WorldWideScience

Sample records for advanced nuclear fuels

  1. Nuclear propulsion technology advanced fuels technology

    Science.gov (United States)

    Stark, Walter A., Jr.

    1993-01-01

    Viewgraphs on advanced fuels technology are presented. Topics covered include: nuclear thermal propulsion reactor and fuel requirements; propulsion efficiency and temperature; uranium fuel compounds; melting point experiments; fabrication techniques; and sintered microspheres.

  2. Fuels for Advanced Nuclear Energy Systems

    International Nuclear Information System (INIS)

    Fuels for advanced nuclear reactors differ greatly from conventional light water reactor fuels and vary widely between the different concepts, due differences in reactor architecture and deployment. Functional requirements of all fuel designs include (1) retention of fission products and fuel nuclides, (2) dimensional stability, and (3) maintaining a coolable geometry. In all cases, the anticipated fuel performance under normal or off-normal conditions is the limiting factor in reactor system design, and cumulative effects of increased exposure to higher burnup degrades fuel performance. In high-temperature (thermal) gas reactor systems, fuel particles of uranium dioxide or uranium oxycarbide particles are coated with layers of carbon and SiC (or ZrC). Such fuels have been used successfully to very high burnup (10-20% of heavy-metal atoms) and can withstand transient temperatures up to 1600 C. Oxide (pellet-type) and metal (pin-type) fuels clad in stainless steel tubes have been successfully used in liquid metal cooled fast reactors, attaining burnup of 20% or more of heavy-metal atoms. Those fuel designs are being adapted for actinide management missions, requiring greater contents of minor actinides (e.g. Am, Np, Cm). The current status of each fuel system is reviewed and technical challenges confronting the implementation of each fuel in the context of the entire advanced reactor fuel cycle (fabrication, reactor performance, recycle) are discussed

  3. Equipment system for advanced nuclear fuel development

    International Nuclear Information System (INIS)

    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

  4. Computational Design of Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Savrasov, Sergey [Univ. of California, Davis, CA (United States); Kotliar, Gabriel [Rutgers Univ., Piscataway, NJ (United States); Haule, Kristjan [Rutgers Univ., Piscataway, NJ (United States)

    2014-06-03

    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.

  5. Options for treatment of legacy and advanced nuclear fuels

    OpenAIRE

    Maher, Christopher John

    2014-01-01

    The treatment of advanced nuclear fuels is relevant to the stabilisation of legacy spent fuels or nuclear materials and fuels from future nuclear reactors. Historically, spent fuel reprocessing has been driven to recover uranium and plutonium for reuse. Future fuel cycles may also recover the minor actinides neptunium, americium and perhaps curium. These actinides would be fabricated into new reactor fuel to produce energy and for transmutation of the minor actinides. This has the potential t...

  6. TALSPEAK Chemistry in Advanced Nuclear Fuel Cycles

    International Nuclear Information System (INIS)

    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)

  7. Technology Readiness Levels for Advanced Nuclear Fuels and Materials Development

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2014-01-01

    The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Advanced nuclear fuels and materials development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Advanced Fuel Campaign (AFC).

  8. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    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

  9. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    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{sup −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.

  10. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ackerman, J.P.; McPheeters, C.C.

    1995-12-31

    More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed.

  11. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

  12. Selection and development of advanced nuclear fuel products

    International Nuclear Information System (INIS)

    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)

  13. The AMP (Advanced MultiPhysics) Nuclear Fuel Performance code

    International Nuclear Information System (INIS)

    Highlights: ► New, three-dimensional, parallel, multi-physics code to simulate fuel behavior in nominal operation. ► Fully-coupled thermomechanics for nominal operation and operation during transients. ► Isotopic depletion using Scale/ORIGEN-S within a fuel performance code. ► Leveraging of existing, validated material models from existing fuel performance codes. ► Initial validation evaluation of an advanced modeling and simulation code for fuel performance. - Abstract: The AMP (Advanced MultiPhysics) Nuclear Fuel Performance code is a new, three-dimensional, multi-physics tool that uses state-of-the-art solution methods and validated nuclear fuel models to simulate the nominal operation and anticipated operational transients of nuclear fuel. The AMP Nuclear Fuel Performance code leverages existing validated material models from traditional fuel performance codes and the Scale/ORIGEN-S spent-fuel characterization code to provide an initial capability that is shown to be sufficiently accurate for a single benchmark problem and anticipated to be accurate for a broad range of problems. The thermomechanics foundation can be solved in a time-dependent or quasi-static approach with any variation of operator-split or fully-coupled solutions at each time step through interoperable interfaces to leading computational mathematics tools, including PETSc, Trilinos, and SUNDIALS. A baseline validation of the AMP Nuclear Fuel Performance code has been performed through the modeling of an experiment in the Halden Reactor Project (IFA-432) that demonstrates the integrated capability and provides a baseline of the initial accuracy of the software.

  14. Advanced nuclear fuel for VVER reactors. Status and operation experience

    International Nuclear Information System (INIS)

    The paper discusses the major VVER fuel trends, aimed at the enhancement of FAs' effectiveness and reliability, flexibility of their operating performances and fuel cycle efficiency, specifically: (i) Fuel burnup increasing is one of the major objectives during the development of improved nuclear fuel and fuel cycles. At present, the achieved fuel rod burn up is 65 MWdays/kgU. The tasks are set and the activities are carried out to achieve fuel rod burnup up to 70 MWdays/kgU and burnup of discharged batch of FAs - up to 60 MWdays/kgU. (ii) Improvement of FA rigidity enables to increase operating reliability of fuel due to gaps reducing between FAs and, as a result, the fall of peak load coefficients. FA geometric stability enables to optimize the speed of handling procedures with fuel. (iii) Increasing of uranium content of FA is aimed at extension of fuel cycles' duration. Fuel weight increase in FA is achieved both due to fuel column height extension and to changes of pellet geometrical size. (iv) Extension of FA service live satisfies the up-to-date NPP requirements for fuel cycles of various duration from 4x320 eff. days to 5x320 eff. days and 3x480 eff. days. (v) The development of new-generation FAs with increased strength characteristics has required the zirconium alloys' improvement. Advanced zirconium alloys shall provide safety and effectiveness of FA and fuel rods during long-life operation up to 40 000 eff. hours. (vi) Utilization of reprocessed uranium enables to use spent nuclear fuel in cycle and to create the partly complete fuel cycle for VVER reactors. This paper summarizes the major operating results of LTAs, which meet the modern and prospective requirements for VVER fuel, at Russian NPPs with VVER-440 and VVER-1000 reactors. (author)

  15. Construction and engineering report for advanced nuclear fuel development facility

    International Nuclear Information System (INIS)

    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. 3,618m2, 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

  16. Construction and engineering report for advanced nuclear fuel development facility

    Energy Technology Data Exchange (ETDEWEB)

    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{sup 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.

  17. Experiences and Trends of Manufacturing Technology of Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    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% 238U + ∼0.7% 235U) and natural thorium (∼100% 232Th) resources. Currently, most power and research reactors use 235U, the only fissile isotope found in nature, as fuel. The fertile isotopes 238U and 232Th are transmuted in the reactor to human-made 239Pu and 233U 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 238U and 235U. 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% PuO2, 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 for sodium

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

    Science.gov (United States)

    Carlsen, Robert W.

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

  19. Feasibility study of advanced fuel burning nuclear reactors

    International Nuclear Information System (INIS)

    An investigation has been conducted to determine both physics, engineering and economic aspects of fusion power reactors based on magnetic confinement and on burning advanced fuels (AFs). DT burning Tokamaks are taken as reference concept. We show that the attractive features of advanced fuels, in particular of neutronlean proton-based AFs, can be combined, in appropriately designed AF reactors (high beta), with power densities comparable to or even higher than those achievable in DT Tokamaks. Moreover we identify physical requirements which would assure Q values well above unity. As an example a semi-open confinement scheme is analyzed based on a self-consistent plasma calculation. We find that a mirror, even if only ''semi-open'' as a result of strong diamagnetism, can barely be expected to achieve high Q values. Therefore confinement schemes such as compact tori, multipole surmacs etc. may be required to burn AFs. We conclude that the economics of AF reactors, as determined by the nuclear boiler power density, may be superior to that of DT-rectors if low recirculating power fractions can be obtained by appropriate plasma tayloring (high fractional transfer of fusion power to ions required). A more detailed investigation is suggested for proton-based fuel cycles. (orig.)

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

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

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

  1. 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.

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

    International Nuclear Information System (INIS)

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

  3. 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.

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

    International Nuclear Information System (INIS)

    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)

  5. 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

  6. Performance of advanced high-temperature fuels for nuclear propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Stark, W.A.; Butt, D.P.; Storms, E.K.; Wallace, T.C. [Los Alamos National Lab., NM (United States)

    1994-12-31

    Nuclear propulsion using hydrogen has been demonstrated to operate at nearly twice the performance level of today`s chemical rockets. However, higher temperatures lead to a variety of degradations that compromise safety and longevity. Foremost among these is the melting of the propulsion reactor fuel. The melting behaviour of the U-Zr-C and U-Nb-C systems have been evaluated.

  7. Advanced methods of quality control in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 CO2 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 planned

  9. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  10. Advanced nuclear fuel study for the utilization of carbon-coated

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee Unviersity, Seoul (Korea)

    1998-03-01

    Advanced nuclear fuel design of carbon coated fuel particles(UCO fuel) was suggested to the current PWRs. Nuclear feasibility studying was forformed for the double heterogeneous UCO fuel by CASMO-3. UCO fuel showed nuclear feasibility when they were packed in the Ulchin3/4 fuel assembly. Nuclear safety was evaluated for the UCO fuel by FTC an dMTC, which had enough safety at operating condition. The average fuel temperature compared with conventional oxide fuel at hot full power condition was reduced by 150 deg K, which was caused by high conductivity of carbon matrix. A core design, used UCO fuel, was possible for same forformance with Ulchin3/4. But, UCO fuel enrichment exceed the PWR fuel enrichment limit 5w/o. Cycle length of UCO duel core was shortened by 90 EFPD satisfied with enrichment limit and thermal power. It is not good for using UCO fuel in PWRs in respect of fuel costs. (author). 19 refs., 71 figs., 25 tabs.

  11. The nuclear fuel cycle with advanced reactor systems - analysis of its economic fundamentals and possibilities

    International Nuclear Information System (INIS)

    The purpose of the study is to analyse the nuclear fuel cycle of alternative advanced reactor systems with respect to their different mass flows of nuclear fuel and to judge the economic feasibility of these advanced nuclear technologies using a specific fuel cycle model. It is the particular importance of this subject that many technical, physical, political and economic coherences are combined in a very complex manner. A detailed description of the problem is given in the introductional chapter 1. The following chapter 2 gives a sufficient survey of the different techniques and technical facilities of the nuclear fuel cycles in question. Part 3 includes an investigation of logical coherences between typical fuel cycle mass flows which consequently leads to a mathematical model. This model is described in part 4. Chapter 5 then deals with the application of this model by the quantitative estimation and valuation of the economic differences between the conventional and advanced nuclear technology. In the final part of this study the influence of a very important parameter in this context, the price of plutonium, is discussed with respect to the time of introduction of the advanced reactor technology under economic conditions. (orig.)

  12. 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

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

    International Nuclear Information System (INIS)

    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

  14. Neutronics analyses for fast spectrum nuclear systems and scenario studies for advanced nuclear fuel cycles

    OpenAIRE

    Grasso, Giacomo

    2010-01-01

    The present PhD thesis summarizes the three-years study about the neutronic investigation of a new concept nuclear reactor aiming at the optimization and the sustainable management of nuclear fuel in a possible European scenario. A new generation nuclear reactor for the nuclear reinassance is indeed desired by the actual industrialized world, both for the solution of the energetic question arising from the continuously growing energy demand together with the corresponding reduction of oil ava...

  15. Nuclear fuel

    International Nuclear Information System (INIS)

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts. (Kako, I.)

  16. Microwave Processing of Simulated Advanced Nuclear Fuel Pellets

    International Nuclear Information System (INIS)

    Throughout the three-year project funded by the Department of Energy (DOE) and lead by Virginia Tech (VT), project tasks were modified by consensus to fit the changing needs of the DOE with respect to developing new inert matrix fuel processing techniques. The focus throughout the project was on the use of microwave energy to sinter fully stabilized zirconia pellets using microwave energy and to evaluate the effectiveness of techniques that were developed. Additionally, the research team was to propose fundamental concepts as to processing radioactive fuels based on the effectiveness of the microwave process in sintering the simulated matrix material.

  17. Microwave Processing of Simulated Advanced Nuclear Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    D.E. Clark; D.C. Folz

    2010-08-29

    Throughout the three-year project funded by the Department of Energy (DOE) and lead by Virginia Tech (VT), project tasks were modified by consensus to fit the changing needs of the DOE with respect to developing new inert matrix fuel processing techniques. The focus throughout the project was on the use of microwave energy to sinter fully stabilized zirconia pellets using microwave energy and to evaluate the effectiveness of techniques that were developed. Additionally, the research team was to propose fundamental concepts as to processing radioactive fuels based on the effectiveness of the microwave process in sintering the simulated matrix material.

  18. Current Status of Advanced Nuclear Fuel Cycle technologies

    International Nuclear Information System (INIS)

    To expand the use of nuclear energy, SNF from nuclear power plants must be managed in a safe and environmental friendly and the problem of decreasing uranium should be solved. To resolve this, a dry processing technology Pyroprocessing is focused on. The government started to develop of Pyroprocessing technology in 1997. According to the decision of government based of Atomic Energy Commission in December 2008, the Korea Atomic Energy Research Institute will construct PRIDE (Pyroprocess Integrated Inactive DEmonstration Facility) by 2011 to prove a consistent process. If Pyroprocessing technology will be developed in the near future, the size of radioactive waste disposal site can be reduced to 100 times compared to the direct disposal. When this technology will be connected to Fast Reactor. high level nuclear waste management of Hundreds of thousands of years may be reduced to hundreds years. However for the commercialization of Pyroprocessing technology, there are some problems to solve. First, because of none commercial facilities in the world of executive experience, so that the facility design, measurement. management and material flow, the critical need for data accumulation. Second, High-level nuclear waste have been known to generate more than the wet methods, it should continue to reduce technology development. In addition, a careful consideration of the residual uranium generating on process also can maximize the efficiency of reducing. The new concept is being developed in Korea Atomic Energy Research Institute Pyroprocessing technology and nuclear waste processing technology to overcome these drawbacks sUQQested a way

  19. Nuclear fuel

    International Nuclear Information System (INIS)

    All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.)

  20. Development of advanced nuclear fuels in the Indian context: advantages and challenges

    International Nuclear Information System (INIS)

    The ever increasing demand on power requirement in the country has opened up need for exploring use of nuclear fuels that could meet such demands. This makes the mission of the department to shift from the first stage of nuclear programme employing natural uranium in PHWRs to the second stage of deploying a large number of fast reactors with plutonium based fuels capable of realising high breeding ratios in addition to energy production. The transition to fast reactors with advanced fuels, capable of higher breeding ratio, opens up a number of scientific and technological challenges in design and operation of such fast reactors. In the Indian context, after successful demonstration of natural uranium based PHWRs, the performance of U-Pu based carbide fuel, as a unique experience in the world, has been demonstrated in FBTR at Kalpakkam. This paper deals with the performance of carbide fuel in FBTR and the programme on development of metallic fuels with appreciably high breeding ratio that would result in considerable reduction in doubling time thereby addressing the increasing demands of power production as well as pave way for introduction of a large number of such fast reactors to provide energy security to the country. The advantages of introduction of metallic fuels as well as the scientific and technological challenges to be faced in doing so and the ongoing efforts towards metallic fuel development are also described in the paper. (author)

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

    International Nuclear Information System (INIS)

    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.

  2. 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.

  3. Radiation Damage in Nuclear Fuel for Advanced Burner Reactors: Modeling and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Niels Gronbech; Asta, Mark; Ozolins, Nigel Browning' Vidvuds; de Walle, Axel van; Wolverton, Christopher

    2011-12-29

    The consortium has completed its existence and we are here highlighting work and accomplishments. As outlined in the proposal, the objective of the work was to advance the theoretical understanding of advanced nuclear fuel materials (oxides) toward a comprehensive modeling strategy that incorporates the different relevant scales involved in radiation damage in oxide fuels. Approaching this we set out to investigate and develop a set of directions: 1) Fission fragment and ion trajectory studies through advanced molecular dynamics methods that allow for statistical multi-scale simulations. This work also includes an investigation of appropriate interatomic force fields useful for the energetic multi-scale phenomena of high energy collisions; 2) Studies of defect and gas bubble formation through electronic structure and Monte Carlo simulations; and 3) an experimental component for the characterization of materials such that comparisons can be obtained between theory and experiment.

  4. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Science.gov (United States)

    2010-01-01

    ... nuclear waste contained in the shipment, as specified in the regulations of DOT in 49 CFR 172.202 and 172... fuel and nuclear waste. 71.97 Section 71.97 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b),...

  5. Advanced Reactor Technology Options for Utilization and Transmutation of Actinides in Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Renewed interest in the potential of nuclear energy to contribute to a sustainable worldwide energy mix is strengthening the IAEA's statutory role in fostering the peaceful uses of nuclear energy, in particular the need for effective exchanges of information and collaborative research and technology development among Member States on advanced nuclear power technologies (Articles III-A.1 and III-A.3). The major challenges facing the long term development of nuclear energy as a part of the world's energy mix are improvement of the economic competitiveness, meeting increasingly stringent safety requirements, adhering to the criteria of sustainable development, and public acceptability. The concern linked to the long life of many of the radioisotopes generated from fission has led to increased R and D efforts to develop a technology aimed at reducing the amount of long lived radioactive waste through transmutation in fission reactors or accelerator driven hybrids. In recent years, in various countries and at an international level, more and more studies have been carried out on advanced and innovative waste management strategies (i.e. actinide separation and elimination). Within the framework of the Project on Technology Advances in Fast Reactors and Accelerator Driven Systems (http://www.iaea.org/inisnkm/nkm/aws/fnss/index.html), the IAEA initiated a number of activities on utilization of plutonium and transmutation of long lived radioactive waste, accelerator driven systems, thorium fuel options, innovative nuclear reactors and fuel cycles, non-conventional nuclear energy systems, and fusion/fission hybrids. These activities are implemented under the guidance and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Fast Reactors (TWG-FR). This publication compiles the analyses and findings of the Coordinated Research Project (CRP) on Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste (2002

  6. Enterprise SRS: Leveraging Ongoing Operations to Advance Nuclear Fuel Cycle Programs - 12579

    International Nuclear Information System (INIS)

    The international leadership in nuclear technology development and deployment long held by the United States has eroded due to the lack of clear national strategies for advanced reactor fuel cycle concepts and for nuclear materials management, as well as to the recent policy decision that halts work on the nuclear fuel repository at Yucca Mountain. Although no national consensus on strategy has yet been reached, a number of recent high-profile reviews and workshops have clearly highlighted a national need for robust research, development and deployment (RD and D) programs in key areas of nuclear technology, especially nuclear separations science and engineering. Collectively, these reviews and workshops provide a picture of the nuclear separations mission needs for three major program offices: Department of Energy Office of-Environmental Management), DOE Office of Nuclear Energy), and the National Nuclear Security Administration (NNSA). While the individual program needs differ significantly in detail and timing, they share common needs in two critical areas of RD and D: - The need for access to and use of multi-purpose engineering-scale demonstration test facilities that can support testing with radioactive material, and - The need for collaborative research enterprises that encompass government research organizations (i.e., national laboratories), commercial industry and the academic community. Such collaborative enterprises effectively integrate theory and modeling with the actual experimental work at all scales, as well as strengthen the technical foundation for research in critical areas. The arguments for engineering-scale collaborative research facilities are compelling. Processing history has shown that test programs and demonstrations conducted with actual nuclear materials are essential to program success. It is widely recognized, however, that such facilities are expensive to build and maintain; creating an imposing, if not prohibitive, financial burden

  7. Evolution of nuclear fuels

    International Nuclear Information System (INIS)

    Nuclear fuel is the primary energy source for sustaining the nuclear fission chain reactions in a reactor. The fuels in the reactor cores are exposed to highly aggressive environment and varieties of advanced fuel materials with improved nuclear properties are continuously being developed to have optimum performance in the existing core conditions. Fabrications of varieties of nuclear fuels used in diverse forms of reactors are mainly based on two naturally occurring nuclear source elements, uranium as fissile 235U and fertile 238U, and thorium as fertile 232Th species. The two metals in the forms of alloys with specific elements, ceramic oxides like MOX and ceramic non-oxide as mixed carbide and nitride with suitable nuclear properties like higher metal density, thermal conductivity, etc. are used as fuels in different reactor designs. In addition, efficiency of various advanced fuels in the forms of dispersion, molten salt and other types are also under investigations. The countries which have large deposits of thorium but limited reserves of uranium, are trying to give special impetus on the development of thorium-based fuels for both thermal and fast reactors in harnessing nuclear energy for peaceful uses of atomic energy. (author)

  8. 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.

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

    International Nuclear Information System (INIS)

    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

  10. Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economic estimates

    International Nuclear Information System (INIS)

    Highlights: • Four fuel cycle scenarios have been analyzed in resources and economic terms. • Scenarios involve Once-Through, Pu burning, and MA transmutation strategies. • No restrictions were found in terms of uranium and plutonium availability. • The best case cost and the impact of their uncertainties to the LCOE were analyzed. - Abstract: Four European fuel cycle scenarios involving transmutation options (in coherence with PATEROS and CP-ESFR EU projects) have been addressed from a point of view of resources utilization and economic estimates. Scenarios include: (i) the current fleet using Light Water Reactor (LWR) technology and open fuel cycle, (ii) full replacement of the initial fleet with Fast Reactors (FR) burning U–Pu MOX fuel, (iii) closed fuel cycle with Minor Actinide (MA) transmutation in a fraction of the FR fleet, and (iv) closed fuel cycle with MA transmutation in dedicated Accelerator Driven Systems (ADS). All scenarios consider an intermediate period of GEN-III+ LWR deployment and they extend for 200 years, looking for long term equilibrium mass flow achievement. The simulations were made using the TREVOL code, capable to assess the management of the nuclear mass streams in the scenario as well as economics for the estimation of the levelized cost of electricity (LCOE) and other costs. Results reveal that all scenarios are feasible according to nuclear resources demand (natural and depleted U, and Pu). Additionally, we have found as expected that the FR scenario reduces considerably the Pu inventory in repositories compared to the reference scenario. The elimination of the LWR MA legacy requires a maximum of 55% fraction (i.e., a peak value of 44 FR units) of the FR fleet dedicated to transmutation (MA in MOX fuel, homogeneous transmutation) or an average of 28 units of ADS plants (i.e., a peak value of 51 ADS units). Regarding the economic analysis, the main usefulness of the provided economic results is for relative comparison of

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

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-09-01

    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

  12. FY2001 Final Report Laboratory Directed Research and Development (LDRD) on Advanced Nuclear Fuel Design in the Future Nuclear Energy Market

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, D.; Choi, J.-S.; DiSabatino, A.; Wirth, B.

    2001-09-30

    This study is to research the maturity of advanced nuclear fuel and cladding technology and to explore the suitability of existing technology for addressing the emerging requirements for Generation IV reactors and emerging thermal/fast spectrum reactors, while simultaneously addressing nuclear waste management, and proliferation resistance concerns.

  13. Some strategic considerations on the development of advance nuclear fuel cycle technologies in China

    International Nuclear Information System (INIS)

    The characteristics of the different fuel cycle options are analyzed from the view point of sustainable development of nuclear fission energy. It is pointed out that the 'once-through' option of fuel cycle does not comply with the sustainability of the nuclear energy development. For the sake of full utilization of uranium resources and the minimization of nuclear waste, the closed fuel cycle of fast breeder reactor is the fundamental way out for the sustainable development of nuclear fission energy. Based on the wide investigations on the present status and R and D trends of the key technologies of fuel cycle both at home and abroad, the strategy for developing China's fuel cycle technologies is explored, some important measures to be taken for achieving the above strategic goal are suggested. (authors)

  14. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    Science.gov (United States)

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  15. The Path to Sustainable Nuclear Energy. Basic and Applied Research Opportunities for Advanced Fuel Cycles, September 12-14, 2005

    International Nuclear Information System (INIS)

    The objective of this report is to identify new basic science that will be the foundation for advances in nuclear fuel-cycle technology in the near term, and for changing the nature of fuel cycles and of the nuclear energy industry in the long term. The goals are to enhance the development of nuclear energy, to maximize energy production in nuclear reactor parks, and to minimize radioactive wastes, other environmental impacts, and proliferation risks. The limitations of the once-through fuel cycle can be overcome by adopting a closed fuel cycle, in which the irradiated fuel is reprocessed and its components are separated into streams that are recycled into a reactor or disposed of in appropriate waste forms. The recycled fuel is irradiated in a reactor, where certain constituents are partially transmuted into heavier isotopes via neutron capture or into lighter isotopes via fission. Fast reactors are required to complete the transmutation of long-lived isotopes. Closed fuel cycles are encompassed by the Department of Energy?s Advanced Fuel Cycle Initiative (AFCI), to which basic scientific research can contribute. Two nuclear reactor system architectures can meet the AFCI objectives: a ?single-tier? system or a ?dual-tier? system. Both begin with light water reactors and incorporate fast reactors. The ?dual-tier? systems transmute some plutonium and neptunium in light water reactors and all remaining transuranic elements (TRUs) in a closed-cycle fast reactor. Basic science initiatives are needed in two broad areas: ? Near-term impacts that can enhance the development of either ?single-tier? or ?dual-tier? AFCI systems, primarily within the next 20 years, through basic research. Examples: Dissolution of spent fuel, separations of elements for TRU recycling and transmutation Design, synthesis, and testing of inert matrix nuclear fuels and non-oxide fuels Invention and development of accurate on-line monitoring systems for chemical and nuclear species in the nuclear

  16. Advanced fuel technology and performance

    International Nuclear Information System (INIS)

    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

  17. Advances in Nuclear Energy

    Science.gov (United States)

    Frois, B.

    2005-04-01

    This paper briefly reviews the next generations of nuclear reactors and the perspectives of development of nuclear energy. Advanced reactors will progressively replace the existing ones during the next two decades. Future systems of the fourth generation are planned to be built beyond 2030. These systems have been studied in the framework of the "Generation IV" International Forum. The goals of these systems is to have a considerable increase in safety, be economically competitive and produce a significantly reduced volume of nuclear wastes. The closed fuel cycle is preferred.

  18. 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.

  19. 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.

  20. Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

    Science.gov (United States)

    Bowles, K. J.

    1974-01-01

    An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

  1. Design concepts and advanced manipulator development for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    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

  2. Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tome, Carlos N [Los Alamos National Laboratory; Caro, J A [Los Alamos National Laboratory; Lebensohn, R A [Los Alamos National Laboratory; Unal, Cetin [Los Alamos National Laboratory; Arsenlis, A [LLNL; Marian, J [LLNL; Pasamehmetoglu, K [INL

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, 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 to develop predictive tools is critical. Not only are fabrication and performance models needed 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. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.

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

    International Nuclear Information System (INIS)

    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

  4. The JRC-ITU approach to the safety of advanced nuclear fuel cycles

    International Nuclear Information System (INIS)

    The JRC-ITU safety studies of advanced fuels and cycles adopt two main axes. First the full exploitation of still available and highly relevant knowledge and samples from past fuel preparation and irradiation campaigns (complementing the limited number of ongoing programmes). Secondly, the shift of focus from simple property measurement towards the understanding of basic mechanisms determining property evolution and behaviour of fuel compounds during normal, off-normal and accident conditions. The final objective of the second axis is the determination of predictive tools applicable to systems and conditions different from those from which they were derived. State of the art experimental facilities, extensive networks of partnerships and collaboration with other organizations worldwide, and a developing programme for training and education are essential in this approach. This strategy has been implemented through various programs and projects. The SUPERFACT programme constitutes the main body of existing knowledge on the behavior in-pile of MOX fuel containing minor actinides. It encompassed all steps of a closed fuel cycle. Another international project investigating the safety of a closed cycle is METAPHIX. In this case a U-Pu19-Zr10 metal alloy containing Np, Am and Cm constitutes the fuel. 9 test pins have been prepared and irradiated. In addition to the PIE (Post Irradiation Examination), pyrometallurgical separation of the irradiated fuel has been performed, to demonstrate all the steps of a multiple recycling closed cycle and characterize their safety relevant aspects. Basic studies like thermodynamic fuel properties, fuel-cladding-coolant interactions have also been carried out at JRC-ITU

  5. Advanced containment research for the Canadian Nuclear Fuel Waste Management Program

    International Nuclear Information System (INIS)

    This document outlines the program on the development of advanced containment systems for the disposal of used fuel in a vault deep in plutonic rock. Possible advanced containment concepts, the strategy adopted in selecting potential container materials, and experimental programs currently underway or planned are presented. Most effort is currently directed toward developing long-term containment systems based on non-metallic materials and massive metal containers. The use of additional independent barriers to extend the lifetime of simple containment systems is also being evaluated. 58 refs

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

    Science.gov (United States)

    Mella, R.; Wenman, M. R.

    2013-06-01

    Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of

  7. FURTHER ASSESSMENTS OF THE ATTRACTIVENESS OF MATERIALS IN ADVANCED NUCLEAR FUEL CYCLES FROM A SAFEGUARDS PERSPECTIVE

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C. G.; Jarvinen, G. D.; Wallace, R. K.; Ireland, J. R.; Johnson, M. W.; Sleaford, Brad W.; Ebbinghaus, B. B.; Bradley, Keith S.; Collins, Brian A.; Smith, Brian W.; Prichard, Andrew W.

    2008-10-01

    This paper summarizes the results of an extension to an earlier study [ ] that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with the PUREX, UREX+, and COEX reprocessing schemes. This study focuses on the materials associated with the UREX, COEX, THOREX, and PYROX reprocessing schemes. This study also examines what is required to render plutonium as “unattractive.” Furthermore, combining the results of this study with those from the earlier study permits a comparison of the uranium and thorium based fuel cycles on the basis of the attractiveness of the SNM associated with each fuel cycle. Both studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of “attractiveness levels” that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities [ ]. The methodology and key findings will be presented. Additionally, how these attractiveness levels relate to proliferation resistance (e.g. by increasing impediments to the diversion, theft, undeclared production of SNM for the purpose of acquiring a nuclear weapon), and how they could be used to help inform policy makers, will be discussed.

  8. Further assessments of the attractiveness of materials in advanced nuclear fuel cycles from a safeguards perspective

    International Nuclear Information System (INIS)

    This paper summarises the results of an extension to an earlier study [1] that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with the Purex, Urex+ and COEX reprocessing schemes. This study focuses on the materials associated with the Urex, COEX, Thorex and PYROX reprocessing schemes. This study also examines what is required to render plutonium as 'unattractive.' Furthermore, combining the results of this study with those from the earlier study permits a comparison of the uranium- and thorium-based fuel cycles on the basis of the attractiveness of the SNM associated with each fuel cycle. Both studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of 'attractiveness levels' that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities [2]. The methodology and key findings will be presented. Additionally, how these attractiveness levels relate to proliferation resistance (e.g. by increasing impediments to the diversion, theft, undeclared production of SNM for the purpose of acquiring a nuclear weapon), and how they could be used to help inform policy makers, will be discussed. (authors)

  9. 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.

  10. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  11. Nuclear fuel lease accounting

    International Nuclear Information System (INIS)

    The subject of nuclear fuel lease accounting is a controversial one that has received much attention over the years. This has occurred during a period when increasing numbers of utilities, seeking alternatives to traditional financing methods, have turned to leasing their nuclear fuel inventories. The purpose of this paper is to examine the current accounting treatment of nuclear fuel leases as prescribed by the Financial Accounting Standards Board (FASB) and the Federal Energy Regulatory Commission's (FERC's) Uniform System of Accounts. Cost accounting for leased nuclear fuel during the fuel cycle is also discussed

  12. Safety related issues of spent nuclear fuel storage : summary of a NATO advanced research workshop

    International Nuclear Information System (INIS)

    Full text: A NATO Advanced Research Workshop was held in Almaty, Kazakhstan, in September 2005. The Workshop was co-sponsored by the IAEA and was concerned with the safety issues associated with spent fuel and waste from three types of reactor: research reactors with Al alloy-clad dispersion fuels, fast reactors with stainless steel-clad UO2, and commercial light-water reactors with Zr alloy-clad UO2. Fifteen presentations dealt with research reactors, five with the BN-350 fast reactor in Kazakhstan-shut down and in decommissioning, and two with commercial reactors in the U.S. and Ukraine. With 657 research reactors built and 274 still operational, corrosion of Al-clad research reactor spent fuel during wet storage was a major subject for discussion. Programs at the IAEA, in the U.S., and elsewhere, have actively studied corrosion of Al-clad fuel since the 1980s and the major mechanisms for aqueous corrosion of both spent fuel and of spent-fuel-pool structural components appear to be now well understood, as are the procedures required to limit corrosion. Nonetheless, avoiding corrosion requires vigilance in monitoring and controlling water quality. Measures to ensure water quality are now being taken at operating research reactors, but are difficult to impose at reactors that are shutdown, where there is less funding (or staff) for the task. It was noted there are about 62,000 spent research reactor fuel assemblies-most of them in wet storage-at many reactor sites around the world, three-quarters in industrialized nations, the remainder in developing countries. Dry storage of research reactor fuel is also being used or actively considered in France, Poland, Russia and the U.S. A variant of simple dry storage-the 'melt-and-dilute' option-casts the spent research reactor fuel with natural U into steel canisters to produce a corrosion-resistant low-enrichment fuel configuration which is suitable for safe long-term storage. The main safety issue of spent fast reactor

  13. MITRA an advanced code to calculate radionuclide release from nuclear fuels under general irradiation conditions

    International Nuclear Information System (INIS)

    The paper presents the computer code Mitra (Multicomponent isotope transport) which has been constructed to calculate the release of radioactive fission products from nuclear fuels under non-stationary conditions. The code is based on a new integration method fo the mass transport equation in the presence of precipitation, re-solution and radioactive decay. The starting equations and the assumed physical models are briefly described in the main part of the report. A very detailed description of the formulae used and of the Mitra subprograms are presented in extended appendices

  14. THE ATTRACTIVENESS OF MATERIALS IN ADVANCED NUCLEAR FUEL CYCLES FOR VARIOUS PROLIFERATION AND THEFT SCENARIOS

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C. G.; Ebbinghaus, Bartley B.; Collins, Brian A.; Sleaford, Brad W.; Hase, Kevin R.; Robel, Martin; Wallace, R. K.; Bradley, Keith S.; Ireland, J. R.; Jarvinen, G. D.; Johnson, M. W.; Prichard, Andrew W.; Smith, Brian W.

    2012-08-29

    We must anticipate that the day is approaching when details of nuclear weapons design and fabrication will become common knowledge. On that day we must be particularly certain that all special nuclear materials (SNM) are adequately accounted for and protected and that we have a clear understanding of the utility of nuclear materials to potential adversaries. To this end, this paper examines the attractiveness of materials mixtures containing SNM and alternate nuclear materials associated with the plutonium-uranium reduction extraction (Purex), uranium extraction (UREX), coextraction (COEX), thorium extraction (THOREX), and PYROX (an electrochemical refining method) reprocessing schemes. This paper provides a set of figures of merit for evaluating material attractiveness that covers a broad range of proliferant state and subnational group capabilities. The primary conclusion of this paper is that all fissile material must be rigorously safeguarded to detect diversion by a state and must be provided the highest levels of physical protection to prevent theft by subnational groups; no 'silver bullet' fuel cycle has been found that will permit the relaxation of current international safeguards or national physical security protection levels. The work reported herein has been performed at the request of the U.S. Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for, the nuclear materials in DOE nuclear facilities. The methodology and findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security is discussed.

  15. ADVANCED FUELS CAMPAIGN 2013 ACCOMPLISHMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-10-01

    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 “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. Accomplishments made during fiscal year (FY) 2013 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.

  16. Development of a CVD silica coating for UK advanced gas-cooled nuclear reactor fuel pins

    International Nuclear Information System (INIS)

    Vapour deposited silica coatings could extend the life of the 20% Cr/25% Ni niobium stabilised (20/25/Nb) stainless steel fuel cladding of the UK advanced gas cooled reactors. A CVD coating process developed originally to be undertaken at atmospheric pressure has now been adapted for operation at reduced pressure. Trials on the LP CVD process have been pursued to the production scale using commercial equipment. The effectiveness of the LP CVD silica coatings in providing protection to 20/25/Nb steel surfaces against oxidation and carbonaceous deposition has been evaluated. (author)

  17. Nuclear criticality safety at global nuclear fuel

    International Nuclear Information System (INIS)

    Nuclear criticality safety is the art and science of preventing or terminating an inadvertent nuclear chain reaction in non-reactor environment. Nuclear criticality safety as part of integrated safety program in the nuclear industry is the responsibility of regulators, management and operators. Over the past 36 years, Global Nuclear Fuel (GNF) has successfully developed an integrated nuclear criticality safety program for its BWR fuel manufacturing business. Implementation of this NRC-approved program includes three fundamental elements: administrative practices, controls and training. These elements establish nuclear criticality safety function responsibilities and nuclear criticality safety design criteria in accordance with double contingency principle. At GNF, a criticality safety computational system has been integrated into nuclear criticality safety program as an incredibly valuable tool for nuclear criticality safety design and control applications. This paper describes select elements of GNF nuclear criticality safety program with emphasis being placed on need for clear criticality safety function responsibilities, nuclear safety design criteria and associated double contingency implementation, as well as advanced Monte Carlo neutron transport codes used to derive subcritical safety limits. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-02-01

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

  19. 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.

  20. Uncertainty Analyses of Advanced Fuel Cycles

    International Nuclear Information System (INIS)

    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

  1. The environmental impacts of Korean advanced nuclear fuel cycle KIEP-21 and disposal concepts

    International Nuclear Information System (INIS)

    We have performed a performance assessment to investigate effects of waste forms and repository designs by comparing the case of direct disposal of used PWR fuel in the Korean Reference Repository System (KRS) concept with the case of Advanced Korean Reference Disposal System (A-KRS) repository containing ILW and HLW from the KIEP-21 system. Numerical evaluations have been made for release rates of actinide and fission product isotopes at the boundaries of the engineered barrier system (EBS) and the natural barrier system (NBS) by the TTB code developed at UC Berkeley. Results show that in both cases, most actinides and their daughters remain as precipitates in the EBS because of their assumed low solubilities. The radionuclides that reach the 1 000-m location in NBS are fission products, 129I, 79Se and 36Cl. They have high solubilities and weak or no sorption with the EBS materials or with the host rock, and are released congruently with waste form alteration. In case of direct disposal, a contribution of 2% of iodine is assumed to be accumulated in the gap between the cladding and fuel pellets released after failure of the waste package and cladding dominates the total release rate. With increase in the waste form alteration time, the peak value of total release rate decreases proportionally because the dominant radionuclides are fission product isotopes, which are released from waste forms congruently with waste form dissolution. It has been shown by PHREEQC simulation that actinide solubilities can be significantly affected by pore water chemistry determined by the evolving EBS materials, waste forms and compositions of groundwater from the far field. (authors)

  2. First-principles modelling of radiation defects in advanced nuclear fuels

    Science.gov (United States)

    Kotomin, E. A.; Gryaznov, D.; Grimes, R. W.; Parfitt, D.; Zhukovskii, Yu. F.; Mastrikov, Yu. A.; Van Uffelen, P.; Rondinella, V. V.; Konings, R. J. M.

    2008-06-01

    We present and discuss the results of the first-principles calculations of Frenkel defects and O impurities in uranium mononitride (UN) perspective for fast reactor nuclear fuels. Special attention is paid to the calculation of defect migration energies. We demonstrate that the interstitialcy mechanism (with the formation of a N-N dumbbell along the [1 1 1] axis) is energetically more favorable than the direct [1 0 0] hops. As a result, for the interstitial N ions we predict a diffusion mechanism similar to that known in isostructural fcc materials with a different chemical nature (KCl, MgO). The calculated effective N charge considerably depends on the ion position and environment (a host lattice site, interstitial or saddle point) which strongly limits the applicability of classical defect modelling based on formal invariant charges. Lastly, the calculated migration energy for the interstitial impurity O ions is quite low (2.84 eV), which indicates their high mobility and ability for reactions with other defects.

  3. Effects of nuclear elastic scattering and modifications of ion-electron equilibration power on advanced-fuel burns

    International Nuclear Information System (INIS)

    The effects of Nuclear Elastic Scattering (NES) of fusion products and modifications of the ion-electron equilibration power on D-T and D-based advanced-fuel fusion plasmas are presented here. The processes causing the modifications to the equilibration power included here are: (1) depletion of low-energy electrons by Coulomb collisions with the ions; and (2) magnetic field effects on the energy transfer between the ions and the electrons. Both NES and the equilibration modifications affect the flow of power to the plasma ions, which is an important factor in the analysis of advanced-fuels. A Hot Ion Mode (HIM) analysis was used to investigate the changes in the minimum ignition requirements for Cat-D and D-3He plasmas, due to the changes in the allowable T/sub i/T/sub e/ for ignition from NES and equilibration modifications. Both of these effects have the strongest influence on the ignition requirements for high temperature (>50 keV), low beta (<15%) plasmas, where the cyclotron radiation power loss from the electrons (which is particularly sensitive to changes in the electron temperature) is large

  4. Advances in Development of the Fission Product Extraction Process for the Separation of Cesium and Strontium from Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    JAck D. Law

    2007-09-01

    The Fission Product Extraction (FPEX) Process is being developed as part of the United States Department of Energy Advanced Fuel Cycle Initiative for the simultaneous separation of cesium (Cs) and strontium (Sr) from spent light water reactor (LWR) fuel. Separation of the Cs and Sr will reduce the short-term heat load in a geological repository, and when combined with the separation of americium (Am) and curium (Cm), could increase the capacity of the geological repository by a factor of approximately 100. The FPEX process is based on two highly specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. Results of flowsheet testing of the FPEX process with a simulated feed solution in 3.3-cm centrifugal contactors are detailed. Removal efficiencies, distribution coefficient data, coextraction of metals, and process hydrodynamic performance are discussed along with recommendations for future flowsheet testing with actual spent nuclear fuel.

  5. 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.

  6. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-03-01

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

  7. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  8. Radioactive waste shipments to Hanford retrievable storage from Westinghouse Advanced Reactors and Nuclear Fuels Divisions, Cheswick, Pennsylvania

    International Nuclear Information System (INIS)

    During the next two decades the transuranic (TRU) waste now stored in the burial trenches and storage facilities at the Hanford Sits in southeastern Washington State is to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico for final disposal. Approximately 5.7 percent of the TRU waste to be retrieved for shipment to WIPP was generated by the decontamination and decommissioning (D ampersand D) of the Westinghouse Advanced Reactors Division (WARD) and the Westinghouse Nuclear Fuels Division (WNFD) in Cheswick, Pennsylvania and shipped to the Hanford Sits for storage. This report characterizes these radioactive solid wastes using process knowledge, existing records, and oral history interviews

  9. Radioactive waste shipments to Hanford retrievable storage from Westinghouse Advanced Reactors and Nuclear Fuels Divisions, Cheswick, Pennsylvania

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, D. [Westinghouse Hanford Co., Richland, WA (United States); Pottmeyer, J.A.; Weyns, M.I.; Dicenso, K.D.; DeLorenzo, D.S. [Los Alamos Technical Associates, Inc., NM (United States)

    1994-04-01

    During the next two decades the transuranic (TRU) waste now stored in the burial trenches and storage facilities at the Hanford Sits in southeastern Washington State is to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico for final disposal. Approximately 5.7 percent of the TRU waste to be retrieved for shipment to WIPP was generated by the decontamination and decommissioning (D&D) of the Westinghouse Advanced Reactors Division (WARD) and the Westinghouse Nuclear Fuels Division (WNFD) in Cheswick, Pennsylvania and shipped to the Hanford Sits for storage. This report characterizes these radioactive solid wastes using process knowledge, existing records, and oral history interviews.

  10. Nuclear Fuel Reprocessing

    International Nuclear Information System (INIS)

    This is a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. Nuclear reprocessing is the chemical treatment of spent fuel involving separation of its various constituents. Principally, it is used to recover useful actinides from the spent fuel. Radioactive waste that cannot be re-used is separated into streams for consolidation into waste forms. The first known application of nuclear reprocessing was within the Manhattan Project to recover material for nuclear weapons. Currently, reprocessing has a peaceful application in the nuclear fuel cycle. A variety of chemical methods have been proposed and demonstrated for reprocessing of nuclear fuel. The two most widely investigated and implemented methods are generally referred to as aqueous reprocessing and pyroprocessing. Each of these technologies is described in detail in Section 3 with numerous references to published articles. Reprocessing of nuclear fuel as part of a fuel cycle can be used both to recover fissionable actinides and to stabilize radioactive fission products into durable waste forms. It can also be used as part of a breeder reactor fuel cycle that could result in a 14-fold or higher increase in energy utilization per unit of natural uranium. Reprocessing can also impact the need for geologic repositories for spent fuel. The volume of waste that needs to be sent to such a repository can be reduced by first subjecting the spent fuel to reprocessing. The extent to which volume reduction can occur is currently under study by the United States Department of Energy via research at various national laboratories and universities. Reprocessing can also separate fissile and non-fissile radioactive elements for transmutation.

  11. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

  12. Advanced biological treatment of aqueous effluent from the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Many of the processing steps in the nuclear fuel cycle generate aqueous effluent streams bearing contaminants that can, because of their chemical or radiological properties, pose an environmental hazard. Concentration of such contaminants must be reduced to acceptable levels before the streams can be discharged to the environment. Two classes of contaminants, nitrates and heavy metals, are addressed in this study. Specific techniques aimed at the removal of nitrates and radioactive heavy metals by biological processes are being developed, tested, and demonstrated. Although cost comparisons between biological processes and current treatment methods will be presented, these comparisons may be misleading because biological processes yield environmentally better end results which are difficult to price. The fluidized-bed biological denitrification process is an environmentally acceptable and economically sound method for the disposal of nonreusable sources of nitrate effluents. A very high denitrification rate can be obtained in a FBR as the result of a high concentration of denitrification bacteria in the bioreactor and the stagewise operation resulting from plug flow in the reactor. The overall denitrification rate in an FBR ranges from 20- to 100-fold greater than that observed for an STR bioreactor. It has been shown that the system can be operated using Ca2+, Na+, or NH4+ cations at nitrate concentrations up to 1 g/liter without inhibition. Biological sorption of uranium and other radionuclides (particularly the actinides) from dilute aqueous waste streams shows considerable promise as a means of recovering these valuable resources and reducing the environmental impact, however, further development efforts are required

  13. DEVELOPMENT OF CERAMIC WASTE FORMS FOR AN ADVANCED NUCLEAR FUEL CYCLE

    Energy Technology Data Exchange (ETDEWEB)

    Marra, J.; Billings, A.; Brinkman, K.; Fox, K.

    2010-11-30

    A series of ceramic waste forms were developed and characterized for the immobilization of a Cesium/Lanthanide (CS/LN) waste stream anticipated to result from nuclear fuel reprocessing. Simple raw materials, including Al{sub 2}O{sub 3} and TiO{sub 2} were combined with simulated waste components to produce multiphase ceramics containing hollandite-type phases, perovskites (particularly BaTiO{sub 3}), pyrochlores and other minor metal titanate phases. Three fabrication methodologies were used, including melting and crystallizing, pressing and sintering, and Spark Plasma Sintering (SPS), with the intent of studying phase evolution under various sintering conditions. X-Ray Diffraction (XRD) and Scanning Electron Microscopy coupled with Energy Dispersive Spectroscopy (SEM/EDS) results showed that the partitioning of the waste elements in the sintered materials was very similar, despite varying stoichiometry of the phases formed. Identification of excess Al{sub 2}O{sub 3} via XRD and SEM/EDS in the first series of compositions led to a Phase II study, with significantly reduced Al{sub 2}O{sub 3} concentrations and increased waste loadings. The Phase II compositions generally contained a reduced amount of unreacted Al{sub 2}O{sub 3} as identified by XRD. Chemical composition measurements showed no significant issues with meeting the target compositions. However, volatilization of Cs and Mo was identified, particularly during melting, since sintering of the pressed pellets and SPS were performed at lower temperatures. Partitioning of some of the waste components was difficult to determine via XRD. SEM/EDS mapping showed that those elements, which were generally present in small concentrations, were well distributed throughout the waste forms.

  14. Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Michael F. Simpson; Jack D. Law

    2010-02-01

    This is an a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. No formal abstract was required for the article. The full article will be attached.

  15. Requirements for a Dynamic Solvent Extraction Module to Support Development of Advanced Technologies for the Recycle of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jack Law; Veronica Rutledge; Candido Pereira; Jackie Copple; Kurt Frey; John Krebs; Laura Maggos; Kevin Nichols; Kent Wardle; Pratap Sadasivan; Valmor DeAlmieda; David Depaoli

    2011-06-01

    The Department of Energy's Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program has been established to create and deploy next generation, verified and validated nuclear energy modeling and simulation capabilities for the design, implementation, and operation of future nuclear energy systems to improve the U.S. energy security. As part of the NEAMS program, Integrated Performance and Safety Codes (IPSC's) are being produced to significantly advance the status of modeling and simulation of energy systems beyond what is currently available to the extent that the new codes be readily functional in the short term and extensible in the longer term. The four IPSC areas include Safeguards and Separations, Reactors, Fuels, and Waste Forms. As part of the Safeguards and Separations (SafeSeps) IPSC effort, interoperable process models are being developed that enable dynamic simulation of an advanced separations plant. A SafeSepss IPSC 'toolkit' is in development to enable the integration of separation process modules and safeguards tools into the design process by providing an environment to compose, verify and validate a simulation application to be used for analysis of various plant configurations and operating conditions. The modules of this toolkit will be implemented on a modern, expandable architecture with the flexibility to explore and evaluate a wide range of process options while preserving their stand-alone usability. Modules implemented at the plant-level will initially incorporate relatively simple representations for each process through a reduced modeling approach. Final versions will incorporate the capability to bridge to subscale models to provide required fidelity in chemical and physical processes. A dynamic solvent extraction model and its module implementation are needed to support the development of this integrated plant model. As a stand-alone application, it will also support solvent development of extraction flowsheets

  16. The Economic, repository and proliferation implications of advanced nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Deinert, Mark; Cady, K B

    2011-09-04

    The goal of this project was to compare the effects of recycling actinides using fast burner reactors, with recycle that would be done using inert matrix fuel burned in conventional light water reactors. In the fast reactor option, actinides from both spent light water and fast reactor fuel would be recycled. In the inert matrix fuel option, actinides from spent light water fuel would be recycled, but the spent inert matrix fuel would not be reprocessed. The comparison was done over a limited 100-year time horizon. The economic, repository and proliferation implications of these options all hinge on the composition of isotopic byproducts of power production. We took the perspective that back-end economics would be affected by the cost of spent fuel reprocessing (whether conventional uranium dioxide fuel, or fast reactor fuel), fuel manufacture, and ultimate disposal of high level waste in a Yucca Mountain like geological repository. Central to understanding these costs was determining the overall amount of reprocessing needed to implement a fast burner, or inert matrix fuel, recycle program. The total quantity of high level waste requiring geological disposal (along with its thermal output), and the cost of reprocessing were also analyzed. A major advantage of the inert matrix fuel option is that it could in principle be implemented using the existing fleet of commercial power reactors. A central finding of this project was that recycling actinides using an inert matrix fuel could achieve reductions in overall actinide production that are nearly very close to those that could be achieved by recycling the actinides using a fast burner reactor.

  17. Nuclear fuel element

    Science.gov (United States)

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

    A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

  18. Development of a Code for the Long Term Radiological Safety Assessment of Radioactive Wastes from Advanced Nuclear Fuel Cycle Facilities in Republic of Korea

    International Nuclear Information System (INIS)

    For the purpose of evaluating annual individual doses from a potential repository disposing of radioactive wastes from the operation of the prospective advanced nuclear fuel cycle facilities in Korea, the new safety assessment code based on the Goldsim has been developed. It was designed to compare the environmental impacts from many fuel cycle options such as direct disposal, wet and dry recycling. The code based on the compartment theory can be applied to assess both normal and what if scenarios

  19. Nuclear fuel manufacture

    International Nuclear Information System (INIS)

    The technologies used to manufacture nuclear fuel from uranium ore are outlined, with particular reference to the light water reactor fuel cycle. Capital and operating cost estimates for the processing stages are given, and the relevance to a developing uranium industry in Australia is discussed

  20. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  1. Nuclear power and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The report provides data and assessments of the status and prospects of nuclear power and the nuclear fuel cycle. The report discusses the economic competitiveness of nuclear electricity generation, the extent of world uranium resources, production and requirements, uranium conversion and enrichment, fuel fabrication, spent fuel treatment and radioactive waste management. A review is given of the status of nuclear fusion research

  2. Evaluation of conceptual flowsheets for incorporating Light Water Reactor (LWR) fuel materials in an advanced nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Bell, J.T.; Burch, W.D.; Collins, E.D.; Forsberg, C.W.; Prince, B.E.; Bond, W.D.; Campbell, D.O.; Delene, J.G.; Mailen, J.C.

    1990-08-01

    A preliminary study by a group of experts at ORNL has generated and evaluated a number of aqueous and non-aqueous flowsheets for recovering transuranium actinides from LWR fuel for use as fuel in an LMR and, at the same time, for transmutation of the wastes to less hazardous materials. The need for proliferation resistance was a consideration in the flowsheets. The current state of development of the flowsheets was evaluated and recommendations for additional study were made. 3 refs., 6 figs.

  3. Advanced coated particle fuels

    International Nuclear Information System (INIS)

    The coated particle fuel (cpf) has been developed for use in high-temperature gas-cooled reactors, but it may find applications in other types of reactors. In JAERI, besides the development of cpf for High Temperature Engineering Test Reactor, conceptual studies of the cpf applications in actinide burner reactors and space reactors have been made. The conceptual design studies as well as the research and development of advanced coatings, ZrC and TiN, are reviewed. (author)

  4. The attractiveness of materials in advanced nuclear fuel cycles for various proliferation and theft scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, Charles G [Los Alamos National Laboratory; Wallace, Richard K [Los Alamos National Laboratory; Ireland, John R [Los Alamos National Laboratory; Johnson, M W [Los Alamos National Laboratory; Hase, Kevin R [Los Alamos National Laboratory; Jarvinen, Gordon D [Los Alamos National Laboratory; Ebbinghaus, Bartley B [LLNL; Sleaford, Brad A [LLNL; Bradley, Keith S [LLNL; Collins, Brian W [PNNL; Smith, Brian W [PNNL; Prichard, Andrew W [PNNL

    2009-01-01

    This paper is an extension to earlier studies that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) and alternate nuclear materials (ANM) associated with the PUREX, UREX, COEX, THOREX, and PYROX reprocessing schemes. This study extends the figure of merit (FOM) for evaluating attractiveness to cover a broad range of proliferant state and sub-national group capabilities. The primary conclusion of this study is that all fissile material needs to be rigorously safeguarded to detect diversion by a state and provided the highest levels of physical protection to prevent theft by sub-national groups; no 'silver bullet' has been found that will permit the relaxation of current international safeguards or national physical security protection levels. This series of studies has been performed at the request of the United States Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for nuclear materials in DOE nuclear facilities. The expanded methodology and updated findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security are discussed.

  5. The Attractiveness of Materials in Advanced Nuclear Fuel Cycles for Various Proliferation and Theft Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C. G.; Wallace, R. K.; Ireland, J. R.; Johnson, M. W.; Hase, Kevin R.; Jarvinen, G. D.; Ebbinghaus, B. B.; Sleaford, Brad W.; Bradley, Keith S.; Collins, Brian A.; Smith, Brian W.; Prichard, Andrew W.

    2010-09-01

    This paper is an extension to earlier studies1,2 that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) and alternate nuclear materials (ANM) associated with the PUREX, UREX, COEX, THOREX, and PYROX reprocessing schemes. This study extends the figure of merit (FOM) for evaluating attractiveness to cover a broad range of proliferant state and sub-national group capabilities. The primary conclusion of this study is that all fissile material needs to be rigorously safeguarded to detect diversion by a state and provided the highest levels of physical protection to prevent theft by sub-national groups; no “silver bullet” has been found that will permit the relaxation of current international safeguards or national physical security protection levels. This series of studies has been performed at the request of the United States Department of Energy (DOE) and is based on the calculation of "attractiveness levels" that are expressed in terms consistent with, but normally reserved for nuclear materials in DOE nuclear facilities.3 The expanded methodology and updated findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security are discussed.

  6. The Attractiveness of Materials in Advanced Nuclear Fuel Cycles for Various Proliferation and Theft Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C.G.; Wallace, R.K.; Hase, K.R.; Jarvinen, G.D.; Ireland, J.R.; Johnson, M.W. [Los Alamos National Laboratory, P.O. Box 1663, MS K575, Los Alamos, NM 87545 (United States); Ebbinghaus, B.B.; Sleaford, B.W.; Bradley, K.S. [Lawrence Livermore National Laboratory (United States); Collins, B.A.; Prichard, A.W.; Smith, B.W. [Pacific Northwest National Laboratory (United States)

    2009-06-15

    This paper is an extension to earlier studies that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) and alternate nuclear materials (ANM) associated with the PUREX, UREX, COEX, THOREX, and PYROX reprocessing schemes. This study extends the figure of merit (FOM) for evaluating attractiveness to cover a broad range of proliferant state and sub-national group capabilities. The primary conclusion of this study is that all fissile material needs to be rigorously safeguarded to detect diversion by a state and provided the highest levels of physical protection to prevent theft by sub-national groups; no 'silver bullet' has been found that will permit the relaxation of current international safeguards or national physical security protection levels. This series of studies has been performed at the request of the United States Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for nuclear materials in DOE nuclear facilities. The expanded methodology and updated findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security are discussed. (authors)

  7. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    The papers presented at the International Conference on The Nuclear Fuel Cycle, held at Stockholm, 28 to 31 October 1975, are reviewed. The meeting, organised by the U.S. Atomic Industrial Forum, and the Swedish Nuclear Forum, was concerned more particularly with economic, political, social and commercial aspects than with tecnology. The papers discussed were considered under the subject heading of current status, uranium resources, enrichment, and reprocessing. (U.K.)

  8. Nuclear Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Dale, Deborah J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-10-28

    These slides will be presented at the training course “International Training Course on Implementing State Systems of Accounting for and Control (SSAC) of Nuclear Material for States with Small Quantity Protocols (SQP),” on November 3-7, 2014 in Santa Fe, New Mexico. The slides provide a basic overview of the Nuclear Fuel Cycle. This is a joint training course provided by NNSA and IAEA.

  9. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To increase the fuel assembly rigidity while making balance in view of the dimension thereby improving the earthquake proofness. Constitution: In a nuclear fuel assembly having a control rod guide thimble tube, the gap between the thimble tube and fuel insert (inner diameter of the guiding thimble tube-outer diameter of the fuel insert) is made greater than 1.0 mm. Further, the wall thickness of the thimble tube is made to about 4 - 5 % of the outer diameter, while the flowing fluid pore cross section S in the thimble tube is set as: S = S0 x A0/A where S0: cross section of the present flowing fluid pore, A: effective cross section after improvement, = Π/4(d2 - D2) in which d is the thimble tube inner diameter and the D is the fuel insert outer diameter. A0: present effective cross section. (Seki, T.)

  10. Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Harold F. McFarlane; Terry Todd

    2013-11-01

    Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore. Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor

  11. Development of the advanced nuclear materials -Development of metallic fuel materials

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Ho; Yang, Young Seok; Hwang, Sung Chan; Ju, Keun Sik; Kim, Ki Whan; Ahn, Hyun Seok; Chang, Se Jeong; Kim, Jeong Do [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    In this report, the electrolysis rate and mechanism between uranium and impurities in molten salts including uranium and other impurities were carried out to separate impurities from the molten salts. Uranium chloride preparation process that is developing as a alternative to uranium fluoride process was reviewed and experimental parameters for chlorination process development were determined based on theoretical review. Also, some elementary experiments were carried out to investigate optimum reaction condition for uranium chloride preparation. U-10wt%Zr alloy which has been used as driver fuel or blanket materials was prepared by vacuum induction melting and investigated into the characteristics of fabrication process and alloy properties. In order to develop uranium alloy of high burn-up fuel, ternary X element was prepared. And the effect of ternary alloying elements with high melting point on the alloy phases and ternary alloying elements with high melting point on the alloy phases and microstructures was investigated. On the other hand rapidly solidified powder of U-10wt%Zr alloy was fabricated and characterized. 17 tabs., 52 figs., 39 refs. (Author).

  12. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

  13. Effects of nuclear elastic scattering and modifications of ion-electron equilibration power on advanced-fuel burns

    International Nuclear Information System (INIS)

    The effects of Nuclear Elastic Scattering (NES) of fusion products and modifications of the ion-electron equilibration power on D-T and D-based advanced-fuel fusion plasmas are presented. The processes causing the modifications to the equilibration power included here are: (1) depletion of low-energy electrons by Coulomb collisions with the ions, and (2) magnetic field effects on the energy transfer between the ions and the electrons. A Hot Ion Mode (HIM) analysis was used to investigate the changes in the minimum ignition requirements for Cat-D and D-3He plasmas, due to the changes in the allowable T/sub i//T/sub e/ for ignition from NES and equilibration modifications. Both of these effects have the strongest influence on the ignition requirements for high temperature (>50 keV), low beta (<15%) plasmas, where the cyclotron radiation power loss from the electrons (which is particularly sensitive to changes in the electron temperature) is large

  14. OSMOSE an experimental program for improving neutronic predictions of advanced nuclear fuels.

    Energy Technology Data Exchange (ETDEWEB)

    Klann, R. T.; Aliberti, G.; Zhong, Z.; Graczyk, D.; Loussi, A.; Nuclear Engineering Division; Commissariat a l Energie Atomique

    2007-10-18

    This report describes the technical results of tasks and activities conducted in FY07 to support the DOE-CEA collaboration on the OSMOSE program. The activities are divided into five high-level tasks: reactor modeling and pre-experiment analysis, sample fabrication and analysis, reactor experiments, data treatment and analysis, and assessment for relevance to high priority advanced reactor programs (such as GNEP and Gen-IV).

  15. Nuclear fuel cycle information workshop

    International Nuclear Information System (INIS)

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

  16. AB INITIO STUDY OF ADVANCED METALLIC NUCLEAR FUELS FOR FAST BREEDER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Landa, A; Soderlind, P; Grabowski, B; Turchi, P A; Ruban, A V; Vitos, L

    2012-04-23

    Density-functional formalism is applied to study the ground state properties of {gamma}-U-Zr and {gamma}-U-Mo solid solutions. Calculated heats of formation are compared with CALPHAD assessments. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components. The decomposition curves for {gamma}-based U-Zr and U-Mo solid solutions are derived from Ising-type Monte Carlo simulations. We explore the idea of stabilization of the {delta}-UZr{sub 2} compound against the {alpha}-Zr (hcp) structure due to increase of Zr d-band occupancy by the addition of U to Zr. We discuss how the specific behavior of the electronic density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. The mechanism of possible Am redistribution in the U-Zr and U-Mo fuels is also discussed.

  17. Advancing Inverse Sensitivity/Uncertainty Methods for Nuclear Fuel Cycle Applications

    Science.gov (United States)

    Arbanas, G.; Williams, M. L.; Leal, L. C.; Dunn, M. E.; Khuwaileh, B. A.; Wang, C.; Abdel-Khalik, H.

    2015-01-01

    The inverse sensitivity/uncertainty quantification (IS/UQ) method has recently been implemented in the Inverse Sensitivity/UnceRtainty Estimator (INSURE) module of the AMPX cross section processing system [M.E. Dunn and N.M. Greene, "AMPX-2000: A Cross-Section Processing System for Generating Nuclear Data for Criticality Safety Applications," Trans. Am. Nucl. Soc. 86, 118-119 (2002)]. The IS/UQ method aims to quantify and prioritize the cross section measurements along with uncertainties needed to yield a given nuclear application(s) target response uncertainty, and doing this at a minimum cost. Since in some cases the extant uncertainties of the differential cross section data are already near the limits of the present-day state-of-the-art measurements, requiring significantly smaller uncertainties may be unrealistic. Therefore, we have incorporated integral benchmark experiments (IBEs) data into the IS/UQ method using the generalized linear least-squares method, and have implemented it in the INSURE module. We show how the IS/UQ method could be applied to systematic and statistical uncertainties in a self-consistent way and how it could be used to optimize uncertainties of IBEs and differential cross section data simultaneously. We itemize contributions to the cost of differential data measurements needed to define a realistic cost function.

  18. Advanced nuclear propulsion concepts

    Energy Technology Data Exchange (ETDEWEB)

    Howe, S.D. [Los Alamos National Lab., NM (United States)

    1994-12-31

    A preliminary analysis has been carried out for two potential advanced nuclear propulsion systems: a contained pulsed nuclear propulsion engine and an antiproton initiated ICF system. The results of these studies indicate that both concepts have a high potential to help enable manned planetary exploration but require substantial development.

  19. Impact of Nuclear Data Uncertainties on Advanced Fuel Cycles and their Irradiated Fuel - a Comparison between Libraries

    Science.gov (United States)

    Díez, C. J.; Cabellos, O.; Martínez, J. S.

    2014-04-01

    The uncertainties on the isotopic composition throughout the burnup due to the nuclear data uncertainties are analysed. The different sources of uncertainties: decay data, fission yield and cross sections; are propagated individually, and their effect assessed. Two applications are studied: EFIT (an ADS-like reactor) and ESFR (Sodium Fast Reactor). The impact of the uncertainties on cross sections provided by the EAF-2010, SCALE6.1 and COMMARA-2.0 libraries are compared. These Uncertainty Quantification (UQ) studies have been carried out with a Monte Carlo sampling approach implemented in the depletion/activation code ACAB. Such implementation has been improved to overcome depletion/activation problems with variations of the neutron spectrum.

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

    International Nuclear Information System (INIS)

    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

  1. 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.

  2. Advancing Inverse Sensitivity/Uncertainty Methods for Nuclear Fuel Cycle Applications

    Energy Technology Data Exchange (ETDEWEB)

    Arbanas, Goran [ORNL; Williams, Mark L [ORNL; Leal, Luiz C [ORNL; Dunn, Michael E [ORNL; Khuwaileh, Bassam A. [North Carolina State University; Wang, C [North Carolina State University; Abdel-Khalik, Hany [North Carolina State University

    2015-01-01

    The inverse sensitivity/uncertainty quantification (IS/UQ) method has recently been implemented in the Inverse Sensitivity/UnceRtainty Estimiator (INSURE) module of the AMPX system [1]. The IS/UQ method aims to quantify and prioritize the cross section measurements along with uncertainties needed to yield a given nuclear application(s) target response uncertainty, and doing this at a minimum cost. Since in some cases the extant uncertainties of the differential cross section data are already near the limits of the present-day state-of-the-art measurements, requiring significantly smaller uncertainties may be unrealistic. Therefore we have incorporated integral benchmark experiments (IBEs) data into the IS/UQ method using the generalized linear least-squares method, and have implemented it in the INSURE module. We show how the IS/UQ method could be applied to systematic and statistical uncertainties in a self-consistent way. We show how the IS/UQ method could be used to optimize uncertainties of IBEs and differential cross section data simultaneously.

  3. nuclear fuel design criteria

    International Nuclear Information System (INIS)

    Nuclear fuel design is strictly dependent on reactor type and experiences obtained from performance of nuclear fuels. The objectives of the design are reliability, and economy. Nuclear fuel design requires an interdisciplinary work which has to cover, at least nuclear design, thermalhydraulic design, mechanical design, and material properties.The procedure of design, as describe in the quality assurance, consist of a number of steps. The most important parts are: Design description or inputs, preliminary design, detailed design and design output, and design verification. The first step covers objectives and requirements, as defined by the customer and by the regulatory authority for product performance,environmental factors, safety, etc. The second describes assumptions and alternatives, safety, economy and engineering analyses. The third covers technical specifications, design drawings, selection of QA program category, etc. The most important form of design verification is design review by qualified independent internal or external reviewers. The scope of the review depends on the specific character of the design work. Personnel involved in verification and review do not assume prime responsibility for detecting errors. Responsibility for the design remains with the personnel involved in the design work

  4. Advancing against nuclear terrorism

    International Nuclear Information System (INIS)

    Meeting a day before the summit, Bush and Putin announced a new Global Initiative to Combat Nuclear Terrorism; a plan for multiple, multilateral guaranteed suppliers of nuclear fuel to States that forgo building their own enrichment plants; and a Civil Nuclear Agreement that will lift restrictions on cooperation between the two countries in developing peaceful nuclear power. Each of these initiatives provides a framework for dozens of specific actions that can measurably reduce the risk of terrorists acquiring a nuclear weapon. The significance of the Global Initiative against Nuclear Terrorism lies not only in its substance but in Russia's visible joint ownership of the Initiative. After years in which Washington lectured Moscow about this threat, Putin's joint leadership in securing nuclear material worldwide should give added impetus to this undertaking inside Russia as well. Globally, this initiative calls for work plans in five arenas: prevention, detection, disruption, mitigation of consequences after an attack, and strengthening domestic laws and export controls against future A.Q. Khans. The guaranteed nuclear fuel supply tightens the noose around Iran as it seeks to exploit a loophole in the global Nuclear Non-Proliferation Treaty. By guaranteeing States that six separate international suppliers will provide backup guarantees against interruption of supply for any reason other that breech of commitments under the NPT, this proposal eliminates Iran's excuse for Natanz-the enrichment plant it is rushing to finish today. This system for supply will be subject to the supervision by the IAEA, which will also have nuclear fuel reserves that allow it to be a supplier of last resort. The Civil Nuclear Agreement will allow joint research on next-generation, proliferation-proof reactors, including technologies where Russian science is the best in the world. It will permit sale to Russia of US technologies that can improve the safety and efficiency of Russian nuclear

  5. 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).

  6. Nuclear reactor fuel element splitter

    International Nuclear Information System (INIS)

    A method and apparatus are disclosed for removing nuclear fuel from a clad fuel element. The fuel element is power driven past laser beams which simultaneously cut the cladding lengthwise into at least two longitudinal pieces. The axially cut lengths of cladding are then separated, causing the nuclear fuel contained therein to drop into a receptacle for later disposition. The cut lengths of cladding comprise nuclear waste which is disposed of in a suitable manner. 6 claims, 10 drawing figures

  7. Sustainability Features of Nuclear Fuel Cycle Options

    Directory of Open Access Journals (Sweden)

    Stefano Passerini

    2012-09-01

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

  8. 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

  9. Convincing about the advanced use of nuclear energy closing the fuel cycle: from a burden to a solution

    International Nuclear Information System (INIS)

    France has associated a closed fuel cycle with its nuclear program, and developed the corresponding treatment recycling capabilities accordingly. This choice was recently consolidated by law. according to the sustainable management of radioactive materials and waste act of June 2006, the volume and radio toxicity reduction of nuclear waste is an objective that can notably be reached with their treatment and conditioning. Presently, used fuel valuable components (U and Pu) are recycled into MOX fuel and RepU, when fission products are conditioned under an extremely solid and resistant form which cannot disperse and dissolve in the environment (High Level Vitrified Waste). Safety and waste minimisation remain the AREVA constant objective. Presently operated treatment and recycling AREVA NC facilities are using mature industrial technologies, which address environment preservation and non proliferation concerns. This french national choice requires a permanent global acceptance strategy towards politicians, media, associations and more generally public opinion: to. be accepted, in needs to be understood. Transparency, dialogue and information are keywords for AREVA NC to be sure that closing the fuel cycle is considered as the best option available now for responsibly managing the waste, respecting the environment, preserving the resource and securing the future. Partnering in this Global Acceptance policy with other countries and customers, who already rely- or plan to do so - on this recycling strategy is both a reality and a permanent axis of development for AREVA NC

  10. Study of advanced professional educational requirements relative to nuclear fuel cycle engineering in industry and government. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jur, T.A.; Huhns, M.N.; Keating, D.A.; Orloff, D.I.; Rhodes, C.A.; Stanford, T.G.; Stephens, L.M.; Tatterson, G.B.; Van Brunt, V.

    1978-12-01

    Under contract with the U.S. Department of Energy, the College of Engineering at the University of South Carolina has conducted an assessment of educational needs among engineers working in nuclear fuel cycle related areas. The study was initiated as a regional effort focusing on the concentration of nuclear industry in the Southeast. Educational needs addressed were those at the post-baccalaureate professional level. The project was envisioned as providing base line information for the eventual implementation of a program in line with the needs of the Southeast's nuclear community. Specific objectives were to establish the content of such a program and to determine those specialized features which would make the program most attractive and useful.

  11. Study of advanced professional educational requirements relative to nuclear fuel cycle engineering in industry and government. Final report

    International Nuclear Information System (INIS)

    Under contract with the U.S. Department of Energy, the College of Engineering at the University of South Carolina has conducted an assessment of educational needs among engineers working in nuclear fuel cycle related areas. The study was initiated as a regional effort focusing on the concentration of nuclear industry in the Southeast. Educational needs addressed were those at the post-baccalaureate professional level. The project was envisioned as providing base line information for the eventual implementation of a program in line with the needs of the Southeast's nuclear community. Specific objectives were to establish the content of such a program and to determine those specialized features which would make the program most attractive and useful

  12. LIGHTBRIDGE corporation advanced metallic fuel

    International Nuclear Information System (INIS)

    Lightbridge Corporation is developing a metallic nuclear fuel which utilizes an innovative fuel rod geometry and composition to improve power plant economics and enhance the performance and safety of commercial light water reactors. The versatile fuel can utilize uranium or plutonium as the fissile component. The fuel is fully compatible with existing light water reactor designs and requires no major changes to reactor operations. The metallic fuel provides a durable solution that is also capable of operating at higher power density than existing fuels allowing for increased power output and cycle length compared to conventional oxide fuels. Lightbridge patented nuclear fuel technologies are designed to significantly enhance nuclear power industry economics and increase power output by: 1) extending fuel cycle length to 24 months or longer while simultaneously increasing power output by 10% or increasing power output by up to 17% with 18-month fuel cycles in existing pressurized water reactors (PWRs); 2) enabling increased reactor power output of up to 30% without changing core size in new build PWRs; and 3) reducing the volume of used fuel per kilowatt-hour as well as enhancing proliferation resistance of spent fuel. (author)

  13. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1972-01-01

    Advances in Nuclear Science and Technology, Volume 6 provides information pertinent to the fundamental aspects of nuclear science and technology. This book covers a variety of topics, including nuclear steam generator, oscillations, fast reactor fuel, gas centrifuge, thermal transport system, and fuel cycle.Organized into six chapters, this volume begins with an overview of the high standards of technical safety for Europe's first nuclear-propelled merchant ship. This text then examines the state of knowledge concerning qualitative results on the behavior of the solutions of the nonlinear poin

  14. Swelling-resistant nuclear fuel

    Science.gov (United States)

    Arsenlis, Athanasios; Satcher, Jr., Joe; Kucheyev, Sergei O.

    2011-12-27

    A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

  15. Nuclear power and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Australian Nuclear Science and Technology Organization maintains an ongoing assessment of the world's nuclear technology developments, as a core activity of its Strategic Plan. This publication reviews the current status of the nuclear power and the nuclear fuel cycle in Australia and around the world. Main issues discussed include: performances and economics of various types of nuclear reactors, uranium resources and requirements, fuel fabrication and technology, radioactive waste management. A brief account of the large international effort to demonstrate the feasibility of fusion power is also given. 11 tabs., ills

  16. Status of advanced carbide fuels: Past, present, and future

    Science.gov (United States)

    Anghaie, Samim; Knight, Travis

    2002-01-01

    Solid solution, mixed uranium/refractory metal carbide fuels such as (U, Zr, Nb)C, so called ternary carbide or tri-carbide fuels have great potential for applications in next generation advanced nuclear power reactors. Because of their high melting points, high thermal conductivity, improved resistance to hot hydrogen corrosion, and good fission product retention, these advanced nuclear fuels have great potential for high performance reactors with increased safety margins. Despite these many benefits, some concerns regarding carbide fuels include compatibility issues with coolant and/or cladding materials and their endurance under the extreme conditions associated with nuclear thermal propulsion. The status of these fuels is reviewed to characterize their performance for space nuclear power applications. Results of current investigations are presented and as well as future directions of study for these advanced nuclear fuels. .

  17. Study of advanced professional educational requirements relative to nuclear fuel cycle engineering in industry and government. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jur, T.A.; Huhns, M.N.; Keating, D.A.; Orloff, D.I.; Rhodes, C.A.; Stanford, T.G.; Stephens, L.M.; Tatterson, G.B.; Van Brunt, V.

    1978-12-01

    An assessment was conducted of educational needs among engineers working in nuclear fuel cycle-related areas, focusing on the nuclear industry in the Southeast. Educational needs addressed were those at the post-baccalaureate professional level. As a result of the study, a list of subject areas has been compiled as best representing the current content of an educational program. In addition to identifying subject areas, a set of course descriptions and reference materials has been developed around each subject. Each course description contains information regarding objectives, anticipated audience, and prerequisites and offers a suggested course outline. An initial modest program of implementation is recommended which would continue to concentrate on the Southeast as a target area.

  18. Study of advanced professional educational requirements relative to nuclear fuel cycle engineering in industry and government. Final report

    International Nuclear Information System (INIS)

    An assessment was conducted of educational needs among engineers working in nuclear fuel cycle-related areas, focusing on the nuclear industry in the Southeast. Educational needs addressed were those at the post-baccalaureate professional level. As a result of the study, a list of subject areas has been compiled as best representing the current content of an educational program. In addition to identifying subject areas, a set of course descriptions and reference materials has been developed around each subject. Each course description contains information regarding objectives, anticipated audience, and prerequisites and offers a suggested course outline. An initial modest program of implementation is recommended which would continue to concentrate on the Southeast as a target area

  19. Report on the workshop "Decay spectroscopy at CARIBU: advanced fuel cycle applications, nuclear structure and astrophysics". 14-16 April 2011, Argonne National Laboratory, USA.

    Energy Technology Data Exchange (ETDEWEB)

    Kondev, F.; Carpenter, M.P.; Chowdhury, P.; Clark, J.A.; Lister, C.J.; Nichols, A.L.; Swewryniak, D. (Nuclear Engineering Division); (Univ. of Massachusetts); (Univ. of Surrey)

    2011-10-06

    A workshop on 'Decay Spectroscopy at CARIBU: Advanced Fuel Cycle Applications, Nuclear Structure and Astrophysics' will be held at Argonne National Laboratory on April 14-16, 2011. The aim of the workshop is to discuss opportunities for decay studies at the Californium Rare Isotope Breeder Upgrade (CARIBU) of the ATLAS facility with emphasis on advanced fuel cycle (AFC) applications, nuclear structure and astrophysics research. The workshop will consist of review and contributed talks. Presentations by members of the local groups, outlining the status of relevant in-house projects and availabile equipment, will also be organized. time will also be set aside to discuss and develop working collaborations for future decay studies at CARIBU. Topics of interest include: (1) Decay data of relevance to AFC applications with emphasis on reactor decay heat; (2) Discrete high-resolution gamma-ray spectroscopy following radioactive decya and related topics; (3) Calorimetric studies of neutron-rich fission framgents using Total ABsorption Gamma-Ray Spectrometry (TAGS) technique; (4) Beta-delayed neutron emissions and related topics; and (5) Decay data needs for nuclear astrophysics.

  20. 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.

  1. Advanced nuclear reactor systems - an Indian perspective

    International Nuclear Information System (INIS)

    The Indian nuclear power programme envisages use of closed nuclear fuel cycle and thorium utilisation as its mainstay for its sustainable growth. The current levels of deployment of nuclear energy in India need to be multiplied nearly hundred fold to reach levels of electricity generation that would facilitate the country to achieve energy independence as well as a developed status. The Indian thorium based nuclear energy systems are being developed to achieve sustainability in respect of fuel resource along with enhanced safety and reduced waste generation. Advanced Heavy Water Reactor and its variants have been designed to meet these objectives. The Indian High Temperature Reactor programme also envisages use of thorium-based fuel with advanced levels of passive safety features. (author)

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

    International Nuclear Information System (INIS)

    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)

  3. The ANF [Advanced Nuclear Fuels Corporation]-RELAP small-break LOCA [loss-of-coolant accident] analysis for the Comanche Peak steam electric station

    International Nuclear Information System (INIS)

    The system response code RELAP/MOD2 Idaho National Engineering Laboratory cycle 36.02, with modifications developed by Advanced Nuclear Fuels Corporation (ANF), was used to perform small-break loss-of-coolant accident (SBLOCA) calculations for the Comanche Peak steam electric station (CPSES) unit 1. The ability of the ANF-RELAP code to calculate the SBLOCA system response for the four-loop pressurized water reactor is presented by discussing the overall system response, the system mass distribution, and the core response

  4. Thermochemical modelling of advanced CANDU reactor fuel

    Science.gov (United States)

    Corcoran, Emily Catherine

    2009-04-01

    With an aging fleet of nuclear generating facilities, the imperative to limit the use of non-renewal fossil fuels and the inevitable need for additional electricity to power Canada's economy, a renaissance in the use of nuclear technology in Canada is at hand. The experience and knowledge of over 40 years of CANDU research, development and operation in Ontario and elsewhere has been applied to a new generation of CANDU, the Advanced CANDU Reactor (ACR). Improved fuel design allows for an extended burnup, which is a significant improvement, enhancing the safety and the economies of the ACR. The use of a Burnable Neutron Absorber (BNA) material and Low Enriched Uranium (LEU) fuel has created a need to understand better these novel materials and fuel types. This thesis documents a work to advance the scientific and technological knowledge of the ACR fuel design with respect to thermodynamic phase stability and fuel oxidation modelling. For the BNA material, a new (BNA) model is created based on the fundamental first principles of Gibbs energy minimization applied to material phase stability. For LEU fuel, the methodology used for the BNA model is applied to the oxidation of irradiated fuel. The pertinent knowledge base for uranium, oxygen and the major fission products is reviewed, updated and integrated to create a model that is applicable to current and future CANDU fuel designs. As part of this thesis, X-Ray Diffraction (XRD) and Coulombic Titration (CT) experiments are compared to the BNA and LEU models, respectively. From the analysis of the CT results, a number of improvements are proposed to enhance the LEU model and provide confidence in its application to ACR fuel. A number of applications for the potential use of these models are proposed and discussed. Keywords: CANDU Fuel, Gibbs Energy Mimimization, Low Enriched Uranium (LEU) Fuel, Burnable Neutron Absorber (BNA) Material, Coulometric Titration, X-Ray Diffraction

  5. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1962-01-01

    Advances in Nuclear Science and Technology, Volume 1 provides an authoritative, complete, coherent, and critical review of the nuclear industry. This book covers a variety of topics, including nuclear power stations, graft polymerization, diffusion in uranium alloys, and conventional power plants.Organized into seven chapters, this volume begins with an overview of the three stages of the operation of a power plant, either nuclear or conventionally fueled. This text then examines the major problems that face the successful development of commercial nuclear power plants. Other chapters consider

  6. ADVANCED NUCLEAR FUEL CYCLE EFFECTS ON THE TREATMENT OF UNCERTAINTY IN THE LONG-TERM ASSESSMENT OF GEOLOGIC DISPOSAL SYSTEMS - EBS INPUT

    Energy Technology Data Exchange (ETDEWEB)

    Sutton, M; Blink, J A; Greenberg, H R; Sharma, M

    2012-04-25

    encapsulated in borosilicate glass. Because the heat load of the glass was much less than the PWR and BWR assemblies, the glass waste form was able to be co-disposed with the open cycle waste, by interspersing glass waste packages among the spent fuel assembly waste packages. In addition, the Yucca Mountain repository was designed to include some research reactor spent fuel and naval reactor spent fuel, within the envelope that was set using the commercial reactor assemblies as the design basis waste form. This milestone report supports Sandia National Laboratory milestone M2FT-12SN0814052, and is intended to be a chapter in that milestone report. The independent technical review of this LLNL milestone was performed at LLNL and is documented in the electronic Information Management (IM) system at LLNL. The objective of this work is to investigate what aspects of quantifying, characterizing, and representing the uncertainty associated with the engineered barrier are affected by implementing different advanced nuclear fuel cycles (e.g., partitioning and transmutation scenarios) together with corresponding designs and thermal constraints.

  7. Nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    The present invention concerns an improvement for corrosion resistance of the welded portion of materials which constitutes a reprocessing plant of spent nuclear fuels. That is, Mo-added austenite stainless steel is used for a plant member at the portion in contact with a nitric acid solution. Then, laser beams are irradiated to the welded portion of the plant member and the surface layer is heated to higher than 1,000degC. If such a heat treatment is applied, the degradation of corrosion resistance of the welded portion can be eliminated at the surface. Further, since laser beams are utilized, heating can be limited only to the surface. Accordingly, undesired thermal deformation of the plant members can be prevented. As a result, the plant member having high pit corrosion resistance against a dissolution solution for spent fuels containing sludges comprising insoluble residue and having resistance to nitric acid solution also in the welded portion substantially equal to that of the matrix can be attained. (I.S.)

  8. Discovery and design of nuclear fuels

    Directory of Open Access Journals (Sweden)

    Marius Stan

    2009-11-01

    Full Text Available To facilitate the discovery and design of innovative nuclear fuels, multi-scale models and simulations are used to predict irradiation effects on properties such as thermal conductivity, oxygen diffusivity, and thermal expansion. The multi-scale approach is illustrated using results on ceramic fuels, with a focus on predictions of point defect concentration, stoichiometry, and phase stability. The high performance computer simulations include coupled heat transport, diffusion, and thermal expansion, and gas bubble formation and evolution in a fuel element consisting of UO2 fuel and metallic cladding. The second part of the paper is dedicated to a discussion of an international strategy for developing advanced, innovative nuclear fuels. Four initiatives are proposed to accelerate the discovery and design of new materials: (a Create Institutes for Materials Discovery and Design, (b Create an International Knowledgebase for experimental data, models (mathematical expressions, and simulations (codes, (c Improve education and (d Set up international collaborations.

  9. Recent advances in fuel product and manufacturing process development

    International Nuclear Information System (INIS)

    This paper discusses advancements in commercial nuclear fuel products and manufacturing made by the Westinghouse Electric Corporation in response to the commercial nuclear fuel industry's demand for high reliability, increased plant availability and improved operating flexibility. The features and benefits of Westinghouse's most advanced fuel products--VANTAGE 5 for PWR plants and QUAD+ for BWR plants--are described, as well as high performance fuel concepts now under development for delivery in the late 1980s. The paper also discusses the importance of in-process quality control throughout manufacturing towards reducing product variability and improving fuel reliability

  10. Recent advances in fuel product and manufacturing process development

    International Nuclear Information System (INIS)

    This paper discusses advancements in commercial nuclear fuel products and manufacturing made by the Westinghouse Electric Corporation in response to the commercial nuclear fuel industry's demand for high reliability, increased plant availability and improved operating flexibility. The features and benefits of Westinghouse's most advanced fuel products--VANTAGE 5 for PWR plants and QUAD+ for BWR plants--are described, as well as 'high performance' fuel concepts now under development for delivery in the late 1980s. The paper also disusses the importance of in-process quality control throughout manufacturing towards reducing product variability and improving fuel reliability. (author)

  11. ARPA advanced fuel cell development

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, L.H.

    1995-08-01

    Fuel cell technology is currently being developed at the Advanced Research Projects Agency (ARPA) for several Department of Defense applications where its inherent advantages such as environmental compatibility, high efficiency, and low noise and vibration are overwhelmingly important. These applications range from man-portable power systems of only a few watts output (e.g., for microclimate cooling and as direct battery replacements) to multimegawatt fixed base systems. The ultimate goal of the ARPA program is to develop an efficient, low-temperature fuel cell power system that operates directly on a military logistics fuel (e.g., DF-2 or JP-8). The absence of a fuel reformer will reduce the size, weight, cost, and complexity of such a unit as well as increase its reliability. In order to reach this goal, ARPA is taking a two-fold, intermediate time-frame approach to: (1) develop a viable, low-temperature proton exchange membrane (PEM) fuel cell that operates directly on a simple hydrocarbon fuel (e.g., methanol or trimethoxymethane) and (2) demonstrate a thermally integrated fuel processor/fuel cell power system operating on a military logistics fuel. This latter program involves solid oxide (SOFC), molten carbonate (MCFC), and phosphoric acid (PAFC) fuel cell technologies and concentrates on the development of efficient fuel processors, impurity scrubbers, and systems integration. A complementary program to develop high performance, light weight H{sub 2}/air PEM and SOFC fuel cell stacks is also underway. Several recent successes of these programs will be highlighted.

  12. Design concepts and advanced telerobotics development for facilities in the back end of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    In the Fuel Recycle Division 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. Five areas of the development effort are primary contributors to the goal of higher operating efficiency for major facilities for the back end of the nuclear fuel cycle. These areas are the single-cell concept, the low-flow ventilation concept, television viewing, equipment-mounting racks, and force-reflecting manipulation. These somewhat innovative directions are products of a design process where the technical scenario to be accomplished, the remote equipment to accomplish the scenario, and the facility design to house the equipment, are considered in an iterative design process to optimize performance, maximize long-term costs effectiveness, and minimize initial capital outlay. 14 refs., 3 figs

  13. Design concepts and advanced telerobotics development for facilities in the back end of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    In the Fuel Recycle Division at the Oak Ridge National Laboratory (ORNL), 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. Five areas of the development effort are primary contributors to the goal of higher operating efficiency for major facilities for the back end of the nuclear fuel cycle. These areas are (1) the single-cell concept, (2) the low-flow ventilation concept, (3) television viewing, (4) equipment-mounting racks, and (5) force-reflecting manipulation. These somewhat innovative directions are products of a design process where the technical scenario to be accomplished, the remote equipment to accomplish the scenario, and the facility design to house the equipment, are considered in an iterative design process to optimize performance, maximize long-term costs effectiveness, and minimize initial capital outlay. (author)

  14. Development of computational technology on heat transfer and fluid flow in a nuclear fuel bundle of advanced reactor

    International Nuclear Information System (INIS)

    The assessment of the RANS(Reynolds-Averaged Navier-Stokes) based turbulence model was conducted to establish the optimal CFD system for turbulent flow and heat transfer in reactor during the first year of the project. The RANS models used in this project are the two-equation models based on the eddy viscosity assumption and the Second-Moment Closure(SMC) models. Since the nuclear fuel assembly loaded in the nuclear reactor is a rod bundle which is square or triangular array, the predictions using the various turbulence models were compared for turbulent flow in bare square and/or triangular rod bundle and the rod bundle with the flow mixing vane. The study for the second year of the project examined the CFD model and the applicability of the CFD code for the turbulent two-phase flow. The numerical predictions of lateral distributions of void fraction, phasic velocities and turbulent kinetic energy were compared against the experimental results for upward and downward bubbly flow in a vertical tube. The boiling flows in vertical tube and rod bundle were also simulated to verify the CFD results

  15. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

  16. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U3O8 were replaced by U3Si2-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is to

  17. Dynamic Systems Analysis Report for Nuclear Fuel Recycle

    Energy Technology Data Exchange (ETDEWEB)

    Brent Dixon; Sonny Kim; David Shropshire; Steven Piet; Gretchen Matthern; Bill Halsey

    2008-12-01

    This report examines the time-dependent dynamics of transitioning from the current United States (U.S.) nuclear fuel cycle where used nuclear fuel is disposed in a repository to a closed fuel cycle where the used fuel is recycled and only fission products and waste are disposed. The report is intended to help inform policy developers, decision makers, and program managers of system-level options and constraints as they guide the formulation and implementation of advanced fuel cycle development and demonstration efforts and move toward deployment of nuclear fuel recycling infrastructure.

  18. Advanced PWR fuel design concepts

    International Nuclear Information System (INIS)

    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)

  19. Application of Spatial Data Modeling Systems, Geographical Information Systems (GIS), and Transportation Routing Optimization Methods for Evaluating Integrated Deployment of Interim Spent Fuel Storage Installations and Advanced Nuclear Plants

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Gary T [ORNL; Belles, Randy [ORNL; Cetiner, Sacit M [ORNL; Howard, Rob L [ORNL; Liu, Cheng [ORNL; Mueller, Don [ORNL; Omitaomu, Olufemi A [ORNL; Peterson, Steven K [ORNL; Scaglione, John M [ORNL

    2012-06-01

    The objective of this siting study work is to support DOE in evaluating integrated advanced nuclear plant and ISFSI deployment options in the future. This study looks at several nuclear power plant growth scenarios that consider the locations of existing and planned commercial nuclear power plants integrated with the establishment of consolidated interim spent fuel storage installations (ISFSIs). This research project is aimed at providing methodologies, information, and insights that inform the process for determining and optimizing candidate areas for new advanced nuclear power generation plants and consolidated ISFSIs to meet projected US electric power demands for the future.

  20. Advanced Measuring (Instrumentation Methods for Nuclear Installations: A Review

    Directory of Open Access Journals (Sweden)

    Wang Qiu-kuan

    2012-01-01

    Full Text Available The nuclear technology has been widely used in the world. The research of measurement in nuclear installations involves many aspects, such as nuclear reactors, nuclear fuel cycle, safety and security, nuclear accident, after action, analysis, and environmental applications. In last decades, many advanced measuring devices and techniques have been widely applied in nuclear installations. This paper mainly introduces the development of the measuring (instrumentation methods for nuclear installations and the applications of these instruments and methods.

  1. LOFT nuclear fuel rod behavior

    International Nuclear Information System (INIS)

    An overview of the calculational models used to predict fuel rod response for Loss-of-Fluid Test (LOFT) data from the first LOFT nuclear test is presented and discussed and a comparison of predictions with experimental data is made

  2. Intelligent Automated Nuclear Fuel Pellet Inspection System

    International Nuclear Information System (INIS)

    At the present time, nuclear pellet inspection is performed manually using naked eyes for judgment and decisionmaking on accepting or rejecting pellets. This current practice of pellet inspection is tedious and subject to inconsistencies and error. Furthermore, unnecessary re-fabrication of pellets is costly and the presence of low quality pellets in a fuel assembly is unacceptable. To improve the quality control in nuclear fuel fabrication plants, an automated pellet inspection system based on advanced techniques is needed. Such a system addresses the following concerns of the current manual inspection method: (1) the reliability of inspection due to typical human errors, (2) radiation exposure to the workers, and (3) speed of inspection and its economical impact. The goal of this research is to develop an automated nuclear fuel pellet inspection system which is based on pellet video (photographic) images and uses artificial intelligence techniques

  3. Innovative Nuclear Fuels: Results and Strategy

    International Nuclear Information System (INIS)

    Materials discovery involves exploring and identifying existing (natural) materials with desirable properties and functionality. Materials design aims at creating new (artificial) materials with predefined properties and functionality. Nuclear fuels are often developed using both methods, with a certain advantage given to discovery. To facilitate the discovery and design of innovative nuclear fuels, multi-scale models and simulations are used to predict irradiation effects on the thermal conductivity, oxygen diffusivity, and thermal expansion of oxide fuels. The scientific method used in this approach covers a large spectrum of time and space scales, from electronic structure to atomistic levels, through meso-scale and all the way to continuum phenomena. The multi-scale approach is illustrated using results on UO2/PuO2 fuels with a focus on predictions of point defect concentrations, stoichiometry, and phase stability. The high performance computer simulations include coupled heat transport, diffusion, and thermal expansion, gas bubble formation and temperature evolution in a fuel element consisting of UO2 fuel and metallic cladding. Uncertainty evaluation reveals that ignoring the composition dependence of fuel properties in the simulations can lead to large errors (>100 k) in the calculations of the centerline temperature. The second part of the talk is dedicated to a discussion of an international strategy for developing advanced, innovative nuclear fuels. It starts with a brief review of the international status of nuclear fuels research, including results from American, European, and Japanese national laboratories and universities. In an effort to improve collaborative work, the status of thermo-chemical databases is used as an example of outstanding opportunities and exciting scientific programs that require better synchronization to advance the research and to avoid excessive redundancy. The presentation ends with a discussion of existing and emerging

  4. 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

  5. Waste Stream Analyses for Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    N. R. Soelberg

    2010-08-01

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

  6. Abundant thorium as an alternative nuclear fuel

    International Nuclear Information System (INIS)

    It has long been known that thorium-232 is a fertile radioactive material that can produce energy in nuclear reactors for conversion to electricity. Thorium-232 is well suited to a variety of reactor types including molten fluoride salt designs, heavy water CANDU configurations, and helium-cooled TRISO-fueled systems. Among contentious commercial nuclear power issues are the questions of what to do with long-lived radioactive waste and how to minimize weapon proliferation dangers. The substitution of thorium for uranium as fuel in nuclear reactors has significant potential for minimizing both problems. Thorium is three times more abundant in nature than uranium. Whereas uranium has to be imported, there is enough thorium in the United States alone to provide adequate grid power for many centuries. A well-designed thorium reactor could produce electricity less expensively than a next-generation coal-fired plant or a current-generation uranium-fueled nuclear reactor. Importantly, thorium reactors produce substantially less long-lived radioactive waste than uranium reactors. Thorium-fueled reactors with molten salt configurations and very high temperature thorium-based TRISO-fueled reactors are both recommended for priority Generation IV funding in the 2030 time frame. - Highlights: • Thorium is an abundant nuclear fuel that is well suited to three advanced reactor configurations. • Important thorium reactor configurations include molten salt, CANDU, and TRISO systems. • Thorium has important nuclear waste disposal advantages relative to pressurized water reactors. • Thorium as a nuclear fuel has important advantages relative to weapon non-proliferation

  7. Advanced fuel chemistry for advanced engines.

    Energy Technology Data Exchange (ETDEWEB)

    Taatjes, Craig A.; Jusinski, Leonard E.; Zador, Judit; Fernandes, Ravi X.; Miller, James A.

    2009-09-01

    Autoignition chemistry is central to predictive modeling of many advanced engine designs that combine high efficiency and low inherent pollutant emissions. This chemistry, and especially its pressure dependence, is poorly known for fuels derived from heavy petroleum and for biofuels, both of which are becoming increasingly prominent in the nation's fuel stream. We have investigated the pressure dependence of key ignition reactions for a series of molecules representative of non-traditional and alternative fuels. These investigations combined experimental characterization of hydroxyl radical production in well-controlled photolytically initiated oxidation and a hybrid modeling strategy that linked detailed quantum chemistry and computational kinetics of critical reactions with rate-equation models of the global chemical system. Comprehensive mechanisms for autoignition generally ignore the pressure dependence of branching fractions in the important alkyl + O{sub 2} reaction systems; however we have demonstrated that pressure-dependent 'formally direct' pathways persist at in-cylinder pressures.

  8. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Center for Nuclear Engineering has shown expertise in the field of nuclear and energy systems ad correlated areas. Due to the experience obtained over decades in research and technological development at Brazilian Nuclear Program personnel has been trained and started to actively participate in the design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in the production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. The Nuclear Fuel Center is responsible for the production of the nuclear fuel necessary for the continuous operation of the IEA-R1 research reactor. Development of new fuel technologies is also a permanent concern

  9. Alternatives for nuclear fuel disposal

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Badillo A, V.; Palacios H, J.; Celis del Angel, L., E-mail: ramon.ramirez@inin.gob.m [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments in the construction of repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution? or, What is the best technology for a specific solution? Many countries have deferred the decision on selecting an option, while other works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However, currently is under process an extended power up rate to 20% of their original power and also there are plans to extend operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. So this work describes some different alternatives that have been studied in Mexico to define which will be the best alternative to follow. (Author)

  10. The development of nuclear fuel for the future

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong Seong; Kim, Shi Hwan; Lim, Kap Soon; Yang, Myeong Soon; Lee, Yeong Uh; Kim, Yeong Jin; Ju, Hyeong Kook; Oh, Soo Yeol; Kim, Yeong Il; No, Jae Man; Lee, Jae Kyeong; Koo, Yang Heon; Song, Keun Uh; Lee, Hee Seong; Kim, Keung Koo; Jeong, Hyeong Kook; Hwang, Dae Heon; Yoo, Yeon Jong; Jang, Jong Hwa; Kim, Jeong Do; Kil, Chung Seop; Choi, Chang Beom; Bae, Ki Kwang; Kim, Han Soo; Choi, Meong Seon; Kim, Hyeong Seop; Lee, Jeong Won; Park, Chun Ho; Jeong, Sang Tae; Park, Jin Ho; Kim, Eung Ho; Kim, Tae Jun; Jeong, Keong Chai; Uh, Moon Sik; Hong, Soon Bok; Kim, Yeon Koo [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1993-05-01

    Goal is to develop nuclear fuels which can improve the safety and economics of nuclear power generation and enhance the effective use of uranium resources. The fuels are aimed for successive power plants after the year 2000. To achieve these goals, it is necessary to analyse characteristics of advanced concept fuels based on currently available fuel design and manufacturing technology and to establish new design and manufacturing technology for the new concept fuels. This first year of the project focused on 1) technical feasibility evaluation of application of existing technology to advanced concept fuel such as extra-high BU fuel or other improved feature fuel, 2) analysis of domestic and foreign circumstances for advanced concept fuel development and drawing-out of plans for its development, and 3) securing of basic technology for development. For analysis of nuclear fuel characteristics, expected change in design parameters at high BU has been evaluated and fuel cycle evaluation models were improved for development of fuel cycle strategy with high BU fuel. For verification and improvement of design methodology, technical feasibility of application of the existing nuclear design system to advanced fuel has been analysed and thermal-hydraulic analysis methodology is under improvement for more thermal design margin. For review of state-of-the-art and study of developmental plans, trends and circumstances of nuclear fuel development in foreign countries have been analysed and several domestic developmental plans were identified. For development of advanced nuclear fuel fabrication technology, preliminary experiments for mixing of additives with UO{sub 2} powder were performed and state-of-the-art of new fabrication techniques which are under development in recent years were analysed...

  11. Advanced Multiphysics Modeling of Fast Reactor Fuel Behavior

    International Nuclear Information System (INIS)

    Evaluation of fast reactor fuel thermo-mechanical performance using fuel performance codes is a key aspect of advanced fast reactors designs. Those fuel performance codes capture the multiphysics nature of fuel behavior during irradiation where different, mostly interdependent, phenomena are taking place. Existing fuel performance codes do not fully capture those interdependencies and present the different phenomena through de-coupled models. Recent developments in multiphysics simulation capabilities and availability of advanced computing platforms led to advancements in simulation of nuclear fuel behavior. This paper presents current experiences in applying different multiphysics simulation platforms to evaluation of fast reactors metallic fuel behavior. Full 3D finite element simulation platforms that include capabilities to fully couple key fuel behavior models are discussed. Issues associated with coupling metallic fuels phenomena, such as fission gas models and constituent distribution models, with thermo-mechanical finite element platforms, as well as different coupling schemes are also discussed. (author)

  12. Nuclear fuel cycle. V. 2

    International Nuclear Information System (INIS)

    Nuclear fuel cycle information in some countries that develop, supply or use nuclear energy is presented. Data about Argentina, Australia, Belgium, Netherlands, Italy, Denmarmark, Norway, Sweden, Switzerland, Finland, Spain and India are included. The information is presented in a tree-like graphic way. (C.S.A.)

  13. Nuclear fuel cycle. V. 1

    International Nuclear Information System (INIS)

    Nuclear fuel cycle information in the main countries that develop, supply or use nuclear energy is presented. Data about Japan, FRG, United Kingdom, France and Canada are included. The information is presented in a tree-like graphic way. (C.S.A.)

  14. Nuclear fuel procurement management at nuclear power plant

    International Nuclear Information System (INIS)

    The market situation of nuclear fuel cycles is highlighted. It also summarises the possible contract models and the elements of effective management for nuclear fuel procurement at nuclear power station based upon the nuclear fuel procurement practice of Guangdong Daya Bay Nuclear Power Station (GNPS)

  15. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    The grid-shaped spacer for PWR fuel elements consists of flat, upright metal bars at right angles to the fuel rods. In one corner of a grid mesh it has a spring with two end parts for the fuel rod. The cut-outs for the end parts start from an end edge of the metal bar parallel to the fuel rods. The transverse metal bar is one of four outer metal bars. Both end parts of the spring have an extension parallel to this outer metal arm, which grips a grid mesh adjacent to this grid mesh at the side in one corner of the spacer and forms an end part of a spring for the fuel rod there on the inside of the outer metal bar. (HP)

  16. Nuclear fuels accounting interface: River Bend experience

    Energy Technology Data Exchange (ETDEWEB)

    Barry, J.E.

    1986-01-01

    This presentation describes nuclear fuel accounting activities from the perspective of nuclear fuels management and its interfaces. Generally, Nuclear Fuels-River Bend Nuclear Group (RBNG) is involved on a day-by-day basis with nuclear fuel materials accounting in carrying out is procurement, contract administration, processing, and inventory management duties, including those associated with its special nuclear materials (SNM)-isotopics accountability oversight responsibilities as the Central Accountability Office for the River Bend Station. As much as possible, these duties are carried out in an integrated, interdependent manner. From these primary functions devolve Nuclear Fuels interfacing activities with fuel cost and tax accounting. Noting that nuclear fuel tax accounting support is of both an esoteric and intermittent nature, Nuclear Fuels-RBNG support of developments and applications associated with nuclear fuel cost accounting is stressed in this presentation.

  17. Nuclear fuels accounting interface: River Bend experience

    International Nuclear Information System (INIS)

    This presentation describes nuclear fuel accounting activities from the perspective of nuclear fuels management and its interfaces. Generally, Nuclear Fuels-River Bend Nuclear Group (RBNG) is involved on a day-by-day basis with nuclear fuel materials accounting in carrying out is procurement, contract administration, processing, and inventory management duties, including those associated with its special nuclear materials (SNM)-isotopics accountability oversight responsibilities as the Central Accountability Office for the River Bend Station. As much as possible, these duties are carried out in an integrated, interdependent manner. From these primary functions devolve Nuclear Fuels interfacing activities with fuel cost and tax accounting. Noting that nuclear fuel tax accounting support is of both an esoteric and intermittent nature, Nuclear Fuels-RBNG support of developments and applications associated with nuclear fuel cost accounting is stressed in this presentation

  18. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1973-01-01

    Advances in Nuclear Science and Technology, Volume 7 provides information pertinent to the fundamental aspects of nuclear science and technology. This book discusses the safe and beneficial development of land-based nuclear power plants.Organized into five chapters, this volume begins with an overview of irradiation-induced void swelling in austenitic stainless steels. This text then examines the importance of various transport processes for fission product redistribution, which depends on the diffusion data, the vaporization properties, and the solubility in the fuel matrix. Other chapters co

  19. Advanced nuclear precleaner

    Energy Technology Data Exchange (ETDEWEB)

    Wright, S.R. [InnovaTech, Inc., Durham, NC (United States)

    1997-10-01

    This Phase II Small Business Innovation Research (SBIR) program`s goal is to develop a dynamic, self-cleaning air precleaner for high-efficiency particulate air (HEPA) filtration systems that would extend significantly the life of HEPA filter banks by reducing the particulate matter that causes filter fouling and increased pack pressure. HEPA filters are widely used in DOE, Department of Defense, and a variety of commercial facilities. InnovaTech, Inc. (Formerly Micro Composite materials Corporation) has developed a proprietary dynamic separation device using a concept called Boundary Layer Momentum Transfer (BLMT) to extract particulate matter from fluid process streams. When used as a prefilter in the HVAC systems or downstream of waste vitrifiers in nuclear power plants, fuel processing facilities, and weapons decommissioning factories, the BLMT filter will dramatically extend the service life and increase the operation efficiency of existing HEPA filtration systems. The BLMT filter is self cleaning, so there will be no degraded flow or increased pressure drop. Because the BLMT filtration process is independent of temperature, it can be designed to work in ambient, medium, or high-temperature applications. During Phase II, the authors are continuing development of the computerized flow simulation model to include turbulence and incorporate expansion into a three-dimensional model that includes airflow behavior inside the filter housing before entering the active BLMT device. A full-scale (1000 ACFM) prototype filter is being designed to meet existing HEPA filter standards and will be fabricated for subsequent testing. Extensive in-house testing will be performed to determine a full range of performance characteristics. Final testing and evaluation of the prototype filter will be conducted at a DOE Quality Assurance Filter Test Station.

  20. Application of life-cycle information for advancement in safety of nuclear fuel cycle facilities. Application of safety information to advanced safety management support system

    International Nuclear Information System (INIS)

    Risk management is major concern to nuclear energy reprocessing plants to improve plant and process reliability and ensure their safety. This is because we are required to predict potential risks before any accident or disaster occurs. The advancement of safety design and safety systems technologies showed large amount of useful safety-related knowledge that can be of great importance to plant operation to reduce operation risks and ensure safety. This research proposes safety knowledge modeling framework on the basis of ontology technologies to systematically construct plant knowledge model, which includes plant structure, operation, and the associated behaviors. In such plant knowledge model safety related information is defined and linked to the different elements of plant knowledge model. Ontology editor is employed to define the basic concepts and their inter-relations, which are used to capture and construct plant safety knowledge. In order to provide detailed safety knowledgebase, HAZOP results are analyzed and structured so that safety-related knowledge are identified and structured within the plant knowledgebase. The target safety knowledgebase includes: failures, deviations, causes, consequences, and fault propagation as mapped to plant knowledge. The proposed ontology-based safety framework is applied on case study nuclear plant to structure failures, causes, consequences, and fault propagation, which are used to support plant operation. (author)

  1. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    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

  2. Nondestructive measurements on spent fuel for the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Nondestructive measurements on spent fuel are being developed to meet safeguards and materials managment requirements at nuclear facilities. Spent-fuel measurement technology and its applications are reviewed

  3. Dry spent nuclear fuel transfer

    International Nuclear Information System (INIS)

    Newport News Shipbuilding, (NNS), has been transferring spent nuclear fuel in a dry condition for over 25 years. It is because of this successful experience that NNS decided to venture into the design, construction and operation of a commercial dry fuel transfer project. NNS is developing a remote handling system for the dry transfer of spent nuclear fuel. The dry fuel transfer system is applicable to spent fuel pool-to-cask or cask-to-cask or both operations. It is designed to be compatible with existing storage cask technology as well as the developing multi-purpose canister design. The basis of NNS' design is simple. It must be capable of transferring all fuel designs, it must be capable of servicing 100 percent of the commercial nuclear plants, it must protect the public and nuclear operators, it must be operated cost efficiently and it must be transportable. Considering the basic design parameters, the following are more specific requirements included in the design: (a) Total weight of transfer cask less than 24 tons; (b) no requirement for permanent site modifications to support system utilization; (c) minimal radiation dose to operating personnel; (d) minimal generation of radioactive waste; (e) adaptability to any size and length fuel or cask; (f) portability of system allowing its efficient movement from site to site; (g) safe system; all possible ''off normal'' situations are being considered, and resultant safety systems are being engineered into NNS' design to mitigate problems. The primary focus of this presentation is to provide an overview of NNS' Dry Spent Nuclear Fuel Transfer System. (author). 5 refs

  4. Review of the IAEA nuclear fuel cycle and material section activities connected with nuclear fuel including WWER fuel

    International Nuclear Information System (INIS)

    Program activities on Nuclear Fuel Cycle and Materials cover the areas of: 1) raw materials (B.1.01); 2) fuel performance and technology (B.1.02); 3) pent fuel (B.1.03); 4) fuel cycle issues and information system (B.1.04); 5) support to technical cooperation activities (B.1.05). The IAEA activities in fuel performance and technology in 2001 include organization of the fuel experts meetings and completion of the Co-ordinate Research Projects (CRP). The special attention is given to the advanced post-irradiation examination techniques for water reactor fuel and fuel behavior under transients and LOCA conditions. An international research program on modeling of activity transfer in primary circuit of NPP is finalized in 2001. A new CRP on fuel modeling at extended burnup (FUMEX II) has planed to be carried out during the period 2002-2006. In the area of spent fuel management the implementation of burnup credit (BUC) in spent fuel management systems has motivated to be used in criticality safety applications, based on economic consideration. An overview of spent fuel storage policy accounting new fuel features as higher enrichment and final burnup, usage of MOX fuel and prolongation of the term of spent fuel storage is also given

  5. Remote maintenance in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Remote maintenance techniques applied in large-scale nuclear fuel reprocessing plants are reviewed with particular attention to the three major maintenance philosophy groupings: contact, remote crane canyon, and remote/contact. Examples are given, and the relative success of each type is discussed. Probable future directions for large-scale reprocessing plant maintenance are described along with advanced manipulation systems for application in the plants. The remote maintenance development program within the Consolidated Fuel Reprocessing Program at the Oak Ridge National Laboratory is also described. 19 refs., 19 figs

  6. Proliferation Resistant Nuclear Reactor Fuel

    International Nuclear Information System (INIS)

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

  7. Spent-fuel-storage studies at the Barnwell Nuclear Fuel Plant. Studies and research concerning BNFP

    International Nuclear Information System (INIS)

    This report contains the results of various studies and demonstrations related to advanced spent-fuel-storage techniques which were performed at the Barnwell Nuclear Fuel Plant (BNFP) in 1982. The demonstrations evaluated various technical aspects of fuel disassembly and canning and dry-storage techniques. The supporting studies examined thermal limitations and criticality concerns

  8. Evaluation of the Use of Synroc to Solidify the Cesium and Strontium Separations Product from Advanced Aqueous Reprocessing of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Julia Tripp; Vince Maio

    2006-03-01

    This report is a literature evaluation on the Synroc process for determining the potential for application to solidification of the Cs/Sr strip product from advanced aqueous fuel separations activities.

  9. Nuclear fuel cycle studies

    International Nuclear Information System (INIS)

    For the metal-matrix encapsulation of radioactive waste, brittle-fracture, leach-rate, and migration studies are being conducted. For fuel reprocessing, annular and centrifugal contactors are being tested and modeled. For the LWBR proof-of-breeding project, the full-scale shear and the prototype dissolver were procured and tested. 5 figures

  10. Contracting for nuclear fuels

    International Nuclear Information System (INIS)

    This paper deals with uranium sales contracts, i.e. with contractual arrangements in the first steps of the fuel cycle, which cover uranium production and conversion. The various types of contract are described and, where appropriate, their underlying business philosophy and their main terms and conditions. Finally, the specific common features of such contracts are reviewed. (NEA)

  11. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Purpose: To enable a tight seal in fuel rods while keeping the sealing gas pressure at an exact predetermined pressure in fuel rods. Constitution: A vent aperture and a valve are provided to the upper end plug of a cladding tube. At first, the valve is opened to fill gas at a predetermined pressure in the fuel can. Then, a conical valve body is closely fitted to a valve seat by the rotation of a needle valve to eliminate the gap in the engaging thread portion and close the vent aperture. After conducting the reduced pressure test for the fuel rod in a water tank, welding joints are formed between the valve and the end plug through welding to completely seal the cladding tube. Since the welding is conducted after the can has been closed by the valve, the predetermined gas pressure can be maintained at an exact level with no efforts from welding heat and with effective gas leak prevention by the double sealing. (Kawakami, Y.)

  12. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  13. Advanced Thermally Stable Jet Fuels

    Energy Technology Data Exchange (ETDEWEB)

    A. Boehman; C. Song; H. H. Schobert; M. M. Coleman; P. G. Hatcher; S. Eser

    1998-01-01

    The Penn State program in advanced thermally stable jet fuels has five components: 1) development of mechanisms of degradation and solids formation; 2) quantitative measurement of growth of sub-micrometer and micrometer-sized particles 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; and 5) assessment of the potential of producing high yields of cycloalkanes and hydroaromatics from coal.

  14. On the International Nuclear Fuel Cycle Evaluation

    International Nuclear Information System (INIS)

    The president of U.S.A. proposed to various countries in his new policy on atomic energy to reevaluate nuclear fuel cycle internationally from the viewpoint of the prevention of nuclear proliferation. It was decided at the summit meeting of seven advanced countries in London from May 7 to 9, 1977, to start the INFCE taking the necessity of promoting atomic energy development and the importance of reducing the danger of nuclear proliferation as the objects. The preliminary conference was held in Paris in June and July, 1977, and the general meeting to establish the INFCE was held in Washington from October 19 to 21, 1977. 40 countries and 4 international organizations took part, and the plan of works to be completed in 2 years thereafter was decided. 8 working groups were set up to carry out the works. The response to these development and the basic concept of Japan are described. Japan was assigned to the chairman country of the 4th working group concerning fuel reprocessing, handling of plutonium and recycle. The state of activities of respective working groups, the intermediate general meeting held from November 27 to 29, 1978, and the technical coordinating committee is reported. As the post-INFCE problems, the concepts of International Plutonium Storage and International Spent Fuel Management and the guarantee system for nuclear fuel supply are discussed. (Kako, I.)

  15. Spent nuclear fuel reprocessing modeling

    International Nuclear Information System (INIS)

    The long-term wide development of nuclear power requires new approaches towards the realization of nuclear fuel cycle, namely, closed nuclear fuel cycle (CNFC) with respect to fission materials. Plant nuclear fuel cycle (PNFC), which is in fact the reprocessing of spent nuclear fuel unloaded from the reactor and the production of new nuclear fuel (NF) at the same place together with reactor plant, can be one variant of CNFC. Developing and projecting of PNFC is a complicated high-technology innovative process that requires modern information support. One of the components of this information support is developed by the authors. This component is the programme conducting calculations for various variants of process flow sheets for reprocessing SNF and production of NF. Central in this programme is the blocks library, where the blocks contain mathematical description of separate processes and operations. The calculating programme itself has such a structure that one can configure the complex of blocks and correlations between blocks, appropriate for any given flow sheet. For the ready sequence of operations balance calculations are made of all flows, i.e. expenses, element and substance makeup, heat emission and radiation rate are determined. The programme is open and the block library can be updated. This means that more complicated and detailed models of technological processes will be added to the library basing on the results of testing processes using real equipment, in test operating mode. The development of the model for the realization of technical-economic analysis of various variants of technologic PNFC schemes and the organization of 'operator's advisor' is expected. (authors)

  16. Advanced fuel technology - A UK perspective

    International Nuclear Information System (INIS)

    The nuclear power industry in the United Kingdom is perhaps more diverse than in any other country. The diversity in design of stations is matched by a diversity in operating responsibility. The SGHWR and PFR are operated by the United Kingdom Atomic Energy Authority (UKAEA), 2 of the Magnox stations are owned and run by BNFL, 2 of the AGR stations and 1 Magnox station are controlled by the South of Scotland Electricity Board (SSEB), and the remaining reactors (including the Sizewell 'B' PWR) currently come under the responsibility of the Central Electricity Generating Board (CEGB) but will pass into the control of a new state-run company when the rest of the CEGB is privatized in 1990. Against this background of a variety of designs and operational responsibilities, there is clearly a great deal of scope for advances in fuel and fuel component technology. It should be noted, however, that the nuclear energy policy within the United Kingdom, particularly with regard to PWR plants, has been to adopt well proven designs wherever possible. Emphasis is therefore directed towards achieving the successful operation of conservative systems, with research and development work on advanced options for future implementation. The following sections give an overview of the areas where advanced designs are either in production or under development for each of the UK reactor systems in turn, together with an indication of possible future developments

  17. Advances in nuclear physics

    CERN Document Server

    Vogt, Erich

    1975-01-01

    Review articles on three topics of considerable current interest make up the present volume. The first, on A-hypernuclei, was solicited by the editors in order to provide nuclear physicists with a general description of the most recent developments in a field which this audience has largely neglected or, perhaps, viewed as a novelty in which a bizarre nuclear system gave some information about the lambda-nuclear intersection. That view was never valid. The very recent developments reviewed here-particularly those pertaining to hypernuclear excitations and the strangeness exchange reactions-emphasize that this field provides important information about the models and central ideas of nuclear physics. The off-shell behavior of the nucleon-nucleon interaction is a topic which was at first received with some embarrassment, abuse, and neglect, but it has recently gained proper attention in many nuclear problems. Interest was first focused on it in nuclear many-body theory, but it threatened nuclear physicists'comf...

  18. Disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste

  19. Disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste.

  20. Passive Safety Systems in Advanced Water Cooled Reactors (AWCRS). Case Studies. A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    This report presents the results from the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) collaborative project (CP) on Advanced Water Cooled Reactor Case Studies in Support of Passive Safety Systems (AWCR), undertaken under the INPRO Programme Area C. INPRO was launched in 2000 - on the basis of a resolution of the IAEA General Conference (GC(44)/RES/21) - to ensure that nuclear energy is available in the 21st century in a sustainable manner, and it seeks to bring together all interested Member States to consider actions to achieve innovation. An important objective of nuclear energy system assessments is to identify 'gaps' in the various technologies and corresponding research and development (R and D) needs. This programme area fosters collaboration among INPRO Member States on selected innovative nuclear technologies to bridge technology gaps. Public concern about nuclear reactor safety has increased after the Fukushima Daiichi nuclear power plant accident caused by the loss of power to pump water for removing residual heat in the core. As a consequence, there has been an increasing interest in designing safety systems for new and advanced reactors that are passive in nature. Compared to active systems, passive safety features do not require operator intervention, active controls, or an external energy source. Passive systems rely only on physical phenomena such as natural circulation, thermal convection, gravity and self-pressurization. Passive safety features, therefore, are increasingly recognized as an essential component of the next-generation advanced reactors. A high level of safety and improved competitiveness are common goals for designing advanced nuclear power plants. Many of these systems incorporate several passive design concepts aimed at improving safety and reliability. The advantages of passive safety systems include simplicity, and avoidance of human intervention, external power or signals. For these reasons, most

  1. Fully ceramic nuclear fuel and related methods

    Science.gov (United States)

    Venneri, Francesco; Katoh, Yutai; Snead, Lance Lewis

    2016-03-29

    Various embodiments of a nuclear fuel for use in various types of nuclear reactors and/or waste disposal systems are disclosed. One exemplary embodiment of a nuclear fuel may include a fuel element having a plurality of tristructural-isotropic fuel particles embedded in a silicon carbide matrix. An exemplary method of manufacturing a nuclear fuel is also disclosed. The method may include providing a plurality of tristructural-isotropic fuel particles, mixing the plurality of tristructural-isotropic fuel particles with silicon carbide powder to form a precursor mixture, and compacting the precursor mixture at a predetermined pressure and temperature.

  2. Three-dimensional porous media based numerical investigation of spatial power distribution effect on advanced nuclear fuel rod bundles critical power

    Energy Technology Data Exchange (ETDEWEB)

    Stosic, Zoran V. [Framatome ANP GmbH . NBTT, Erlangen (Germany)], e-mail: Zoran.Stosic@Framatome-ANP.de; Stevanovic, Vladimir D. [Framatome ANP GmbH, Erlangen (Germany); Iguchi, Tadashi [Japan Atomic Energy Research Institute (JAERI), Ibaraki (Japan)

    2001-07-01

    The influence of spatial power generation shape on thermal-hydraulics behaviour of the fuel rod bundle has been investigated. Particularly, the occurrence of the local Boiling Transition has been analysed, indicating that conditions for the Critical Heat Flux (CHF) are reached somewhere within the boiling water channels in the assembly. The two-phase coolant flow within the bundle is represented with the two-fluid model in 3D space. The porous medium concept is applied in the simulation of the two-phase flow through the rod bundle implying nonequilibrium thermal and flow conditions. The governing equations in three-dimensions are discretized with the control volume method. The 3D numerical simulation and analyses of thermal-hydraulics in a complex geometry of an advanced nuclear fuel assembly are performed for conditions of a partial and/or complete rods uncovering indicating occurrence of high quality CHF - Dryout. The obtained results from numerical simulations are compared with experimental Critical Power data obtained from full scale tests. Employed is an electrically heated test rod bundle with real 1:1 geometry. Different radial and axial power distributions are used with wide range of inlet mass flow rates (2 - 19 kg/s) and coolant inlet subcooling (25 - 185 kJ/kg). The coolant pressure, equal to 6.9 MPa, is typical for BWRs conditions. Comparison of the predicted Critical Power values with measured data shows encouraging agreements for all analysed power distributions and the results completely reflect measured two-phase mixture cross flows, steam void distribution and spatial positions of Dryout onsets. Based on performed numerical investigation, an improvement of Dryout criteria is proposed. Dynamic effects of power shape change on spatial thermal hydraulics and hence on CHF occurrence as well as the influence of transfer function on thermal hydraulics under cyclic power and/or flow rate changes are also being analysed. Experiments for such verifications

  3. Method and facility for reprocessing nuclear fuels

    International Nuclear Information System (INIS)

    For reprocessing of nuclear fuels used in fuel elements with several metallic cladding tubes that are especially applied for light water reactors, the cladding tubes separated from the fuel element structure are individually cut in longitudinal direction so that the nuclear fuel can be removed from the metal parts. The nuclear fuel then is filled into an acid bath for further treatment, whereas the metal parts are conditioned in solid form for ultimate storage by embedding them in a binder. (orig./RW)

  4. Apparatus for locating defective nuclear fuel elements

    International Nuclear Information System (INIS)

    An ultrasonic search unit for locating defective fuel elements within a fuel assembly used in a water cooled nuclear reactor is presented. The unit is capable of freely traversing the restricted spaces between the fuel elements

  5. Impacts of the use of spent nuclear fuel burnup credit on DOE advanced technology legal weight truck cask GA-4 fleet size

    International Nuclear Information System (INIS)

    The object of this paper is to study the impact of full and partial spent fuel burnup credit on the capacity of the Legal Weight Truck Spent Fuel Shipping Cask (GA-4) and to determine the numbers of additional spent fuel assemblies which could be accommodated as a result. The scope of the study comprised performing nuclear criticality safety scoping calculations using the SCALE-PC software package and the 1993 spent fuel database to determine logistics for number of spent fuel assemblies to be shipped. The results of the study indicate that more capacity than 2 or 3 pressurized water reactor assemblies could be gained for GA-4 casks when burnup credit is considered. Reduction in GA-4 fleet size and number of shipments are expected to result from the acceptance of spent fuel burnup credit

  6. Compositions and methods for treating nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

    2014-01-28

    Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

  7. Compositions and methods for treating nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

    2013-08-13

    Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

  8. Recent advances in nuclear cardiology

    DEFF Research Database (Denmark)

    Gutte, H.; Petersen, C. Leth; Kjaer, A.;

    2008-01-01

    Nuclear cardiology is an essential part of functional, non-invasive, cardiac imaging. Significant advances have been made in nuclear cardiology since planar (201)thallium ((201)TI) scintigraphy was introduced for the evaluation of left ventricular (LV) perfusion nearly 40 years ago. The use...... of nuclear cardiology has been steadily increasing over the last 20 years with important steps being the introduction of (99m)technetium- ((99m)Tc)-labelled perfusion radiotracers, the change from only planar to now much more single photon emission computed tomography (SPECT) and positron emission tomography......-coronary cardiac diseases. The advances in nuclear cardiology are discussed under the four headlines of: 1) myocardial perfusion, 2) cardiac performance including LV and right ventricular (RV) function, 3) myocardial metabolism, and 4) experimental nuclear cardiology Udgivelsesdato: 2008/6...

  9. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

  10. Advanced research reactor fuel development

    International Nuclear Information System (INIS)

    The fabrication technology of the U3Si 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 U3Si2 dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U3Si2 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 ∼ 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 γ-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 U3Si2. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49

  11. Transition Towards a Sustainable Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    To support the evaluation of R and D needs and relevant technology requirements for future nuclear fuel cycles, the OECD/NEA WPFC Expert Group on Advanced Fuel Cycle Scenarios was created in 2010, replacing the WPFC Expert Group on Fuel Cycle Transition Scenario Studies (1) to assemble, organise and understand the scientific issues of advanced fuel cycles and (2) to provide a framework for assessing specific national needs related to the implementation of advanced fuel cycles. In this framework, a simulation of world transition scenarios towards possible future fuel cycles with fast reactors has been performed, using both a homogeneous and a heterogeneous approach involving different world regions. In fact, it has been found that a crucial feature of any world scenario study is to provide not only trends for an idealised 'homogeneous' description of the world, but also trends for different regions in the world, selected with simple criteria (mostly of geographical type), in order to apply different hypotheses to energy demand growth, different fuel cycle strategies and different reactor types implementation in the different regions. This approach was an attempt to avoid focusing on selected countries, in particular on those where no new spectacular energy demand growth is expected, but to provide trends and conclusions that account for the features of countries that will be major future players in the world's energy development. The heterogeneous approach considered a subdivision of the world in four main macro-regions (where countries have been grouped together according to their economic development dynamics). An original global electricity production envelope was used in simulations and a specific regional energy share was defined. In the regional approach two different fuel cycles were analysed: a once-through LWR cycle was used as the reference and a transition to fast reactor closed cycle to enable a better management of resources and minimisation of waste

  12. Proliferation Resistant Nuclear Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-02-18

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

  13. Grids for nuclear fuel elements

    International Nuclear Information System (INIS)

    This invention relates to grids for nuclear fuel assemblies with the object of providing an improved grid, tending to have greater strength and tending to offer better location of the fuel pins. It comprises sets of generally parallel strips arranged to intersect to define a structure of cellular form, at least some of the intersections including a strip which is keyed to another strip at more than one point. One type of strip may be dimpled along its length and another type of strip may have slots for keying with the dimples. (Auth.)

  14. Nuclear fuel microsphere gamma analyzer

    Science.gov (United States)

    Valentine, Kenneth H.; Long, Jr., Ernest L.; Willey, Melvin G.

    1977-01-01

    A gamma analyzer system is provided for the analysis of nuclear fuel microspheres and other radioactive particles. The system consists of an analysis turntable with means for loading, in sequence, a plurality of stations within the turntable; a gamma ray detector for determining the spectrum of a sample in one section; means for analyzing the spectrum; and a receiver turntable to collect the analyzed material in stations according to the spectrum analysis. Accordingly, particles may be sorted according to their quality; e.g., fuel particles with fractured coatings may be separated from those that are not fractured, or according to other properties.

  15. Coal and nuclear electricity fuels

    International Nuclear Information System (INIS)

    Comparative economic analysis is used to contrast the economic advantages of nuclear and coal-fired electric generating stations for Canadian regions. A simplified cash flow method is used with present value techniques to yield a single levelized total unit energy cost over the lifetime of a generating station. Sensitivity analysis illustrates the effects of significant changes in some of the cost data. The analysis indicates that in Quebec, Ontario, Manitoba and British Columbia nuclear energy is less costly than coal for electric power generation. In the base case scenario the nuclear advantage is 24 percent in Quebec, 29 percent in Ontario, 34 percent in Manitoba, and 16 percent in British Columbia. Total unit energy cost is sensitive to variations in both capital and fuel costs for both nuclear and coal-fuelled power stations, but are not very sensitive to operating and maintenance costs

  16. The DOE Advanced Gas Reactor Fuel Development and Qualification Program

    International Nuclear Information System (INIS)

    The high outlet temperatures and high thermal-energy conversion efficiency of modular High Temperature Gas-cooled Reactors (HTGRs) enable an efficient and cost effective integration of the reactor system with non-electricity generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300 C and 900 C. The Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission-product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete, fundamental understanding of the relationship between the fuel fabrication process and key fuel properties, the irradiation and accident safety performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. An overview of the program and recent progress is presented.

  17. Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

    Science.gov (United States)

    Stacey, W. M.

    2009-09-01

    The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

  18. Nuclear fuel cycle under progressing preparation of its systemisation

    International Nuclear Information System (INIS)

    Trends of nuclear development in Japan show more remarkable advancements in 2000, such as new addition of nuclear power plant, nuclear fuel cycling business, and so on. Based on an instruction of the criticality accident in JCO formed on September, 1999, government made efforts on revision of the law on regulation of nuclear reactor and so forth and establishment of a law on protection of nuclear accident as sooner, to enforce nuclear safety management and nuclear accident protective countermeasure. On the other hand, the nuclear industry field develops some new actions such as establishment of Nuclear Safety Network (NSnet)', mutual evaluation of nuclear-relative works (pier review), and so forth. And, on the high level radioactive wastes disposal of the most important subject remained in nuclear development, the Nuclear Waste Management Organization of Japan' of its main business body was established on October, 1999 together with establishment of the new law, to begin a business for embodiment of the last disposal aiming at 2030s to 2040s. On the same October, the Japan Nuclear Fuel Limited. concluded a safety agreement on premise of full-dress transportation of the used fuels to the Rokkasho Reprocessing Plant in Aomori prefecture with local government, to begin their transportation from every electric company since its year end. Here were described on development of the nuclear fuel cycling business in Japan, establishment of nuclear fuel cycling, disposal on the high level radioactive wastes, R and D on geological disposal of the high level radioactive wastes, establishment on cycle back-end of nuclear fuels, and full-dressing of nuclear fuel cycling. (G.K.)

  19. IMPULSE - advanced nuclear thermal propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Ivanenok, J.F. III; Wett, J.F. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1993-12-31

    The IMPULSE nuclear thermal rocket concept provides an evolutionary step toward high thrust-to-weight and specific impulse over a wide operating range. Most of the components and features of the concept are based on demonstrated or proven technology from the NER VA/Rover program. The performance increase is due to the use of a new solid nuclear fuel shape. The new fuel shape provides a large flow area while maintaining flow control and eliminating hot spots due to fuel-to-fuel contact. The control and eliminating hot spots due to fuel-to-fuel contact. The IMPULSE reactor utilizes a multi-pass, series flow configuration to provide excess turbine power while improving the thermal efficiency of the overall system. This configuration also provides a large area for moderator. The IMPULSE concept can provide a specific impulse of up to 1000 seconds and trust to weight ratios approaching 40. The improved performance will reduce the Initial Mass In Low Earth Orbit (IMLEO) and provide a consequent reduction in launch costs and logistics problems.

  20. Proceedings of the 2006 International Meeting on LWR fuel performance 'Nuclear Fuel: Addressing the future' - TopFuel 2006 Transactions

    International Nuclear Information System (INIS)

    From 22-26 October, 340 researchers, nuclear engineers and scientists from across Europe and beyond congregated in the ancient university city of Salamanca, Spain, to discuss the challenges facing the developers and manufacturers of new high-performance nuclear fuels-fuels that will help meet current and future energy demand and reduce man's over dependence upon CO2-emitting fossil fuels. TopFuel is an annual topical meeting organised by ENS, the American Nuclear Society and the Atomic Energy Society of Japan. This year it was co-sponsored by the IAEA, the OECD/NEA and the Spanish Nuclear Society (SNE). TopFuel's primary objective was to bring together leading specialists in the field from around the world to analyse advances in nuclear fuel management technology and to use the findings of the latest cutting-edge research to help manufacture the high performance nuclear fuels of today and tomorrow. The TopFuel 2006 agenda revolved around ten technical sessions dedicated to priority issues such as security of supply, new fuel and reactor core designs, fuel cycle strategies and spent fuel management. Among the many topics under discussion were new developments in fuel performance modelling, advanced fuel assembly design and the improved conditioning and processing of spent fuel. During the week, a poster exhibition also gave delegates the opportunity to display and discuss the results of their latest work and to network with fellow professionals. One important statement to emerge from TopFuel 2006 was that the world has enough reserves of uranium to support the large-scale and long-term production of nuclear energy. The OECD/NEA and the IAEA recently published a report entitled Uranium 2005: Resources, Production and Demand (the Red Book). The report, which makes a comprehensive assessment of uranium supplies and projected demand up until the year 2025, concludes by saying 'the uranium resource base is adequate to meet projected future requirements'. With the global

  1. Nuclear Fuels: Present and Future

    Directory of Open Access Journals (Sweden)

    Donald R. Olander

    2009-02-01

    Full Text Available The important new developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of these fuels and the reactors they power are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel-rod designs, the hydride fuel with liquid metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the Very High Temperature Reactor and the Sodium Fast Reactor, and the accompanying reprocessing technologies, aqueous-based UREX and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the material's behavior under irradiation and in the reprocessing schemes are emphasized.

  2. Sustainability Features of Nuclear Fuel Cycle Options

    OpenAIRE

    Stefano Passerini; Mujid Kazimi

    2012-01-01

    The nuclear fuel cycle is the series of stages that nuclear fuel materials go through in a cradle to grave framework. The Once Through Cycle (OTC) is the current fuel cycle implemented in the United States; in which an appropriate form of the fuel is irradiated through a nuclear reactor only once before it is disposed of as waste. The discharged fuel contains materials that can be suitable for use as fuel. Thus, different types of fuel recycling technologies may be introduced in order to more...

  3. Antineutrino monitoring of spent nuclear fuel

    OpenAIRE

    Brdar, Vedran; Huber, Patrick; Kopp, Joachim

    2016-01-01

    Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries world wide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this letter, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel eleme...

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Paul Wilson

    2009-11-02

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

  6. Evaluation and development of advanced nuclear materials: IAEA activities

    International Nuclear Information System (INIS)

    Economical, environmental and non-proliferation issues associated with sustainable development of nuclear power bring about a need for optimization of fuel cycles and implementation of advanced nuclear systems. While a number of physical and design concepts are available for innovative reactors, the absence of reliable materials able to sustain new challenging irradiation conditions represents the real bottle-neck for practical implementation of these promising ideas. Materials performance and integrity are key issues for the safety and competitiveness of future nuclear installations being developed for sustainable nuclear energy production incorporating fuel recycling and waste transmutation systems. These systems will feature high thermal operational efficiency, improved utilization of resources (both fissile and fertile materials) and reduced production of nuclear waste. They will require development, qualification and deployment of new and advanced fuel and structural materials with improved mechanical and chemical properties combined with high radiation and corrosion resistance. The extensive, diverse, and expensive efforts toward the development of these materials can be more effectively organized within international collaborative programmes with wide participation of research, design and engineering communities. IAEA carries out a number of international projects supporting interested Member States with the use of available IAEA program implementation tools (Coordinated Research Projects, Technical Meetings, Expert Reviews, etc). The presentation summarizes the activities targeting material developments for advanced nuclear systems, with particular emphasis on fast reactors, which are the focal topics of IAEA Coordinated Research Projects 'Accelerator Simulation and Theoretical Modelling of Radiation Effects' (on-going), 'Benchmarking of Structural Materials Pre-Selected for Advanced Nuclear Reactors', 'Examination of advanced fast reactor fuel and core

  7. General overview of CANDU advanced fuel cycles program

    International Nuclear Information System (INIS)

    The R and D program for CANDU advanced fuel cycles may be roughly divided into two components which have a near-and long-term focus, respectively. The near-term focus is on the technology to implement improved once-through cycles and mixed oxide (plutonium-uranium oxides) recycle in CANDU and on technologies to separate zirconium isotopes. Included is work on those technologies which would allow a CANDU-LWR strategy to be developed in a growing nuclear power system. For the longer-term, activities are focused on those technologies and fuel cycles which would be appropriate in a period when nuclear fuel demand significantly exceeds mined uranium supplies. Fuel cycles and systems under study are thorium recycle, CANDU fast breeder systems and electro-nuclear fissile breeders. The paper will discuss the rationale underlying these activities, together with a brief description of activities currently under way in each of the fuel cycle technology areas

  8. VISION - Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics

    International Nuclear Information System (INIS)

    The U.S. DOE Advanced Fuel Cycle Initiative's (AFCI) fundamental objective is to provide technology options that--if implemented--would enable long-term growth of nuclear power while improving sustainability and energy security. The AFCI organization structure consists of four areas; Systems Analysis, Fuels, Separations and Transmutations. The Systems Analysis Working Group is tasked with bridging the program technical areas and providing the models, tools, and analyses required to assess the feasibility of design and deployment options and inform key decision makers. An integral part of the Systems Analysis tool set is the development of a system level model that can be used to examine the implications of the different mixes of reactors, implications of fuel reprocessing, impact of deployment technologies, as well as potential ''exit'' or ''off ramp'' approaches to phase out technologies, waste management issues and long-term repository needs. The Verifiable Fuel Cycle Simulation Model (VISION) is a computer-based simulation model that allows performing dynamic simulations of fuel cycles to quantify infrastructure requirements and identify key trade-offs between alternatives. It is based on the current AFCI system analysis tool ''DYMOND-US'' functionalities in addition to economics, isotopic decay, and other new functionalities. VISION is intended to serve as a broad systems analysis and study tool applicable to work conducted as part of the AFCI and Generation IV reactor development studies

  9. Fuel Cycle Services the Heart of Nuclear Energy

    Directory of Open Access Journals (Sweden)

    S. Soentono

    2007-01-01

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

  10. Fuel Cycle Services The Heart of Nuclear Energy

    International Nuclear Information System (INIS)

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

  11. Advanced research workshop: nuclear materials safety

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J; Moshkov, M M

    1999-01-28

    The Advanced Research Workshop (ARW) on Nuclear Materials Safety held June 8-10, 1998, in St. Petersburg, Russia, was attended by 27 Russian experts from 14 different Russian organizations, seven European experts from six different organizations, and 14 U.S. experts from seven different organizations. The ARW was conducted at the State Education Center (SEC), a former Minatom nuclear training center in St. Petersburg. Thirty-three technical presentations were made using simultaneous translations. These presentations are reprinted in this volume as a formal ARW Proceedings in the NATO Science Series. The representative technical papers contained here cover nuclear material safety topics on the storage and disposition of excess plutonium and high enriched uranium (HEU) fissile materials, including vitrification, mixed oxide (MOX) fuel fabrication, plutonium ceramics, reprocessing, geologic disposal, transportation, and Russian regulatory processes. This ARW completed discussions by experts of the nuclear materials safety topics that were not covered in the previous, companion ARW on Nuclear Materials Safety held in Amarillo, Texas, in March 1997. These two workshops, when viewed together as a set, have addressed most nuclear material aspects of the storage and disposition operations required for excess HEU and plutonium. As a result, specific experts in nuclear materials safety have been identified, know each other from their participation in t he two ARW interactions, and have developed a partial consensus and dialogue on the most urgent nuclear materials safety topics to be addressed in a formal bilateral program on t he subject. A strong basis now exists for maintaining and developing a continuing dialogue between Russian, European, and U.S. experts in nuclear materials safety that will improve the safety of future nuclear materials operations in all the countries involved because of t he positive synergistic effects of focusing these diverse backgrounds of

  12. Chemical Kinetic Modeling of Advanced Transportation Fuels

    Energy Technology Data Exchange (ETDEWEB)

    PItz, W J; Westbrook, C K; Herbinet, O

    2009-01-20

    Development of detailed chemical kinetic models for advanced petroleum-based and nonpetroleum based fuels is a difficult challenge because of the hundreds to thousands of different components in these fuels and because some of these fuels contain components that have not been considered in the past. It is important to develop detailed chemical kinetic models for these fuels since the models can be put into engine simulation codes used for optimizing engine design for maximum efficiency and minimal pollutant emissions. For example, these chemistry-enabled engine codes can be used to optimize combustion chamber shape and fuel injection timing. They also allow insight into how the composition of advanced petroleum-based and non-petroleum based fuels affect engine performance characteristics. Additionally, chemical kinetic models can be used separately to interpret important in-cylinder experimental data and gain insight into advanced engine combustion processes such as HCCI and lean burn engines. The objectives are: (1) Develop detailed chemical kinetic reaction models for components of advanced petroleum-based and non-petroleum based fuels. These fuels models include components from vegetable-oil-derived biodiesel, oil-sand derived fuel, alcohol fuels and other advanced bio-based and alternative fuels. (2) Develop detailed chemical kinetic reaction models for mixtures of non-petroleum and petroleum-based components to represent real fuels and lead to efficient reduced combustion models needed for engine modeling codes. (3) Characterize the role of fuel composition on efficiency and pollutant emissions from practical automotive engines.

  13. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

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

    2010-02-23

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

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

    Science.gov (United States)

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

    2013-09-03

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

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

    Science.gov (United States)

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

    2011-03-01

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

  16. Advances in nuclear and radiochemistry. Extended abstracts

    International Nuclear Information System (INIS)

    The publication compiles extended abstracts of the conference. Conference topics were: fundamental nuclear chemistry (nuclear reactions, radioactive decay), actinides, transactinides, radioanalytics (nuclear and non-nuclear methods), nuclear technology (techniques, cross sections, radionuclide production), radiotracers in life sciences (radiopharmaceutical chemistry), radioactive indicators in research and chemistry, radionuclides in geochemistry and cosmochemistry, nuclear fuel cycle (waste management, transmutation, partitioning), radioecology and environmental sciences. (uke)

  17. Nuclear power performance and safety. V.5. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    The International Conference on Nuclear Power Performance and Safety, organized by the International Atomic Energy Agency, was held at the Austria Centre Vienna (ACV) in Vienna, Austria, from 28 September to 2 October 1987. The objective of the Conference was to promote an exchange of worldwide information on the current trends in the performance and safety of nuclear power and its fuel cycle, and to take a forward look at the expectations and objectives for the 1990s. Policy decisions for waste management have already been taken in many countries and the 1990s should be a period of demonstration and implementation of these policies. As ilustrated by data presented from a number of countries, many years of experience in radioactive waste management have been achieved and the technology exists to implement the national plans and policies that have been developed. The establishment of criteria, the development of safety performance methodology and site investigation work are key activities essential to the successful selection, characterization and construction of geological repositories for the final disposal of radioactive waste. Considerable work has been done in these areas over the last ten years and will continue into the 1990s. However, countries that are considering geological disposal for high level waste now recognize the need for relating the technical aspects to public understanding and acceptance of the concept and decision making activities. The real challenge for the 1990s in waste disposal will be successfully to integrate technological activities within a process which responds to institutional and public concern. Volume 5 of the Proceedings comprehends the contributions on waste management in the 1990s. Decontamination and decommissioning, waste management, treatment and disposal, nuclear fuel cycle - present and future. Enrichment services and advanced reactor fuels, improvements in reactor fuel utilization and performance, spent fuel management

  18. An introduction to the nuclear fuel cycle

    International Nuclear Information System (INIS)

    This overview of the nuclear fuel cycle is divided into three parts. First, is a brief discussion of the basic principles of how nuclear reactors work;second, is a look at the major types of nuclear reactors being used and world-wide nuclear capacity;and third, is an overview of the nuclear fuel cycle and the present industrial capability in the US. 34 figs., 10 tabs

  19. Nuclear fuel alloys or mixtures and method of making thereof

    Energy Technology Data Exchange (ETDEWEB)

    Mariani, Robert Dominick; Porter, Douglas Lloyd

    2016-04-05

    Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

  20. Advanced Fuels Campaign FY 2014 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase; W. Edgar May

    2014-10-01

    The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance.

  1. Advanced nuclear plants meet the economic challenge

    International Nuclear Information System (INIS)

    Nuclear power plants operated in the baseload regime are economically competitive even when compared with plants burning fossil fuels. As they do not produce emissions when operated, they do not pollute the environment. This is clearly reflected also in the internalized costs. After 2000, many new power plants are expected to be constructed in the USA and worldwide. An important role in this phase will be played by advanced light water reactors of the ABWR and SBWR types representing the future state of the art in technology and safety as well as in cost and plant operations management. (orig.)

  2. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    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

  3. Development of advanced mixed oxide fuels for plutonium management

    International Nuclear Information System (INIS)

    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

  4. Recent Advances in Nuclear Cardiology.

    Science.gov (United States)

    Lee, Won Woo

    2016-09-01

    Nuclear cardiology is one of the major fields of nuclear medicine practice. Myocardial perfusion studies using single-photon emission computed tomography (SPECT) have played a crucial role in the management of coronary artery diseases. Positron emission tomography (PET) has also been considered an important tool for the assessment of myocardial viability and perfusion. However, the recent development of computed tomography (CT)/magnetic resonance imaging (MRI) technologies and growing concerns about the radiation exposure of patients remain serious challenges for nuclear cardiology. In response to these challenges, remarkable achievements and improvements are currently in progress in the field of myocardial perfusion imaging regarding the applicable software and hardware. Additionally, myocardial perfusion positron emission tomography (PET) is receiving increasing attention owing to its unique capability of absolute myocardial blood flow estimation. An F-18-labeled perfusion agent for PET is under clinical trial with promising interim results. The applications of F-18 fluorodeoxyglucose (FDG) and F-18 sodium fluoride (NaF) to cardiovascular diseases have revealed details on the basic pathophysiology of ischemic heart diseases. PET/MRI seems to be particularly promising for nuclear cardiology in the future. Restrictive diseases, such as cardiac sarcoidosis and amyloidosis, are effectively evaluated using a variety of nuclear imaging tools. Considering these advances, the current challenges of nuclear cardiology will become opportunities if more collaborative efforts are devoted to this exciting field of nuclear medicine. PMID:27540423

  5. Structural behaviour from MAEF (Modified Advanced European Fuel) in the PWR Spanish Nuclear Power Plants; Comportamiento estructural del MAEF en las centrales PWR nucleares espanolas

    Energy Technology Data Exchange (ETDEWEB)

    Morales, M.; Garcia-Infanta, J. M.

    2010-07-01

    This article focuses in the results obtained in dimensional inspections carried out under the Coordinated Research Plan and other plans in Spanish Nuclear Power Plants. Data from one hundred thirty irradiated combustible elements, in three different reactors and in a wide range of burnt, are available. The results show an excellent behaviour from MAEF compared to the previous products, and confirm the achievement of the objectives set out during the design.

  6. Cermet-fueled reactors for advanced space applications

    International Nuclear Information System (INIS)

    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

  7. Communicating about advanced nuclear energy plants

    International Nuclear Information System (INIS)

    The success of advanced nuclear energy plants, as with any new product, will not depend on design alone. Success will require public support and good communications to achieve that support. In the past, communication weaknesses - including mixed and confusing messages - have sometimes created barriers between the technical community and the public. Several lessons learned from a decade of social science research in the United States of America have implications for communicating effectively about advanced design nuclear energy plants: (1) Most audiences are open-minded and receptive to communications on this topic. They view nuclear energy as a fuel of the future and want to be comfortable about the future. Most people in the USA (82%) expect future nuclear energy plants to be safer, so the improvements being made are simply consistent with public expectations. (2) Few people pay close attention to energy issues. (3) Communications must be simple and free of jargon. Because people do not pay close attention to the issues, their knowledge is limited. Some terms used by the industry to describe advanced design plants are misinterpreted. (4) Good communications focus on consumer wants and values, not industry needs or problems. People care about generational responsibility, planning for the future, environmental protection and security. (5) Benefits and safeguards should be shown instead of risk comparisons. Generic benefits of nuclear energy, such as clean air, are important to consumers. (6) Pictures and hand-on demonstrations help in communicating about nuclear energy plants, because many of the discussion concepts are abstract. (7) Trust is crucial and is established now for tomorrow through word and deed. (author)

  8. Advanced nuclear systems in comparison

    International Nuclear Information System (INIS)

    This study aims at a comparison of future reactor concepts, paying particular attention to aspects of safety, of the fuel cycle, the economics, the experience-base and the state of development. Representative examples of typical development lines, that could possibly be 'of interest' within a time horizon of 50 years were selected for comparison. This can be divided into three phases: - Phase I includes the next 10 years and will be characterised mainly by evolutionary developments of light water reactors (LWR) of large size; representative: EPR, - Phase II: i.e. the time between 2005 and 2020 approximately, encompasses the forecasted doubling of today's world-wide installed nuclear capacity; along with evolutionary reactors, innovative systems like AP600, PIUS, MHTGR, EFR will emerge, - Phase III covers the time between 2020 and 2050 and is characterised by the issue of sufficient fissile material resources; novel fast reactor systems including hybrid systems can, thus, become available; representatives: IFR, EA, ITER (the latter being). The evaluated concepts foresee partly different fuel cycles. Fission reactors can be operated in principle on the basis of either a Uranium-Plutonium-cycle or a Thorium-Uranium-cycle, while combinations of these cycles among them or with other reactor concepts than proposed are possible. With today's nuclear park (comprising mainly LWRs), the world-wide plutonium excess increases annually by about 100 t. Besides strategies based on reprocessing like: - recycling in thermal and fast reactors with mixed oxide fuels, - plutonium 'burning' in reactors with novel fuels without uranium or in 'hybrid' systems, allowing a reduction of this excess, direct disposal of spent fuel elements including their plutonium content ('one-through') is being considered. (author) figs., tabs., 32 refs

  9. Studies of nuclear fuel by means of nuclear spectroscopy methods

    Energy Technology Data Exchange (ETDEWEB)

    Jansson, Peter

    2000-02-01

    This paper is a summary text of several works performed by the author regarding spectroscopic measurements on spent nuclear fuel. Methods for determining the decay heat of spent nuclear fuel by means of gamma-ray spectroscopy and for verifying the integrity of nuclear fuel by means of tomography is presented. A summary of work performed regarding gamma-ray detector technology for studies of fission gas release is presented.

  10. A review of nuclear fuel cycle options for developing nations

    International Nuclear Information System (INIS)

    A study of several nuclear reactor and fuel cycle options for developing nations was performed. All reactor choices were considered under a GNEP framework. Two advanced alternative reactor types, a nuclear battery-type reactor and a fuel reprocessing fast reactor were examined and compared with a conventional Generation III+ LWR reactor. The burn of nuclear fuel was simulated using ORIGEN 2.2 for each reactor type and the resulting information was used to compare the options in terms of waste produced, waste quality and repository impact. The ORIGEN data was also used to evaluate the economics of the fuel cycles using unit costs, discount rates and present value functions with the material balances. The comparison of the fuel cycles and reactors developed in this work provides a basis for the evaluation of subsidy programs and cost-benefit comparisons for various reactor parameters such as repository impact and proliferation risk versus economic considerations. (authors)

  11. Nuclear Spent Fuel Management in Spain

    International Nuclear Information System (INIS)

    The radioactive waste management policy is established by the Spanish Government through the Ministry of Industry, Tourism and Commerce. This policy is described in the Cabinet-approved General Radioactive Waste Plan. ENRESA is the Spanish organization in charge of radioactive waste and nuclear SFM and nuclear installations decommissioning. The priority goal in SFM is the construction of the centralized storage facility named Almacén Temporal Centralizado (ATC), whose generic design was approved by the safety authority, Consejo de Seguridad Nuclear. This facility is planned for some 6.700 tons of heavy metal. The ATC site selection process, based on a volunteer community’s scheme, has been launched by the Government in December 2009. After the selection of a site in a participative and transparent process, the site characterization and licensing activities will support the construction of the facility. Meanwhile, extension of the on-site storage capacity has been implemented at the seven nuclear power plants sites, including past reracking at all sites. More recent activities are: reracking performed at Cofrentes NPP; dual purpose casks re-licensing for higher burnup at Trillo NPP; transfer of the spent fuel inventory at Jose Cabrera NPP to a dry-storage system, to allow decommissioning operations; and licence application of a dry-storage installation at Ascó NPP, to provide the needed capacity until the ATC facility operation. For financing planning purposes, the long-term management of spent fuel is based on direct disposal. A final decision about major fuel management options is not made yet. To assist the decision makers a number of activities are under way, including basic designs of a geological disposal facility for clay and granite host rocks, together with associated performance assessment, and supported by a R&D programme, which also includes research projects in other options like advanced separation and transmutation. (author)

  12. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    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

  13. 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

  14. Laser cutting system for nuclear fuel disassembly

    International Nuclear Information System (INIS)

    A significant advancement in fuel reprocessing technology has been made by utilizing a multikilowatt, carbon dioxide laser to perform cutting operations necessary to remove unprocessible hardware from reactor fuel assemblies. 10 figs

  15. OECD - HRP Summer School on Nuclear Fuel

    International Nuclear Information System (INIS)

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures

  16. International Summer School on Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

  17. Nuclear Fusion Fuel Cycle Research Perspectives

    International Nuclear Information System (INIS)

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

  18. OECD - HRP Summer School on Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

  19. Concepts for institutional arrangements for the nuclear fuel cycle

    International Nuclear Information System (INIS)

    These concepts deal with establishing a framework for the analysis of institutional arrangements, with institutional arrangements under consideration in the working groups on fuel and heavy water availability, enrichment availability, assurances of long-term supply, reprocessing-plutonium handling-recycling, fast breeder reactors, spent fuel management, waste management and disposal, and advanced reactor concepts. The standardization of nuclear practices, joint commercial and development undertakings, nuclear supply assurances, developing a consensus in international nuclear co-operation, and settlements of disputes are treated

  20. Variants of closing the nuclear fuel cycle

    Science.gov (United States)

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

    2015-12-01

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

  1. Advanced Recycling Reactor with Minor Actinide Fuel

    International Nuclear Information System (INIS)

    The Advanced Recycling Reactor (ARR) with minor actinide fuel has been studied. This paper presents the pre-conceptual design of the ARR proposed by the International Nuclear Recycling Alliance (INRA) for FOA study sponsored by DOE of the United States of America (U.S.). Although the basic reactor concept is technically mature, it is not suitable for commercial use due to the need to reduce capital costs. As a result of INRA's extensive experience, it is anticipated that a non-commercial ARR1 will be viable and meet U.S. requirements by 2025. Commercial Advanced Recycling Reactor (ARR) operations are expected to be feasible in competition with LWRs by 2050, based on construction of ARR2 in 2035. The ARR based on the Japan Sodium-cooled Fast Reactor (JSFR) is a loop-typed sodium cooled reactor with MOX fuel that is selected because of much experience of SFRs in the world. Major features of key technology enhancements incorporated into the ARR are the following: Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop system and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The reactor core of the ARR1 is 70 cm high and the volume fraction of fuel is 31.6%. The conversion ratio of fissile is set up less than 0.65 and the amount of burned TRU is 45-51 kg/TWeh. According to survey of more effective TRU burning core, the oxide fuel core containing high TRU (MA 15%, Pu 35% average) with moderate pins of 12% arranged driver fuel assemblies can decrease TRU conversion ratio to 0.33 and improve TRU burning capability to 67 kg/TWeh. The moderator can enhance TRU burning, while increasing the Doppler effect and reducing the positive sodium void effect. High TRU fraction promotes TRU burning by curbing plutonium production. High Am fraction and Am blanket promote Am transmutation. The ARR1 consists of a reactor building (including

  2. Spent nuclear fuel disposal liability insurance

    International Nuclear Information System (INIS)

    This thesis examines the social efficiency of nuclear power when the risks of accidental releases of spent fuel radionuclides from a spent fuel disposal facility are considered. The analysis consists of two major parts. First, a theoretical economic model of the use of nuclear power including the risks associated with releases of radionuclides from a disposal facility is developed. Second, the costs of nuclear power, including the risks associated with a radionuclide release, are empirically compared to the costs of fossil fuel-fired generation of electricity. Under the provisions of the Nuclear Waste Policy Act of 1982, the federally owned and operated spent nuclear fuel disposal facility is not required to maintain a reserve fund to cover damages from an accidental radionuclide release. Thus, the risks of a harmful radionuclide release are not included in the spent nuclear fuel disposal fee charged to the electric utilities. Since the electric utilities do not pay the full, social costs of spent fuel disposal, they use nuclear fuel in excess of the social optimum. An insurance mechanism is proposed to internalize the risks associated with spent fueled disposal. Under this proposal, the Federal government is required to insure the disposal facility against any liabilities arising from accidental releases of spent fuel radionuclides

  3. Spent Nuclear Fuel (SNF) Project Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  4. Spent Nuclear Fuel (SNF) Project Execution Plan

    International Nuclear Information System (INIS)

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities

  5. Nuclear Fuel Cycle Information System. A directory of nuclear fuel cycle facilities. 2009 ed

    International Nuclear Information System (INIS)

    The Nuclear Fuel Cycle Information System (NFCIS) is an international directory of civilian nuclear fuel cycle facilities, published online as part of the Integrated Nuclear Fuel Cycle Information System (iNFCIS: http://www-nfcis.iaea.org/). This is the fourth hardcopy publication in almost 30 years and it represents a snapshot of the NFCIS database as of the end of 2008. Together with the attached CD-ROM, it provides information on 650 civilian nuclear fuel cycle facilities in 53 countries, thus helping to improve the transparency of global nuclear fuel cycle activities

  6. The safety of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The nuclear fuel cycle covers the procurement and preparation of fuel for nuclear power reactors, its recovery and recycling after use and the safe storage of all wastes generated through these operations. The facilities associated with these activities have an extensive and well documented safety record accumulated over the past 40 years by technical experts and safety authorities. This report constitutes an up-to-date analysis of the safety of the nuclear fuel cycle, based on the available experience in OECD countries. It addresses the technical aspects of fuel cycle operations, provides information on operating practices and looks ahead to future activities

  7. Nuclear fuels for very high temperature applications

    Energy Technology Data Exchange (ETDEWEB)

    Lundberg, L.B.; Hobbins, R.R.

    1992-08-01

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO{sub 2} or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

  8. Nuclear fuels for very high temperature applications

    International Nuclear Information System (INIS)

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO2 or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures

  9. Nuclear fuels for very high temperature applications

    Energy Technology Data Exchange (ETDEWEB)

    Lundberg, L.B.; Hobbins, R.R.

    1992-01-01

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO{sub 2} or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

  10. Advanced fuel technology and performance: Current status and trends

    International Nuclear Information System (INIS)

    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

  11. LWR spent fuel storage technology: Advances and experience

    International Nuclear Information System (INIS)

    By 2003, the year the US Department of Energy (DOE) currently predicts a repository will be available, 58 domestic commercial nuclear-power plant units are expected to run out of wet storage space for LWR spent fuel. To alleviate this problem, utilities implemented advances in storage methods that increased storage capacity as well as reduced the rate of generating spent fuel. Those advances include (1) transhipping spent-fuel assemblies between pools within the same utility system, (2) reracking pools to accommodate additional spent-fuel assemblies, (3) taking credit for fuel burnup in pool storage rack designs, (4) extending fuel burnup, (5) rod consolidation, and (6) dry storage. The focus of this paper is on advances in rod consolidation and dry storage. Wet storage continues to be the predominant US spent-fuel management technology, but as a measure to enhance at-reactor storage capacity, the Nuclear Waste Policy Act of 1982 authorized DOE to assist utilities with licensing at-reactor dry storage. Information exchanges with other nations, laboratory testing and modeling, and cask tests cooperatively funded by US utilities and DOE produced a strong technical basis to develop confidence that LWR spent fuel can be stored safely for several decades in both wet and dry modes. Licensed dry storage of spent fuel in an inert atmosphere was first achieved in the US in 1986. Studies are underway in several countries to determine acceptable conditions for storing LWR spent fuel in air. Rod-consolidation technology is being developed and demonstrated to enhance the capacity for both wet and dry storage. Large-scale commercial implementation is awaiting optimization of practical and economical mechanical systems. 22 refs., 1 fig

  12. Overview of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The nuclear fuel cycle is substantially more complicated than the energy production cycles of conventional fuels because of the very low abundance of uranium 235, the presence of radioactivity, the potential for producing fissile nuclides from irradiation, and the risk that fissile materials will be used for nuclear weapons. These factors add enrichment, recycling, spent fuel storage, and safeguards to the cycle, besides making the conventional steps of exploration, mining, processing, use, waste disposal, and transportation more difficult

  13. Fuel Cycle Services the Heart of Nuclear Energy

    OpenAIRE

    S. Soentono

    2007-01-01

    Fuel is essential for development whether for survival and or wealth creation purposes. In this century the utilization of fuels need to be improved although energy mix is still to be the most rational choice. The large amount utilization of un-renewable fossil has some disadvantages since its low energy content requires massive extraction, transport, and processing while emitting CO2 resulting degradation of the environment. In the mean time the advancement of nuclear science and technology ...

  14. Impact of actinide recycle on nuclear fuel cycle health risks

    International Nuclear Information System (INIS)

    The purpose of this background paper is to summarize what is presently known about potential impacts on the impacts on the health risk of the nuclear fuel cycle form deployment of the Advanced Liquid Metal Reactor (ALMR)1 and Integral Fast Reactor (IF)2 technology as an actinide burning system. In a companion paper the impact on waste repository risk is addressed in some detail. Therefore, this paper focuses on the remainder of the fuel cycle

  15. Used Nuclear Fuel: From Liability to Benefit

    Science.gov (United States)

    Orbach, Raymond L.

    2011-03-01

    Nuclear power has proven safe and reliable, with operating efficiencies in the U.S. exceeding 90%. It provides a carbon-free source of electricity (with about a 10% penalty arising from CO2 released from construction and the fuel cycle). However, used fuel from nuclear reactors is highly toxic and presents a challenge for permanent disposal -- both from technical and policy perspectives. The half-life of the ``bad actors'' is relatively short (of the order of decades) while the very long lived isotopes are relatively benign. At present, spent fuel is stored on-site in cooling ponds. Once the used fuel pools are full, the fuel is moved to dry cask storage on-site. Though the local storage is capable of handling used fuel safely and securely for many decades, the law requires DOE to assume responsibility for the used fuel and remove it from reactor sites. The nuclear industry pays a tithe to support sequestration of used fuel (but not research). However, there is currently no national policy in place to deal with the permanent disposal of nuclear fuel. This administration is opposed to underground storage at Yucca Mountain. There is no national policy for interim storage---removal of spent fuel from reactor sites and storage at a central location. And there is no national policy for liberating the energy contained in used fuel through recycling (separating out the fissionable components for subsequent use as nuclear fuel). A ``Blue Ribbon Commission'' has been formed to consider alternatives, but will not report until 2012. This paper will examine alternatives for used fuel disposition, their drawbacks (e.g. proliferation issues arising from recycling), and their benefits. For recycle options to emerge as a viable technology, research is required to develop cost effective methods for treating used nuclear fuel, with attention to policy as well as technical issues.

  16. Studies of Nuclear Fuel by Means of Nuclear Spectroscopic Methods

    OpenAIRE

    Jansson, Peter

    2002-01-01

    The increasing demand for characterization of nuclear fuel, both from an operator and authority point of view, motivates the development of new experimental and, preferable, non-destructive methods. In this thesis, some methods based on nuclear spectroscopic techniques are presented. Various parameters of irradiated fuel are shown to be determined with high accuracy and confidence by utilizing gamma-ray scanning, tomography and passive neutron assay. Specifically, fuel parameters relevant for...

  17. Candu advanced fuel cycles: key to energy sustainability

    International Nuclear Information System (INIS)

    A primary rationale for Indonesia to proceed with a nuclear power program is to diversity its energy sources and achieve freedom from future resource constraints. While other considerations, such as economy of power supply, hedging against potential future increases in the price of fossil fuels, fostering the technological development of the Indonesia economy and minimizing greenhouse and other gaseous are important, the strategic resource issue is key. In considering candidate nuclear power technologies upon which to base such a program, a major consideration will be the potential for those technologies to be economically sustained in the face of large future increases in demand for nuclear fuels. the technology or technologies selected should be amenable to evaluation in a rapidly changing technical, economic, resource and environmental policy environment. the world's proven uranium resources which can be economically recovered represent a fairly modest energy resource if utilization is based on the currently commercialized fuel cycles, even with the use of recovered plutonium in mixed oxide fuels. In the long term, fuel cycles relying solely on the use of light water reactors will encounter increasing fuel supply constraints. Because of its outstanding neutron economy and the flexibility of on-power refueling, Candu reactors are the most fuel resource efficient commercial reactors and offer the potential for accommodating an almost unlimited variety of advanced and even more fuel efficient cycles. Most of these cycles utilize nuclear fuel which are too low grade to be used in light water reactors, including many products now considered to be waste, such as spent light water reactor fuel and reprocessing products such as recovered uranium. The fuel-cycle flexibility of the Candu reactor provides a ready path to sustainable energy development in both the short and the long terms. Most of the potential Candu fuel cycle developments can be accommodated in existing

  18. Investigation of Spent Nuclear Fuel Pool Coolability

    OpenAIRE

    Nimander, Fredrik

    2011-01-01

    The natural catastrophe at Fukushima Dai-ichi 2011 enlightened the nuclear community. This master thesis reveals the non-negligible risks regarding the short term storage of spent nuclear fuel. The thesis has also investigated the possibility of using natural circulation of air in a passive safety system to cool the spent nuclear fuel pools. The results where conclusive: The temperature difference between the heated air and ambient air is far too low for natural circulation of air to remove a...

  19. Establishment of China Nuclear Fuel Assembly Database

    Institute of Scientific and Technical Information of China (English)

    CHENPeng; ZHANGYing-chao; LIUTing-jin; JINYong-li

    2003-01-01

    During researching, designing, manufacturing and post irradiation, a large amount of data on fuel assembly of China nuclear power plants has been accumulated. It is necessary to collect the data together,so that the researchers, designers, manufactures and managers could use the data conveniently. It was proposed to establish a China Nuclear Fuel Assembly Database through the Internet on workstations during the year of 2003 to 2006, so the data would be shared in China nuclear industry.

  20. 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.

  1. Risk management and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    If nuclear fuel is the answer to the future energy crisis, more must be done in the area of protecting financial interests. This paper discusses what has been done in the area of insurance to protect the owner, processor, vendors, etc. What is available in the insurance market is reviewed; the Nuclear Energy Liability Property Insurance Association is virtually the only nuclear insuror, except for the mutual company Nuclear Mutual Limited in Bermuda. Methods being used today to insure each phase of the processing for nuclear fuel are reviewed next. There are basically three (overlapping) types of primary insurance for the fuel cycle: conventional insurance, nuclear insurance pools, and Price-Anderson indemnification. There is no clearcut assumption of risk because the contract between owner, converter, fabricator or reprocessor is usually completed before insurance is considered. The need to educate the insurors about nuclear matters is emphasized

  2. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    This report, the fifth of a series of annual reports, reviews the progress that has been made in the research and development program for the safe management and disposal of Canada's nuclear fuel waste. The report summarizes activities over the past year in the following areas: public interaction; used fuel storage and transportation; immobilization of used fuel and fuel recycle waste; geoscience research related to deep underground disposal; environmental research; and environmental and safety assessment

  3. PNC`s proposal on the Advanced Fuel Recycle concept

    Energy Technology Data Exchange (ETDEWEB)

    Kamiya, Masayoshi; Shinoda, Yoshihiko; Ojima, Hisao [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works

    1998-03-01

    MOX fuel for FBR is allowed to contain impurities within several thousand ppm, which means less than 1000 of decontamination factor (DF) in reprocessing is enough for Pu and U recycle use. The Advanced Fuel Recycle proposed by PNC is on this basis. The concept consists of innovations on both MOX fuel fabrication and aqueous reprocessing technologies based on the Purex process and it is believed that successful optimization of fuel cycle interface condition is the key issue to realize the concept. The lower DF such as 1000 can be easily obtained by the simplified Purex flowsheet which has no purification steps. However, new subject arises in MOX fuel fabrication, that is, fabrication is conducted in the shielding cell using equipment which is maintained remotely. A simplified fabrication technology becomes essential to establish the remote maintenance system and is one of the critical path for achieving the Advanced Fuel Recycle. The PNC`s proposal on the advanced fuel recycle concept consists of modified PUREX process having single extraction cycle and crystallization, Remote fuel fabrication such as gelation and vibro-packing. In the Advanced Fuel Recycle concept, as it is low DF cycle system, all processes should be installed in remote maintenance cells. Then both reprocessing and fabrication facility would be able to be integrated into a same building. Integrated fuel cycle plant has several merits. No transportation of nuclear material between reprocessing and fabrication enhances non-proriferation aspect in addition to the low-DF concept. Cost performance is also improved because of optimization and rationalization of auxiliary equipment, and so on. (author)

  4. National Policy on Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    National policy on nuclear fuel cycle is aimed at attaining the expected condition, i.e. being able to support optimality the national energy policy and other related Government policies taking into account current domestic nuclear fuel cycle condition and the trend of international nuclear fuel cycle development, the national strength, weakness, thread and opportunity in the field of energy. This policy has to be followed by the strategy to accomplish covering the optimization of domestic efforts, cooperation with other countries, and or purchasing licences. These policy and strategy have to be broken down into various nuclear fuel cycle programmes covering basically assesment of the whole cycle, performing research and development of the whole cycle without enrichment and reprocessing being able for weapon, as well as programmes for industrialization of the fuel cycle stepwisery commencing with the middle part of the cycle and ending with the edge of the back-end of the cycle

  5. Advanced fuel developments to improve fuel cycle cost in PWR

    International Nuclear Information System (INIS)

    Increasingly lower fuel cycle costs and higher plant availability factors have been two crucial components in keeping the overall cost of electricity produced by nuclear low and competitive with respect to other energy sources. The continuous quest to reduce fuel cycle cost has resulted in some consolidated trends in LWR fuel management schemes: smaller number of feed fuel assemblies with longer residence time; longer cycles, with 18-month cycle as the predominant option, and some plants already operating on, or considering, 24-month refueling intervals; higher power ratings with many plants undergoing power uprates. In order to maintain or improve fuel utilization for the longer cycles and/or higher power ratings, the licensed limits in fuel fissile content (5.0 w/o U235 enrichment) and discharge burnup (62 GWd/tHM for the peak pin) have been approached. In addition, Zr-based fuel cladding materials are also being challenged by the resulting increased duty. For the above reasons further improvements in fuel cycle cost have to overcome one or more of the current limits. This paper discusses an option to break through this 'stalemate', i.e. uranium nitride (UN) fuel with SiC clad. In UN the higher density of the nitride with respect to the oxide fuel leads to higher fissile content and reduction in the number of feed assemblies, improved fuel utilization and potentially higher specific powers. The SiC clad, among other benefits, enables higher clad irradiation, thereby exploiting the full potential of UN fuel. An alternative to employing UN fuel is to maintain UO2 fuel but boost the fissile content increasing the U235 enrichment beyond the 5 w/o limit. The paper describes and compares the potential benefits on fuel cycle cost of either option using realistic full-core calculations and ensuing economic analysis performed using Westinghouse in-house reactor physics tools and methodologies. (author)

  6. Advanced ceramic materials for next-generation nuclear applications

    Science.gov (United States)

    Marra, John

    2011-10-01

    The nuclear industry is at the eye of a 'perfect storm' with fuel oil and natural gas prices near record highs, worldwide energy demands increasing at an alarming rate, and increased concerns about greenhouse gas (GHG) emissions that have caused many to look negatively at long-term use of fossil fuels. This convergence of factors has led to a growing interest in revitalization of the nuclear power industry within the United States and across the globe. Many are surprised to learn that nuclear power provides approximately 20% of the electrical power in the US and approximately 16% of the world-wide electric power. With the above factors in mind, world-wide over 130 new reactor projects are being considered with approximately 25 new permit applications in the US. Materials have long played a very important role in the nuclear industry with applications throughout the entire fuel cycle; from fuel fabrication to waste stabilization. As the international community begins to look at advanced reactor systems and fuel cycles that minimize waste and increase proliferation resistance, materials will play an even larger role. Many of the advanced reactor concepts being evaluated operate at high-temperature requiring the use of durable, heat-resistant materials. Advanced metallic and ceramic fuels are being investigated for a variety of Generation IV reactor concepts. These include the traditional TRISO-coated particles, advanced alloy fuels for 'deep-burn' applications, as well as advanced inert-matrix fuels. In order to minimize wastes and legacy materials, a number of fuel reprocessing operations are being investigated. Advanced materials continue to provide a vital contribution in 'closing the fuel cycle' by stabilization of associated low-level and high-level wastes in highly durable cements, ceramics, and glasses. Beyond this fission energy application, fusion energy will demand advanced materials capable of withstanding the extreme environments of high

  7. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-03-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) organized a Cladding and Coatings operational meeting February 12-13, 2013, at Oak Ridge National Laboratory (ORNL). Representatives from the U.S. Department of Energy (DOE), national laboratories, industry, and universities attended the two-day meeting. The purpose of the meeting was to discuss advanced cladding and cladding coating research and development (R&D); review experimental testing capabilities for assessing accident tolerant fuels; and review industry/university plans and experience in light water reactor (LWR) cladding and coating R&D.

  8. Strategic research of advanced fuel cycle technologies in JNC

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  9. Strategic research of advanced fuel cycle technologies in JNC

    International Nuclear Information System (INIS)

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

  10. Nuclear fuels and materials irradiation technology development in HANARO

    International Nuclear Information System (INIS)

    The equipments for the irradiation tests of nuclear fuels and materials in the HANARO are classified into a capsule and an FTL (Fuel Test Loop). Capsules for irradiation tests of nuclear fuels and materials in HANARO have been developed. Also, extensive efforts have been made to establish the design/manufacturing and irradiation technologies for irradiating nuclear fuels and materials by using these capsules and their control systems, which should be compatible with HANARO's characteristics. Other devices consisting of a fixing of the capsule during an irradiation test in the HANARO, a cutting and a transporting of the capsule main body after an irradiation test were also developed. These capsules and others have been actively utilized for various material irradiation tests requested by users. Based on the accumulated experiences and a user's sophisticated requirements, capsules for a creep test and a fatigue test of materials during an irradiation in HANARO have been developed. And, the irradiation plans related to developing the Gen-IV reactor systems by using capsules in HANARO will mean more emphasis on the development of capsules by focusing on the irradiation tests of materials or nuclear fuels for Gen-IV reactor systems, such as the SFR and the VHTR. The FTL is one of the irradiation devices, which can conduct an irradiation test of a nuclear fuel in HANARO under the operating conditions of commercial nuclear power plants. The 3-test fuel rods can be irradiated in HANARO by using the FTL. The installation of the FTL was completed in March 2007. Currently, the commissioning test of the FTL is being performed. At first the FTL will be used for the irradiation test of an advanced nuclear fuel for a PWR from the end of this year. In this paper, the status and the perspective in the field of material irradiation tests in HANARO are described. (author)

  11. Annotated Bibliography for Drying Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rebecca E. Smith

    2011-09-01

    Internationally, the nuclear industry is represented by both commercial utilities and research institutions. Over the past two decades many of these entities have had to relocate inventories of spent nuclear fuel from underwater storage to dry storage. These efforts were primarily prompted by two factors: insufficient storage capacity (potentially precipitated by an open-ended nuclear fuel cycle) or deteriorating quality of existing underwater facilities. The intent of developing this bibliography is to assess what issues associated with fuel drying have been identified, to consider where concerns have been satisfactorily addressed, and to recommend where additional research would offer the most value to the commercial industry and the U. S. Department of Energy.

  12. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-05-01

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

  13. Survey of advanced nuclear technologies for potential applications of sonoprocessing.

    Science.gov (United States)

    Rubio, Floren; Blandford, Edward D; Bond, Leonard J

    2016-09-01

    Ultrasonics has been used in many industrial applications for both sensing at low power and processing at higher power. Generally, the high power applications fall within the categories of liquid stream degassing, impurity separation, and sonochemical enhancement of chemical processes. Examples of such industrial applications include metal production, food processing, chemical production, and pharmaceutical production. There are many nuclear process streams that have similar physical and chemical processes to those applications listed above. These nuclear processes could potentially benefit from the use of high-power ultrasonics. There are also potential benefits to applying these techniques in advanced nuclear fuel cycle processes, and these benefits have not been fully investigated. Currently the dominant use of ultrasonic technology in the nuclear industry has been using low power ultrasonics for non-destructive testing/evaluation (NDT/NDE), where it is primarily used for inspections and for characterizing material degradation. Because there has been very little consideration given to how sonoprocessing can potentially improve efficiency and add value to important process streams throughout the nuclear fuel cycle, there are numerous opportunities for improvement in current and future nuclear technologies. In this paper, the relevant fundamental theory underlying sonoprocessing is highlighted, and some potential applications to advanced nuclear technologies throughout the nuclear fuel cycle are discussed.

  14. Survey of advanced nuclear technologies for potential applications of sonoprocessing.

    Science.gov (United States)

    Rubio, Floren; Blandford, Edward D; Bond, Leonard J

    2016-09-01

    Ultrasonics has been used in many industrial applications for both sensing at low power and processing at higher power. Generally, the high power applications fall within the categories of liquid stream degassing, impurity separation, and sonochemical enhancement of chemical processes. Examples of such industrial applications include metal production, food processing, chemical production, and pharmaceutical production. There are many nuclear process streams that have similar physical and chemical processes to those applications listed above. These nuclear processes could potentially benefit from the use of high-power ultrasonics. There are also potential benefits to applying these techniques in advanced nuclear fuel cycle processes, and these benefits have not been fully investigated. Currently the dominant use of ultrasonic technology in the nuclear industry has been using low power ultrasonics for non-destructive testing/evaluation (NDT/NDE), where it is primarily used for inspections and for characterizing material degradation. Because there has been very little consideration given to how sonoprocessing can potentially improve efficiency and add value to important process streams throughout the nuclear fuel cycle, there are numerous opportunities for improvement in current and future nuclear technologies. In this paper, the relevant fundamental theory underlying sonoprocessing is highlighted, and some potential applications to advanced nuclear technologies throughout the nuclear fuel cycle are discussed. PMID:27400217

  15. Nuclear fuel conversion and fabrication chemistry

    International Nuclear Information System (INIS)

    Following irradiation and reprocessing of nuclear fuel, two operations are performed to prepare the fuel for subsequent reuse as fuel: fuel conversion, and fuel fabrication. These operations complete the classical nuclear fuel cycle. Fuel conversion involves generating a solid form suitable for fabrication into nuclear fuel. For plutonium based fuels, either a pure PuO2 material or a mixed PuO2-UO2 fuel material is generated. Several methods are available for preparation of the pure PuO2 including: oxalate or peroxide precipitation; or direct denitration. Once the pure PuO2 is formed, it is fabricated into fuel by mechanically blending it with ceramic grade UO2. The UO2 can be prepared by several methods which include direct denitration. ADU precipitation, AUC precipitation, and peroxide precipitation. Alternatively, UO2-PuO2 can be generated directly using coprecipitation, direct co-denitration, or gel sphere processes. In coprecipitation, uranium and plutonium are either precipitated as ammonium diuranate and plutonium hydroxide or as a mixture of ammonium uranyl-plutonyl carbonate, filtered and dried. In direct thermal denitration, solutions of uranium and plutonium nitrates are heated causing concentration and, subsequently, direct denitration. In gel sphere conversion, solutions of uranium and plutonium nitrate containing additives are formed into spherical droplets, gelled, washed and dried. Refabrication of these UO3-PuO2 starting materials is accomplished by calcination-reduction to UO2-PuO2 followed by pellet fabrication. (orig.)

  16. Overview of the nuclear fuel cycle

    International Nuclear Information System (INIS)

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

  17. Radioecology of nuclear fuel cycles

    International Nuclear Information System (INIS)

    Sites where radioactive wastes are found are solid waste burial grounds, soils below liquid stoage areas, surface ditches and ponds, and the terrestrial environment around chemical processing facilities that discharge airborne radioactive debris from stacks. This study provides information to help assess the environmental impacts and certain potentiall human hazards associated with nuclear fuel cycles. A data base is being developed to define and quantify biological transport routes which will permit credible predictions and assessment of routine and potential large-scale releases of radionuclides and other toxic materials. These data, used in assessment models, will increase the accuracy of estimating radiation doses to man and other life forms. Information obtained from existing storage and disposal sites will provide a meaningful radioecological perspective with which to improve the effectiveness of waste management practices. Results will provide information to determine if waste management procedures on the Hanford Site have caused ecological perturbations, and if so, to determine the source, nature, and magnitude of such disturbances

  18. Next generation advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    Growing energy demand by technological developments and the increase of the world population and gradually diminishing energy resources made nuclear power an indispensable option. The renewable energy sources like solar, wind and geothermal may be suited to meet some local needs. Environment friendly nuclear energy which is a suitable solution to large scale demands tends to develop highly economical, advanced next generation reactors by incorporating technological developments and years of operating experience. The enhancement of safety and reliability, facilitation of maintainability, impeccable compatibility with the environment are the goals of the new generation reactors. The protection of the investment and property is considered as well as the protection of the environment and mankind. They became economically attractive compared to fossil-fired units by the use of standard designs, replacing some active systems by passive, reducing construction time and increasing the operation lifetime. The evolutionary designs were introduced at first by ameliorating the conventional plants, than revolutionary systems which are denoted as generation IV were verged to meet future needs. The investigations on the advanced, proliferation resistant fuel cycle technologies were initiated to minimize the radioactive waste burden by using new generation fast reactors and ADS transmuters.

  19. World nuclear fuel cycle requirements 1989

    International Nuclear Information System (INIS)

    This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under two nuclear supply scenarios. These two scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries in the World Outside Centrally Planned Economic Areas (WOCA). A No New Orders scenarios is also presented for the Unites States. This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the WOCA projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel; discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2020 for the Lower and Upper Reference cases and through 2036, the last year in which spent fuel is discharged, for the No New Orders case

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

    Energy Technology Data Exchange (ETDEWEB)

    J. Carmack; L. Braase; F. Goldner

    2015-09-01

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

  1. 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).

  2. Nuclear Fuel Cycle Technologies: Current Challenges and Future Plans - 12558

    International Nuclear Information System (INIS)

    The mission of the Office of Nuclear Energy's Fuel Cycle Technologies office (FCT program) is to provide options for possible future changes in national nuclear energy programs. While the recent draft report of the Blue Ribbon Commission on America's Nuclear Future stressed the need for organization changes, interim waste storage and the establishment of a permanent repository for nuclear waste management, it also recognized the potential value of alternate fuel cycles and recommended continued research and development in that area. With constrained budgets and great expectations, the current challenges are significant. The FCT program now performs R and D covering the entire fuel cycle. This broad R and D scope is a result of the assignment of new research and development (R and D) responsibilities to the Office of Nuclear Energy (NE), as well as reorganization within NE. This scope includes uranium extraction from seawater and uranium enrichment R and D, used nuclear fuel recycling technology, advanced fuel development, and a fresh look at a range of disposal geologies. Additionally, the FCT program performs the necessary systems analysis and screening of fuel cycle alternatives that will identify the most promising approaches and areas of technology gaps. Finally, the FCT program is responsible for a focused effort to consider features of fuel cycle technology in a way that promotes nonproliferation and security, such as Safeguards and Security by Design, and advanced monitoring and predictive modeling capabilities. This paper and presentation will provide an overview of the FCT program R and D scope and discuss plans to analyze fuel cycle options and support identified R and D priorities into the future. The FCT program is making progress in implanting a science based, engineering driven research and development program that is evaluating options for a sustainable fuel cycle in the U.S. Responding to the BRC recommendations, any resulting legislative changes

  3. The US Advanced Fuel Cycle Programme: Objectives and Accomplishments

    International Nuclear Information System (INIS)

    For approximately a decade, the United States Department of Energy has been conducting an advanced fuel cycle programme, presently named the Fuel Cycle R and D Program, devoted to lessening both the environmental burden of nuclear energy and the proliferation risk of accumulating used nuclear fuel. Currently, the programme is being redirected towards a science based, goal oriented focus with the objective of deploying successfully demonstrated technology in the 2040-2050 time frame. The present paper reports the key considerations of the science based research approach, the elements of the technical programme and the accomplishments in fast reactor research and development, the goal of which is to improve the primary issues that have inhibited fast reactor introduction in the past, namely, economics and safety. (author)

  4. Levelized nuclear fueling cost in Israel

    International Nuclear Information System (INIS)

    Basic nuclear fuel cycle mode options are discussed as they apply to PWR-type reactors. Forecast fueling costs have been computed and are reported for the two main choices - basic front-end cost supplemented by either a throw- away mode option or a reprocessing mode option. It is concluded that reprocessing could result in total unit fueling costs ranging from a minimum slightly lower, through a maximum about 30% higher than the total unit fueling cost using the throw-away mode option. Moreover, in massive breeder development the total unit fueling cost can extend even below the numerically calculated limit. (H.K.)

  5. Reference Neutron Radiographs of Nuclear Reactor Fuel

    OpenAIRE

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as...

  6. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    Energy Technology Data Exchange (ETDEWEB)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  7. The Technology Trend of Japanese Patent for the Nuclear Fuel Assembly Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo; Lee, Nam Ho; Jeong, Kyung Min; Suh, Yong Chil; Kim, Chang Hoi; Shin, Jung Cheol

    2008-06-15

    Japanese technology patents for the nuclear fuel assembly inspection unit, from the year 1993 to the year 2006, were investigated. The fuel rods which contain fissile material are grouped together in a closely-spaced array within the fuel assembly. Various kinds of reactor including the PWR reactor are being operated in Japan. There are many kinds of nuclear fuel assemblies in Japan, and the shape and the size of these nuclear fuel assemblies are various also. As the structure of these various fuel assemblies is a regular square as the same as the Korean one, the inspection method described in Japanese technology patent can be applied to the inspection of the nuclear fuel assembly of the Korea. This report focuses on advances in VIT(visual inspection test) of nuclear fuel assembly using the state-of-the-art CCD camera system.

  8. Development of advanced expansion due to compression (A-EDC) test method for safety evaluation of degraded nuclear fuel cladding materials

    International Nuclear Information System (INIS)

    Expansion due to compression (EDC) test has been applied to evaluate the performance of nuclear fuel claddings where pellet-cladding mechanical interaction (PCMI) is introduced by swelling of fuel pellets and is triggered by the larger hoop deformation of the pellets, especially during accidental transients. The purpose of this study is to modify the EDC test to describe PCMI, specimen volume reduction and others. Ring-shaped specimens were cut from Zry-4 cladding tubes. Cylindrical metal pellets with 8 mm in diameter and 15 mm in maximum height were used as inner pellets. Expansion of the specimens due to the inner pellet compression was performed at room temperature. The experimental data were further analyzed by finite element method. Through the survey in the variation of the specimen and core, specimen size and inner pellet geometry were optimized. Excellent reproducibility with less error was confirmed. The uniaxial tension condition in the hoop direction up to the specimen failure was confirmed. Hoop stress–hoop strain curves were successfully derived. (author)

  9. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  10. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  11. Structural evaluation of Siemens advanced fuel channel under accident loadings

    International Nuclear Information System (INIS)

    As a part of an effort to develop an advanced BWR fuel channel design, Siemens Power Corporation (SPC) and the Siemens AG Power Generation Group (KWU) performed structural analyses to verify the acceptability of the fuel channel design under combined seismic/LOCA (Loss Of. Coolant Accident) loadings. The results of the analyses give some interesting insights into the problem: 1) fluid-structure interaction (FSI) effects are significant and should be considered, 2) the problem may simplified by using a linear analysis despite non-linear features (gaps) between interfacing components, and 3) sufficient accuracy may be obtained by using only the first mode of vibration. The channeled fuel assembly can be considered to be a beam where the flexural stiffness is primarily determined by the fuel channel and the mass is given by the fuel assembly. The results from the analyses show the advanced fuel channel design meets applicable design criteria with adequate margins while at the same time exhibiting superior nuclear performance compared to a conventional BWR fuel channel. (author)

  12. Dry Refabrication Technology Development of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Key technical data on advanced nuclear fuel cycle technology development for the spent fuel recycling have been produced in this study. In the frame work of DUPIC, dry process oxide products fabrication, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remote modulated welding equipment has been designed and fabricated. In the area of advanced pre-treatment process development, a rotary-type oxidizer and spherical particle fabrication process were developed by using SIMFUEL and off-gas treatment technology and zircalloy tube treatment technology were studied. In the area of the property characteristics of dry process products, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data

  13. 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.

  14. Development of advanced LWR fuel cladding

    International Nuclear Information System (INIS)

    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

  15. Spacer for fuel rods in nuclear fuel elements

    International Nuclear Information System (INIS)

    Spacers for fuel rods in nuclear reactor fuel elements are described, especially for use aboard ships. Spacers are used in a grid formed by web plates orthogonally intersecting and assembled together in a tooth-comb fashion forming a plurality of channels. The web plates are joined together and each of the web plates includes apertures through which resilient and separator members are joined. The resilient and separator members are joined. The resilient and separator members are in adjacent channels and with other similar members in the same channel, contact a fuel rod in the channel. The contact pressure between the members and fuel rod is radially directed

  16. Nuclear spent fuel management. Experience and options

    International Nuclear Information System (INIS)

    Spent nuclear fuel can be stored safely for long periods at relatively low cost, but some form of permanent disposal will eventually be necessary. This report examines the options for spent fuel management, explores the future prospects for each stage of the back-end of the fuel cycle and provides a thorough review of past experience and the technical status of the alternatives. Current policies and practices in twelve OECD countries are surveyed

  17. On-Going Comparison of Advanced Fuel Cycle Options

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Ralph G. Bennett; Brent W. Dixon; J. Stephen Herring; David E. Shropshire; Mark Roth; J. D. Smith; Robert Hill; James Laidler; Kemal Pasamehmetoglu

    2004-10-01

    The Advanced Fuel Cycle Initiative (AFCI) program is addressing key issues associated with critical national needs. This paper compares the major options with these major “outcome” objectives - waste geological repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety as well as “process” objectives associated with readiness to proceed and adaptability and robustness in the face of uncertainties. Working together, separation, transmutation, and fuel technologies provide complete energy systems that can improve waste management compared to the current “once-through/no separation” approach. Future work will further increase confidence in potential solutions, optimize solutions for the mixtures of objectives, and develop attractive development and deployment paths for selected options. This will allow the nation to address nearer-term issues such as avoiding the need for additional geological repositories while making nuclear energy a more sustainable energy option for the long-term. While the Generation IV Initiative is exploring multiple reactor options for future nuclear energy for both electricity generation and additional applications, the AFCI is assessing fuel cycles options for either a continuation or expansion of nuclear energy in the United States. This report compares strategies and technology options for managing the associated spent fuel. There are four major potential strategies, as follows: · The current U.S. strategy is once through: standard nuclear power plants, standard fuel burnup, direct geological disposal of spent fuel. Variants include higher burnup fuels in water-cooled power plants, once-through gas-cooled power plants, and separation (without recycling) of spent fuel to reduce the number and cost of geological waste packages. · The second strategy is thermal recycle, recycling some fuel components in thermal reactors. This strategy extends the useful life of

  18. Nuclear Fuel Cycle Reasoner: PNNL FY12 Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-03

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

  19. Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Weber, William J. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Materials Science and Engineering; Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Materials Science and Engineering

    2016-09-20

    This is the final report of the NEUP project “Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms.” This project started on July 1, 2012 and was successfully completed on June 30, 2016. This report provides an overview of the main achievements, results and findings through the duration of the project. Additional details can be found in the main body of this report and in the individual Quarterly Reports and associated Deliverables of this project, which have been uploaded in PICS-NE. The objective of this research was to advance understanding and develop validated models on the effects of self-radiation from beta and alpha decay on the response of used nuclear fuel and nuclear waste forms during high-temperature interim storage and long-term permanent disposition. To achieve this objective, model used-fuel materials and model waste form materials were identified, fabricated, and studied.

  20. Waste management and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The present lecture deals with energy needs and nuclear power, the importance of waste and its relative place in the fuel cycle, the games of controversies over nuclear waste in the strategies of energy and finally with missions and functions of the IAEA for privileging the rational approach and facilitating the transfer of technology. (RW)

  1. 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.

  2. Integrated spent nuclear fuel database system

    International Nuclear Information System (INIS)

    The Distributed Information Systems software Unit at the Idaho National Engineering Laboratory has designed and developed an Integrated Spent Nuclear Fuel Database System (ISNFDS), which maintains a computerized inventory of all US Department of Energy (DOE) spent nuclear fuel (SNF). Commercial SNF is not included in the ISNFDS unless it is owned or stored by DOE. The ISNFDS is an integrated, single data source containing accurate, traceable, and consistent data and provides extensive data for each fuel, extensive facility data for every facility, and numerous data reports and queries

  3. Advances in HTGR spent fuel treatment technology

    International Nuclear Information System (INIS)

    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

  4. Advanced fuel cycle on the basis of pyroelectrochemical process for irradiated fuel reprocessing and vibropacking technology

    International Nuclear Information System (INIS)

    For advanced nuclear fuel cycle in SSC RIAR there is developed the pyroelectrochemical process to reprocess irradiated fuel and produce granulated oxide fuel UO2, PuO2 or (U,Pu)O2 from chloride melts. The basic technological stage is the extraction of oxides as a crystal product with the methods either of the electrolysis (UO2 and UO2-PuO2) or of the precipitating crystalIization (PuO2). After treating the granulated fuel is ready for direct use to manufacture vibropacking fuel pins. Electrochemical model for (U,Pu)O2 coprecipitation is described. There are new processes being developed: electroprecipitation of mixed oxides - (U,Np)O2, (U,Pu,Np)O2, (U,Am)O2 and (U,Pu,Am)O2. Pyroelectrochemical production of mixed actinide oxides is used both for reprocessing spent fuel and for producing actinide fuel. Both the efficiency of pyroelectrochemical methods application for reprocessing nuclear fuel and of vibropac technology for plutonium recovery are estimated. (author)

  5. Long term wet spent nuclear fuel storage

    International Nuclear Information System (INIS)

    The meeting showed that there is continuing confidence in the use of wet storage for spent nuclear fuel and that long-term wet storage of fuel clad in zirconium alloys can be readily achieved. The importance of maintaining good water chemistry has been identified. The long-term wet storage behaviour of sensitized stainless steel clad fuel involves, as yet, some uncertainties. However, great reliance will be placed on long-term wet storage of spent fuel into the future. The following topics were treated to some extent: Oxidation of the external surface of fuel clad, rod consolidation, radiation protection, optimum methods of treating spent fuel storage water, physical radiation effects, and the behaviour of spent fuel assemblies of long-term wet storage conditions. A number of papers on national experience are included

  6. Nuclear Fuels & Materials Spotlight Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    I. J. van Rooyen,; T. M. Lillo; Y. Q. WU; P.A. Demkowicz; L. Scott; D.M. Scates; E. L. Reber; J. H. Jackson; J. A. Smith; D.L. Cottle; B.H. Rabin; M.R. Tonks; S.B. Biner; Y. Zhang; R.L. Williamson; S.R. Novascone; B.W. Spencer; J.D. Hales; D.R. Gaston; C.J. Permann; D. Anders; S.L. Hayes; P.C. Millett; D. Andersson; C. Stanek; R. Ali; S.L. Garrett; J.E. Daw; J.L. Rempe; J. Palmer; B. Tittmann; B. Reinhardt; G. Kohse; P. Ramuhali; H.T. Chien; T. Unruh; B.M. Chase; D.W. Nigg; G. Imel; J. T. Harris

    2014-04-01

    As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

  7. Vertical integration in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Vertical integration in the nuclear fuel cycle and its contribution to market power of integrated fuel suppliers were studied. The industry subdivision analyzed is the uranium raw materials sector. The hypotheses demonstrated are that (1) this sector of the industry is trending toward vertical integration between production of uranium raw materials and the manufacture of nuclear fuel elements, and (2) this vertical integration confers upon integrated firms a significant market advantage over non-integrated fuel manufacturers. Under microeconomic concepts the rationale for vertical integration is the pursuit of efficiency, and it is beneficial because it increases physical output and decreases price. The Market Advantage Model developed is an arithmetical statement of the relative market power (in terms of price) between non-integrated nuclear fuel manufacturers and integrated raw material/fuel suppliers, based on the concept of the ''squeeze.'' In operation, the model compares net profit and return on sales of nuclear fuel elements between the competitors, under different price and cost circumstances. The model shows that, if integrated and non-integrated competitors sell their final product at identical prices, the non-integrated manufacturer returns a net profit only 17% of the integrated firm. Also, the integrated supplier can price his product 35% below the non-integrated producer's price and still return the same net profit. Vertical integration confers a definite market advantage to the integrated supplier, and the basic source of that advantage is the cost-price differential of the raw material, uranium

  8. The State of the WWER Nuclear Spent Fuel Management in Ukraine and Trends on the Optimal Choice of Spent Fuel Management Strategy

    International Nuclear Information System (INIS)

    The paper gives an overview of the operational state of the away from reactor interim spent fuel storage facility at the Zaporizhzhya NPP site, commissioned according to the dry storage cask technology that is developed by the US company Sierra Nuclear Corporation. The trends of the SFM at the WWER nuclear units were noted, which are caused by the advanced nuclear fuel implementation in order to improve the nuclear fuel utilization. (author)

  9. Advanced fuel in the Budapest research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hargitai, T.; Vidovsky, I. [KFKI Atomic Energy Research Inst., Budapest (Hungary)

    1997-07-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  10. Nuclear Fuel Cycle Evaluation and Real Options

    Directory of Open Access Journals (Sweden)

    L. Havlíček

    2008-01-01

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

  11. Significant incidents in nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    In contrast to nuclear power plants, events in nuclear fuel cycle facilities are not well documented. The INES database covers all the nuclear fuel cycle facilities; however, it was developed in the early 1990s and does not contain information on events prior to that. The purpose of the present report is to collect significant events and analyze them in order to give a safety related overview of nuclear fuel cycle facilities. Significant incidents were selected using the following criteria: release of radioactive material or exposure to radiation; degradation of items important to safety; and deficiencies in design, quality assurance, etc. which include criticality incidents, fire, explosion, radioactive release and contamination. This report includes an explanation, where possible, of root causes, lessons learned and action taken. 4 refs, 4 tabs

  12. Fundamental aspects of nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D.R.

    1976-01-01

    The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

  13. Fuel element for nuclear reactor

    International Nuclear Information System (INIS)

    In order to avoid a can box or an adjacent fuel element sitting on the spacer of a fuel element in the corner during assembly, the top and bottom edges of the outer bars of the spacers are provided with deflector bars, which have projections projecting beyond the outside of the outer bars. (orig.)

  14. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

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

  15. Dry Transfer Systems for Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Brett W. Carlsen; Michaele BradyRaap

    2012-05-01

    The potential need for a dry transfer system (DTS) to enable retrieval of used nuclear fuel (UNF) for inspection or repackaging will increase as the duration and quantity of fuel in dry storage increases. This report explores the uses for a DTS, identifies associated general functional requirements, and reviews existing and proposed systems that currently perform dry fuel transfers. The focus of this paper is on the need for a DTS to enable transfer of bare fuel assemblies. Dry transfer systems for UNF canisters are currently available and in use for transferring loaded canisters between the drying station and storage and transportation casks.

  16. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-05-01

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

  17. Advanced methods for fabrication of PHWR and LMFBR fuels

    International Nuclear Information System (INIS)

    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 UO2 pellet-pins. The advanced PHWR fuels are UO2-PuO2 (≤ 2 per cent), ThO2-PuO2 (≤ 4 per cent) and ThO2-U233O2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O2 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 UO2, PuO2 and ThO2 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

  18. Nuclear integrated database and design advancement system

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Jae Joo; Jeong, Kwang Sub; Kim, Seung Hwan; Choi, Sun Young

    1997-01-01

    The objective of NuIDEAS is to computerize design processes through an integrated database by eliminating the current work style of delivering hardcopy documents and drawings. The major research contents of NuIDEAS are the advancement of design processes by computerization, the establishment of design database and 3 dimensional visualization of design data. KSNP (Korea Standard Nuclear Power Plant) is the target of legacy database and 3 dimensional model, so that can be utilized in the next plant design. In the first year, the blueprint of NuIDEAS is proposed, and its prototype is developed by applying the rapidly revolutionizing computer technology. The major results of the first year research were to establish the architecture of the integrated database ensuring data consistency, and to build design database of reactor coolant system and heavy components. Also various softwares were developed to search, share and utilize the data through networks, and the detailed 3 dimensional CAD models of nuclear fuel and heavy components were constructed, and walk-through simulation using the models are developed. This report contains the major additions and modifications to the object oriented database and associated program, using methods and Javascript.. (author). 36 refs., 1 tab., 32 figs.

  19. Nuclear integrated database and design advancement system

    International Nuclear Information System (INIS)

    The objective of NuIDEAS is to computerize design processes through an integrated database by eliminating the current work style of delivering hardcopy documents and drawings. The major research contents of NuIDEAS are the advancement of design processes by computerization, the establishment of design database and 3 dimensional visualization of design data. KSNP (Korea Standard Nuclear Power Plant) is the target of legacy database and 3 dimensional model, so that can be utilized in the next plant design. In the first year, the blueprint of NuIDEAS is proposed, and its prototype is developed by applying the rapidly revolutionizing computer technology. The major results of the first year research were to establish the architecture of the integrated database ensuring data consistency, and to build design database of reactor coolant system and heavy components. Also various softwares were developed to search, share and utilize the data through networks, and the detailed 3 dimensional CAD models of nuclear fuel and heavy components were constructed, and walk-through simulation using the models are developed. This report contains the major additions and modifications to the object oriented database and associated program, using methods and Javascript.. (author). 36 refs., 1 tab., 32 figs

  20. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  1. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

  2. Panorama 2010: Nuclear fuel resources

    International Nuclear Information System (INIS)

    The abundance of projects to build nuclear power plants, the desire of new countries to acquire civil atomic power, contracts sometimes deemed fantastically high for the operation of uranium mines, etc. All of these signals indicate a return to nuclear power in a context dominated by the fight against global warming. But can nuclear power make a durable contribution to the effort to meet the ever-increasing demand for energy? (author)

  3. Development of a computer program of fast calculation for the pre design of advanced nuclear fuel 10 x 10 for BWR type reactors

    International Nuclear Information System (INIS)

    In the National Institute of Nuclear Research (ININ) a methodology is developed to optimize the design of cells 10x10 of assemble fuels for reactors of water in boil or BWR. It was proposed a lineal calculation formula based on a coefficients matrix (of the change reason of the relative power due to changes in the enrichment of U-235) for estimate the relative powers by pin of a cell. With this it was developed the computer program of fast calculation named PreDiCeldas. The one which by means of a simple search algorithm allows to minimize the relative power peak maximum of cell or LPPF. This is achieved varying the distribution of U-235 inside the cell, maintaining in turn fixed its average enrichment. The accuracy in the estimation of the relative powers for pin is of the order from 1.9% when comparing it with results of the 'best estimate' HELIOS code. With the PreDiCeldas it was possible, at one minimum time of calculation, to re-design a reference cell diminishing the LPPF, to the beginning of the life, of 1.44 to a value of 1.31. With the cell design with low LPPF is sought to even design cycles but extensive that those reached at the moment in the BWR of the Laguna Verde Central. (Author)

  4. International nuclear fuel cycle evaluation (INFCE)

    International Nuclear Information System (INIS)

    The study describes and analyzes the structures, the procedures and decision making processes of the International Nuclear Fuel Cycle Evaluation (INFCE). INFCE was agreed by the Organizing Conference to be a technical and analytical study and not a negotiation. The results were to be transmitted to governments for their consideration in developing their nuclear energy policies and in international discussions concerning nuclear energy cooperation and related controls and safeguards. Thus INFCE provided a unique example for decision making by consensus in the nuclear world. It was carried through under mutual respect for each country's choices and decisions, without jeopardizing their respective fuel cycle policies or international co-operation agreements and contracts for the peaceful use of nuclear energy, provided that agreed safeguards are applied. (orig.)

  5. Assessment and balancing of nuclear fuels

    International Nuclear Information System (INIS)

    In 1981 nuclear energy had a share of ca. 17% in the electric power supply of the F.R. of Germany. The amount of nuclear fuels required is equal to ca. 15 million tce. In public technical discussions the economic importance which must be assigned to nuclear energy, e.g. with regard to curbing the energy price development or relieving our balance of payments, is discussed in detail. On the other hand, a number of industrial aspects of nuclear energy utilization - problems of commercial or fiscal law - have been little considered in the technical literature. The following contribution is to present the principles of commercial and fiscal law which have taken shape in connection with the assessment and balancing of the single stages of the nuclear fuel cycle. (orig./UA)

  6. Sustaining nuclear fuel science and technology base

    International Nuclear Information System (INIS)

    To fulfil energy demand, the Indonesian Government has made efforts to optimize the use of various-fossil and non fossil-potential energy resources in synergy (energy mix), which is stated in national energy policy. According to national energy policy, Indonesia is going to use nuclear energy for electricity supply, and up to 2025, the use of nuclear energy is projected at about 2% of the total primary energy or 4 to 5% of the national electricity supply. This energy demand is described in NPP road map, which consists of NPP preparation, construction and operation up to 2025. To sustain the activity of nuclear power plants, the continuity of nuclear reactor fuel supply is an absolute necessity; therefore, it will become industrially prospective and have an effect on national industries. As a nuclear research center and guidance in nuclear energy system in Indonesia, Batan also plays a role to promote this prospect and to increase the national content at NPP construction. In this point of view, Batan should have the competency especially in nuclear fuel cycle technology, and in this case PTBN is viewed as the competent center since PTBN's main task is to conduct the development of Nuclear Fuel Technology. This competency is performed as mastering its science and technology base. In this case, PTBN is noticed to have the capability to function suitably since PTBN is equipped with documents for fuel fabrication industry such as bidding, construction and commissioning and qualified man power. Basically, PTBN does not have the mandatory to operate nuclear fuel fabrication commercially. However, PTBN has the capability to prepare competent man power through training and coaching in nuclear fuel fabrication. In fact, the present condition shows that some of the equipments does not function properly or are not utilized optimally or are not operable. Besides, the process documents available have not yet validated and qualified, and the man power is not qualified yet

  7. An analysis of international nuclear fuel supply options

    Science.gov (United States)

    Taylor, J'tia Patrice

    As the global demand for energy grows, many nations are considering developing or increasing nuclear capacity as a viable, long-term power source. To assess the possible expansion of nuclear power and the intricate relationships---which cover the range of economics, security, and material supply and demand---between established and aspirant nuclear generating entities requires models and system analysis tools that integrate all aspects of the nuclear enterprise. Computational tools and methods now exist across diverse research areas, such as operations research and nuclear engineering, to develop such a tool. This dissertation aims to develop methodologies and employ and expand on existing sources to develop a multipurpose tool to analyze international nuclear fuel supply options. The dissertation is comprised of two distinct components: the development of the Material, Economics, and Proliferation Assessment Tool (MEPAT), and analysis of fuel cycle scenarios using the tool. Development of MEPAT is aimed for unrestricted distribution and therefore uses publicly available and open-source codes in its development when possible. MEPAT is built using the Powersim Studio platform that is widely used in systems analysis. MEPAT development is divided into three modules focusing on: material movement; nonproliferation; and economics. The material movement module tracks material quantity in each process of the fuel cycle and in each nuclear program with respect to ownership, location and composition. The material movement module builds on techniques employed by fuel cycle models such as the Verifiable Fuel Cycle Simulation (VISION) code developed at the Idaho National Laboratory under the Advanced Fuel Cycle Initiative (AFCI) for the analysis of domestic fuel cycle. Material movement parameters such as lending and reactor preference, as well as fuel cycle parameters such as process times and material factors are user-specified through a Microsoft Excel(c) data spreadsheet

  8. MOLTEN FLUORIDE NUCLEAR REACTOR FUEL

    Science.gov (United States)

    Barton, C.J.; Grimes, W.R.

    1960-01-01

    Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

  9. Nuclear fuel supply view in Argentina

    Energy Technology Data Exchange (ETDEWEB)

    Cirimello, R.O. [Comision Nacional de Energia Atomica, Conuar SA (Argentina)

    1997-07-01

    The Argentine Atomic Energy Commission promoted and participated in a unique achievement in the R and D system in Argentina: the integration of science technology and production based on a central core of knowledge for the control and management of the nuclear fuel cycle technology. CONUAR SA, as a fuel manufacturer, FAE SA, the manufacturer of Zircaloy tubes, CNEA and now DIOXITEC SA producer of Uranium Dioxide, have been supply, in the last ten years, the amount of products required for about 1300 Tn of equivalent U content in fuels. The most promising changes for the fuel cycle economy is the Slight Enriched Uranium project which begun in Atucha I reactor. In 1997 seventy five fuel assemblies, equivalent to 900 Candu fuel bundles, will complete its irradiation. (author)

  10. Nuclear reactor fuel assembly with fuel rod removal means

    International Nuclear Information System (INIS)

    A fuel assembly is described for a nuclear reactor. The assembly has a bottom nozzle, at least one longitudinally extending control rod guide thimble attached to and projecting upwardly from the bottom nozzle and transverse grids spaced along the thimble. An organized array of elongated fuel rods are transversely spaced and supported by the grids and axially captured between the bottom nozzle and a top nozzle. The assembly comprises: (a) a transversely extending adapter plate formed by an arrangement of integral cross-laced ligaments defining a plurality of coolant flow openings; (b) means for mounting the adapter plate on an upper end portion of the thimble and spaced axially above and disposed transversely over the upper ends of all of the fuel rods present in the fuel assembly such that ones of the ligaments overlie corresponding ones of the fuel rods so as to prevent the fuel rods from moving upwardly through the coolant flow openings; and (c) removable plug means confined within the adapter plate and positioned over and spaced axially above selected ones of the fuel rods in providing access to at least one fuel rod for removal thereof upwardly through the axially spaced adapter plate without removing the top nozzle from the fuel assembly

  11. 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.

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

    Science.gov (United States)

    Settle, Frank A.

    2009-01-01

    The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and engineering of controlled fission are central to the generation of nuclear power, chemistry…

  13. Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad Abir; Fahima Islam; Hyoung Koo Lee; Daniel Wachs

    2014-11-01

    The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the High Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.

  14. Modeling of advanced fossil fuel power plants

    Science.gov (United States)

    Zabihian, Farshid

    The first part of this thesis deals with greenhouse gas (GHG) emissions from fossil fuel-fired power stations. The GHG emission estimation from fossil fuel power generation industry signifies that emissions from this industry can be significantly reduced by fuel switching and adaption of advanced power generation technologies. In the second part of the thesis, steady-state models of some of the advanced fossil fuel power generation technologies are presented. The impacts of various parameters on the solid oxide fuel cell (SOFC) overpotentials and outputs are investigated. The detail analyses of operation of the hybrid SOFC-gas turbine (GT) cycle when fuelled with methane and syngas demonstrate that the efficiencies of the cycles with and without anode exhaust recirculation are close, but the specific power of the former is much higher. The parametric analysis of the performance of the hybrid SOFC-GT cycle indicates that increasing the system operating pressure and SOFC operating temperature and fuel utilization factor improves cycle efficiency, but the effects of the increasing SOFC current density and turbine inlet temperature are not favourable. The analysis of the operation of the system when fuelled with a wide range of fuel types demonstrates that the hybrid SOFC-GT cycle efficiency can be between 59% and 75%, depending on the inlet fuel type. Then, the system performance is investigated when methane as a reference fuel is replaced with various species that can be found in the fuel, i.e., H2, CO2, CO, and N 2. The results point out that influence of various species can be significant and different for each case. The experimental and numerical analyses of a biodiesel fuelled micro gas turbine indicate that fuel switching from petrodiesel to biodiesel can influence operational parameters of the system. The modeling results of gas turbine-based power plants signify that relatively simple models can predict plant performance with acceptable accuracy. The unique

  15. Nuclear energy and its fuel cycle, prospects to 2025

    International Nuclear Information System (INIS)

    Nuclear power will supply an increasing share of the world's electricity but will expand more slowly than had been expected, and no shortages of uranium or other fuel cycle services are foreseen before the end of the century. While exploration for new uranium deposits should continue to ensure long-term supplies, advances in reactor design and enrichment and reprocessing techniques could achieve reductions in uranium demand

  16. Nuclear fuel element having oxidation resistant cladding

    International Nuclear Information System (INIS)

    This patent describes an improved nuclear fuel element of the type including a zirconium alloy tube, a zirconium barrier layer metallurgically bonded to the inside surface of the alloy tube, and a central core of nuclear fuel material partially filling the inside of the tube so as to leave a gap between the sponge zirconium barrier and the nuclear fuel material. The improvement comprising an alloy layer formed on the inside surface of the zirconium barrier layer. The alloy layer being composed of one or more impurities present in a thin layer region of the zirconium barrier in amounts less than 1% by weight but sufficient to inhibit the oxidation of the inside surface of the zirconium barrier layer without substantially affecting the plastic properties of the barrier layer, wherein the impurities are selected from the group consisting of iron, chromium, copper, nitrogen, and niobium

  17. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1992-05-01

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

  18. International nuclear fuel cycle fact book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1988-01-01

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

  19. International nuclear fuel cycle fact book

    International Nuclear Information System (INIS)

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

  20. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I W; Mitchell, S J

    1990-01-01

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

  1. International Nuclear Fuel Cycle Fact Book

    International Nuclear Information System (INIS)

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

  2. International nuclear fuel cycle fact book

    International Nuclear Information System (INIS)

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

  3. Leaf spring puller for nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Fogg, J.L.

    1981-11-03

    A fuel rod puller in the form of a collet for pulling fuel rods from a storage area into grids of a nuclear reactor fuel assembly. The rod puller moves longitudinally through the grids to a storage area where projections on the end of leaf springs grasp onto an end plug in a fuel rod. Drive apparatus then pulls the rod puller and connected fuel rod from the storage area into the fuel assembly grids. The rod puller includes an outer tube having leaf springs on one end thereof in one modification, mounted within the outer tube is a movable plunger which acts to urge the leaf springs outwardly to a position to permit passing or with the end of a end plug. Upon withdrawal of the plunger, the leaf springs move into a groove formed in the end of a fuel rod end plug, and the fuel rod subsequently is pulled into the fuel assembly grids. In another modification, the leaf springs on the outer rod are biased in an outward direction and a longitudinally movable tube on the outer rod is moved in a direction to contract the leaf springs into a position where the projections thereof engage the groove formed in a fuel rod end plug.

  4. Nuclear fuel supply: challenges and opportunities

    Energy Technology Data Exchange (ETDEWEB)

    Lowen, S. [Cameco Corp., Saskatoon, Saskatchewan (Canada)

    2006-07-01

    Prices of uranium, conversion services and enrichment services have all significantly increased in the last few years. These price increases have generally been driven by a tightening in the supply of these products and services, mostly due to long lead times required to bring these products and services to the market. This paper will describe the various steps in the nuclear fuel cycle for natural and enriched uranium fuel, will discuss the development of the front-end fuel cycle for low void reactivity fuel, and will address the challenges faced in the long-term supply of each component, particularly in the light of potential demand increases as a result of a nuclear renaissance. The opportunities for new capacity and uranium production will be outlined and the process required to achieve sufficient new supply will be discussed. (author)

  5. Economic projection for advanced fuel fabrication

    International Nuclear Information System (INIS)

    This paper presents a fabrication cost evaluation based on the status of MTR fuel development at NUKEM as of November 1980, and includes advanced chemical and mechanical processes starting with UF6 (including chemical scrap recovery). It also assumes that there will be just one fabrication line for each fuel type. UAlx-Al fuel with a uranium density of 1.0 g/cm3 was taken as the 100% cost reference. Excluding enrichment for the moment, the 100% reference values for all fuels are individually different for each reactor because of the different fuel element designs. Plate type, thickness of meat, length of meat, number of plates per element, bent or flat plates/tubes, upper and lower components, assembly design, materials, element quantity per order, etc., are all important in determining the final costs. New activity in research and development (R and D) has been started at NUKEM. In the very beginning, all the steps involved in fabrication were examined, starting with UF6 conversion and chemical treatment of the fuel (mainly, the oxide fuels), and continuing through powder production and plate fabrication, including intermediate chemical scrap recovery and final scrap recovery. The latter is very important because these procedures can also be used in reprocessing later on. If the mechanical or technical capability and/or the costs no longer makes sense, that R and D activity is stopped if there is an alternative fuel. At the moment, the prospects for alternative fuels are promising. The R and D on a particular fuel is stopped if its cost to the reactor operator would be increased by more than 30% over the current fuel cost. For this reason, the R and D on UAl alloy fuel was stopped about two years ago at a uranium density of roughly 1.2 g/cm3. Similarly, the R and D on UAIx fuel is now being stopped at around 2.2 g U/cm3 because the 30% limit has been reached. The R and D on U3O8 fuel is continuing up to about 3.2 g U/cm3, but there is some possibility of achieving

  6. Spent nuclear fuel project integrated schedule plan

    Energy Technology Data Exchange (ETDEWEB)

    Squires, K.G.

    1995-03-06

    The Spent Nuclear Fuel Integrated Schedule Plan establishes the organizational responsibilities, rules for developing, maintain and status of the SNF integrated schedule, and an implementation plan for the integrated schedule. The mission of the SNFP on the Hanford site is to provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it to final disposition. This particularly involves K Basin fuel.

  7. Method of manufacturing nuclear fuel sintered product

    International Nuclear Information System (INIS)

    Purpose: To eliminate various restrictions in view of the production such as addition amount of organic additives and obtain sintered fuels of excellent burning property. Method: Metal oxide powder for use in nuclear fuels is selected from UO2 and Gd2O3. Further, organic material additives are selected from those constituted with carbon and at least one of nitrogen, oxygen and hydrogen, such as succinic acid and maleic acid. Further, another metal oxide powder for use in nuclear fuels is selected from U3O8, (Gd,O)3O8 at a higher oxidized state than that of the previously mentioned metal oxide powder for use in nuclear fuels. These materials are mixed and molded into starting powder for nuclear fuels. Then, the molding products are sintered in a reducing atmosphere. It is thus possible to obtain normal fine structures by sintering in a usual reducing atmosphere while eliminating the restrictions for the addition amount of the organic additives or using no particular additive removing furnace. (T.M.)

  8. World nuclear fuel market. Seventeenth annual meeting

    International Nuclear Information System (INIS)

    The papers presented at the seventeenth World Nuclear Fuels Market meeting are cataloged individually. This volume includes information on the following areas of interest: historical and current aspects of the uranium and plutonium market with respect to supply and demand, pricing, spot market purchasing, and other market phenomena; impact of reprocessing and recycling uranium, plutonium, and mixed oxide fuels; role of individual countries in the market: Hungary, Germany, the Soviet Union, Czechoslovakia, France, and the US; the impact of public opinion and radioactive waste management on the nuclear industry, and a debate regarding long term versus short term contracting by electric utilities for uranium and enrichment services

  9. Long island to Limerick, nuclear fuel transfer

    International Nuclear Information System (INIS)

    The issue described is: how to move 33 shipments of radioactive nuclear fuel - 200 tons of enriched uranium pellets - on rail cars through the heart of Philadelphia, without upsetting politicians, the media and anti-nuclear activists, after a similar plan to move the fuel through New York City had been rejected in a political disaster. The answer to this is: Strategic Communications Planning. At PECO Energy's department of Corporate and Public Affairs, the research is quite clear that in risk management situations like this, the side that gets out front with the most credible information inevitably wins. That is exactly what was set out to do

  10. Reference Neutron Radiographs of Nuclear Reactor Fuel

    DEFF Research Database (Denmark)

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

  11. Chemical reprocessing of spent nuclear fuels

    International Nuclear Information System (INIS)

    The reprocessing of nuclear fuels from atomic power stations has a twofold goal. On the one hand it is serving for fuel supply by recovering the fissile materials which have not been consumed or which have been freshly generated in the reactor. On the other hand the radioactive waste products from nuclear power generation are pretreated for long-term safe disposal. The core element of the chemical processing is the PUREX Process, a counter-current solvent extraction procedure using tributyl phosphate (TBP) as the solvent for uranium and plutonium. The chemical basis and the technological performance of the process are discussed. (orig.)

  12. Seismic response of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Hlaváč Z.

    2014-06-01

    Full Text Available The paper deals with mathematical modelling and computer simulation of the seismic response of fuel assembly components. The seismic response is investigated by numerical integration method in time domain. The seismic excitation is given by two horizontal and one vertical synthetic accelerograms at the level of the pressure vessel seating. Dynamic response of the hexagonal type nuclear fuel assembly is caused by spatial motion of the support plates in the reactor core investigated on the reactor global model. The modal synthesis method with condensation is used for calculation of the fuel assembly component displacements and speeds on the level of the spacer grid cells.

  13. Nuclear fuel assembly identification using computer vision

    International Nuclear Information System (INIS)

    This report describes an improved method of remotely identifying irradiated nuclear fuel assemblies. The method uses existing in-cell TV cameras to input an image of the notch-coded top of the fuel assemblies into a computer vision system, which then produces the identifying number for that assembly. This system replaces systems that use either a mechanical mechanism to feel the notches or use human operators to locate notches visually. The system was developed for identifying fuel assemblies from the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor, but could be used for other reactor assembly identification, as appropriate

  14. Nuclear Thermal Propulsion for Advanced Space Exploration

    Science.gov (United States)

    Houts, M. G.; Borowski, S. K.; George, J. A.; Kim, T.; Emrich, W. J.; Hickman, R. R.; Broadway, J. W.; Gerrish, H. P.; Adams, R. B.

    2012-01-01

    The fundamental capability of Nuclear Thermal Propulsion (NTP) is game changing for space exploration. A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on NTP could provide high thrust at a specific impulse above 900 s, roughly double that of state of the art chemical engines. Characteristics of fission and NTP indicate that useful first generation systems will provide a foundation for future systems with extremely high performance. The role of the NCPS in the development of advanced nuclear propulsion systems could be analogous to the role of the DC-3 in the development of advanced aviation. Progress made under the NCPS project could help enable both advanced NTP and advanced Nuclear Electric Propulsion (NEP).

  15. Options contracts in the nuclear fuel industry

    International Nuclear Information System (INIS)

    This article discusses options trading in the nuclear fuels industry. Although there now exists no formal options market in the nuclear industry, flexibilities, or embedded options, are actually quite common in the long-term supply contracts. The value of these flexibilities can be estimated by applying the methods used to evaluate options. The method used is the Black-Scholes Model, and it is applied to a number of examples

  16. On the applicability of a correlation of critical heat flux prediction for advanced nuclear fuel from Angra-1; Confirmando a aplicabilidade de uma correlacao de previsao do fluxo de calor critico para o combustivel nuclear avancado de Angra-1

    Energy Technology Data Exchange (ETDEWEB)

    Palheiros, Franklin L.; Gomes, Sydney [Industrias Nucleares do Brasil SA, Rio de Janeiro, RJ (Brazil). Gerencia de Analise Tecnica do Combustivel]. E-mail: franklin@inb.gov.br; sydney@inb.gov.br

    2005-07-01

    To establish the nuclear core operation limits it is requested that the safe condition of heat transfer between the fuel rod and the primary system coolant shall be well known. Those limits are set based on the critical heat flux marking the upper limit of nucleate boiling where the interaction of the coolant causes a departure from nucleate boiling - DNB [1]. For the new fuel type (16NGF) to be used at the Angra-1, the Kori-2, and the Krsko sites, the Critical Heat Flux must be characterized as well. Compared to the 16x16STD fuel (a 25+ year old Westinghouse design), the 16NGF design has many new features, including: three Intermediate Flow Mixers (IFM's) were added in the grid spans 4, 5 and 6 to improve the mixing behavior and consequently the thermal margin; an optimized vane pattern was used as well as an improved vane design (RFA Mod-1 vane). Others important thermal characteristics are the use of an optimized fuel rod diameter- from the U/H ratio point of view (0.360'' instead of 0.374'') and others features (Reference 2).Therefore, to demonstrate the applicability of an existing CHF correlation for the 16NGF design, two tests were performed in the Columbia University's Heat Transfer Research Facility, New York City (HTRF). Denominated Tests number 112 and number 113, they employed a 5x5 bundle array with a heated length of 12 ft. and cosine axial power shape. The purpose of this paper is then to present the data analysis from the 16NGF CHF tests and demonstrate, through a statistical evaluation, that the resulting data when evaluated is 'poolable' (combinable) with the WRB-2 database. In so doing, it can be thus stated that the WRB-2 correlation is applicable to 16 NGF Fuel Assemblies. (author)

  17. SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

    Science.gov (United States)

    Schluderberg, D.C.; Ryon, J.W.

    1962-05-01

    A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

  18. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

    Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)

  19. Nuclear policies: fuel without the bomb

    International Nuclear Information System (INIS)

    The essays, developed from studies conducted by the California seminar on arms control and foreign policy, address technical, political, and economic aspects of nonproliferation. How to halt nuclear proliferation commands worldwide attention today. The search for new energy resources by industrial as well as nonindustral nations has led to the spread of nuclear technology and the production of weapons grade fuel materials such as plutonium and enriched uranium in the name of energy independence. The background and consequences of this growing danger and possible solutions to it are the substance of the essays. Conceding the desirability (if not necessity) of developing nuclear power as an energy source, the writers focus on the different reactor technologies; an historical perspective of proliferation through the example of India; the rationales for stringent international monitoring; and finally, the link between proliferation and the spread of nuclear weapons. The chapters are: Nuclear technology: essential elements for decisionmakers, Robert Gillette; Must we decide now for worldwide commerce in plutonium fuel, Albert Wohlstetter; US peaceful aid and the Indian bomb, Roberta Wohlstetter; International discipline over the uses of nuclear energy, Victor Gilinsky; and Nuclear energy and the proliferation of nuclear weapons, Victor Gilinsky

  20. World Nuclear Association position statement: Safe management of nuclear waste and used nuclear fuel

    International Nuclear Information System (INIS)

    This WNA Position Statement summarises the worldwide nuclear industry's record, progress and plans in safely managing nuclear waste and used nuclear fuel. The global industry's safe waste management practices cover the entire nuclear fuel-cycle, from the mining of uranium to the long-term disposal of end products from nuclear power reactors. The Statement's aim is to provide, in clear and accurate terms, the nuclear industry's 'story' on a crucially important subject often clouded by misinformation. Inevitably, each country and each company employs a management strategy appropriate to a specific national and technical context. This Position Statement reflects a confident industry consensus that a common dedication to sound practices throughout the nuclear industry worldwide is continuing to enhance an already robust global record of safe management of nuclear waste and used nuclear fuel. This text focuses solely on modern civil programmes of nuclear-electricity generation. It does not deal with the substantial quantities of waste from military or early civil nuclear programmes. These wastes fall into the category of 'legacy activities' and are generally accepted as a responsibility of national governments. The clean-up of wastes resulting from 'legacy activities' should not be confused with the limited volume of end products that are routinely produced and safely managed by today's nuclear energy industry. On the significant subject of 'Decommissioning of Nuclear Facilities', which is integral to modern civil nuclear power programmes, the WNA will offer a separate Position Statement covering the industry's safe management of nuclear waste in this context. The paper's conclusion is that the safe management of nuclear waste and used nuclear fuel is a widespread, well-demonstrated reality. This strong safety record reflects a high degree of nuclear industry expertise and of industry responsibility toward the well-being of current and future generations. Accumulating

  1. Proceeding of the Fifth Scientific Presentation on Nuclear Fuel Cycle: Development of Nuclear Fuel Cycle Technology in Third Millennium

    International Nuclear Information System (INIS)

    The proceeding contains papers presented in the Fifth Scientific Presentation on Nuclear Fuel Element Cycle with theme of Development of Nuclear Fuel Cycle Technology in Third Millennium, held on 22 February in Jakarta, Indonesia. These papers were divided by three groups that are technology of exploration, processing, purification and analysis of nuclear materials; technology of nuclear fuel elements and structures; and technology of waste management, safety and management of nuclear fuel cycle. There are 35 papers indexed individually. (id)

  2. Nuclear fuel waste policy in Canada

    International Nuclear Information System (INIS)

    The 1996 Policy Framework for Radioactive Waste established the approach in Canada for dealing with all radioactive waste, and defined the respective roles of Government and waste producers and owners. The Policy Framework sets the stage for the development of institutional and financial arrangements to implement long-term waste management solutions in a safe, environmentally sound, comprehensive, cost-effective and integrated manner. For nuclear fuel waste, a 10-year environmental review of the concept to bury nuclear fuel waste bundles at a depth of 500 m to 1000 m in stable rock of the Canadian Shield was completed in March 1998. The Review Panel found that while the concept was technically safe, it did not have the required level of public acceptability to be adopted at this time as Canada's approach for managing its nuclear fuel waste. The Panel recommended that a Waste Management Organization be established at arm's length from the nuclear industry, entirely funded by the waste producers and owners, and that it be subject to oversight by the Government. In its December 1998 Response to the Review Panel, the Government of Canada provided policy direction for the next steps towards developing Canada's approach for the long-term management of nuclear fuel waste. The Government chose to maintain the responsibility for long-term management of nuclear fuel waste close with the producers and owners of the waste. This is consistent with its 1996 Policy Framework for Radioactive Waste. This approach is also consistent with experience in many countries. In addition, the federal government identified the need for credible federal oversight. Cabinet directed the Minister of NRCan to consult with stakeholders, including the public, and return to ministers within 12 months with recommendations on means to implement federal oversight. (author)

  3. Advances in HTGR fuel performance models

    International Nuclear Information System (INIS)

    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)

  4. Thorium nuclear fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  5. Securing the nuclear fuel cycle: What next?

    International Nuclear Information System (INIS)

    The greatest challenge to the international nuclear non-proliferation regime is posed by nuclear energy's dual nature for both peaceful and military purposes. Uranium enrichment and spent nuclear fuel (SNF) reprocessing (here after called sensitive nuclear technologies) are critical from the non-proliferation viewpoint because they may be used to produce weapons-grade nuclear materials: highly enriched uranium and separated plutonium. Alongside measures to limit the spread of sensitive nuclear technologies, multilateral approaches to the nuclear fuel cycle (NFC) started to be discussed. Spiralling prices for hydrocarbons and prospects of their imminent extinction are encouraging more and more countries to look at nuclear energy as an alternative means to ensure their sustainable development. To this end, it's becoming increasingly important to link the objective need for an expanded use of nuclear energy with strengthening nuclear non-proliferation by, in particular, preventing the spread of sensitive nuclear technologies and securing access for interested countries to NFC products and services. With this in mind, at the IAEA General Conference in 2003, IAEA Director General Mohamed ElBaradei called for establishing an international experts group on multilateral nuclear approaches. The proposal was supported, and in February 2005 the international experts, headed by Bruno Pellaud, issued a report (published by the IAEA as INFCIRC-640; see www.iaea.org) with recommendations on different multilateral approaches. The recommendations can be generalized as follows: reinforcement of existing market mechanisms; involvement of governments and the IAEA in the assurance of supply, including the establishment of low-enriched uranium (LEU) stocks as reserves; conversion of existing national uranium enrichment and SNF reprocessing enterprises into multilateral ones under international management and control, and setting up new multilateral enterprises on regional and

  6. International nuclear fuel cycle centers in global nuclear power infrastructure

    International Nuclear Information System (INIS)

    Interest among the nations of the world to use nuclear energy is increasing due to economic, environmental and energy security reasons. The increase in the number of nations using nuclear energy might raise political risk of non-peaceful use of sensitive nuclear technologies. Therefore, additional measures should be taken in order to minimize risk of proliferation in connection with the awaited renaissance of nuclear power. The problem of nuclear nonproliferation is an extremely complicated one and in order to mitigate it different dimensions should be taken into account: political, technological and institutional. Early in 2006 Russia proposed an initiative on global nuclear power infrastructure which will permit nondiscrimination access to nuclear energy of all interested countries observing requirements of nonproliferation regime. The key element of such infrastructure should be system of International Centers (IC) to provide services of nuclear fuel cycle including at first stage uranium enrichment and later on management of spent nuclear fuel (SNF) under the IAEA control. For effective management of SNF it is necessary to have developed technologies at least in four areas including fast reactors (FR) and closed fuel cycle technologies, SNF reprocessing, transuranium (TRU) fuel fabrication, nuclear waste management. At present the technology for only one area mentioned above have reached commercial level - LWR SNF aqueous reprocessing. Two other areas - technologies of sodium FR, MOX fuel for FR - have been demonstrated at semi-industrial level. Other technologies are still at R and D level - reprocessing of FR SNF, multi recycling of TRU fuel in FR, and nuclear waste management. Business as usual scenario of ICs establishment for SNF management might be to wait until some nations commercialize all associated with FR and closed fuel cycle areas driven mainly by national interest in addressing uranium resource shortages. Obviously this way needs significant time

  7. Nuclear Energy and Synthetic Liquid Transportation Fuels

    Science.gov (United States)

    McDonald, Richard

    2012-10-01

    This talk will propose a plan to combine nuclear reactors with the Fischer-Tropsch (F-T) process to produce synthetic carbon-neutral liquid transportation fuels from sea water. These fuels can be formed from the hydrogen and carbon dioxide in sea water and will burn to water and carbon dioxide in a cycle powered by nuclear reactors. The F-T process was developed nearly 100 years ago as a method of synthesizing liquid fuels from coal. This process presently provides commercial liquid fuels in South Africa, Malaysia, and Qatar, mainly using natural gas as a feedstock. Nuclear energy can be used to separate water into hydrogen and oxygen as well as to extract carbon dioxide from sea water using ion exchange technology. The carbon dioxide and hydrogen react to form synthesis gas, the mixture needed at the beginning of the F-T process. Following further refining, the products, typically diesel and Jet-A, can use existing infrastructure and can power conventional engines with little or no modification. We can then use these carbon-neutral liquid fuels conveniently long into the future with few adverse environmental impacts.

  8. Nuclear Propulsion using Thin Foiled Fuel

    Science.gov (United States)

    Takahashi, H.

    1998-11-01

    A new way to produce plasma for nuclear propulsion is proposed. A thin foiled fuel can be used for converting fission energy to propulsion energy efficiently. The fission products coming out of the thin foil directly ionize the hydrogen molecules which are used for propulsion. Thus very small portion of fission energy deposited in the thin foil, and integrity of the thin foiled fuel can be maintained even in high nuclear power. Fuel material with large thermal fission cross-section is preferable to make thin foiled fuel and the heat deposition in the foil can be reduced. To get high power from the foiled fuel assembly, thermal neutrons which are created out from the assembly can be supplied, or the assembly itself can create the high intensity thermal neutrons by self-multiplication. A flexible design of a highly efficient nuclear propulsion system can be made. The thickness of the foil and the maintenance of the thermo-mechanical integrity can be determined from the fission cross-section and the slowing down power for fission products. The talk discusses the issues related to heat removal from the assembly.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-01

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

  11. Spent Nuclear Fuel Storage Program user's guide

    International Nuclear Information System (INIS)

    The purpose of this manual is to present procedures to execute the Spent Nuclear Fuel Storage Model (SNFSM) program. This manual includes an overview of the model, operating environment, input and output specifications and user procedures. An example of the execution of the program is included to assist potential users

  12. Decision Analysis For Nuclear Fuel Cycle Policy

    International Nuclear Information System (INIS)

    The prime objective in this talk is to explore the impact of widely different (or hypothetical) fuel cycle requirement rather than to attempt to predict a probable scenario. In the course of preparation of this talk, it was realized that, despite the very speculative nature of this kind of endeavor, studies like these are considered essential to the long-range planning needs of the national nuclear power industry, utilities and those providing supporting services, even though the current presentation are extremely primitive in that purpose. A nuclear electricity utility tries to reduce fuel cycle costs. But the problems have to be approached with a long-term perspective, and the logical conclusion is that utility has to make technical progress. As nuclear generation gradually become great, supplies of the fuel cycle services are responsible for the R and D about the nuclear fuel cycle services which is useful to implement the technical choices they propose. Then it is for the utility to choose according to his knowledge, if necessary by carrying out additional research. But only the utility acquires real operating experience and prototype reactor or laboratory tests offer limited knowledge quantities. One way to ensure a good guarantee of supply is, obviously, to make the order far enough ahead of time to have a stock. But, on the other hand, stocks are expensive and should be kept to a strict minimum. Therefore, a detailed analysis of uncertainties is required, as well as an effort to optimize the handling of the overall problem. As mentioned earlier, in recent years, specifically the right way to handle the back-end of the fuel cycle has been always hotly contested and ultimately it was a question of reprocessing or direct disposal of spent fuel elements. Direct disposal of spent fuel is, at present, the only possibility of spent fuel disposal option available to the Korean utility. Korea, having virtually no indigenous uranium resources, can hardly afford to

  13. Multidimensional multiphysics simulation of nuclear fuel behavior

    Science.gov (United States)

    Williamson, R. L.; Hales, J. D.; Novascone, S. R.; Tonks, M. R.; Gaston, D. R.; Permann, C. J.; Andrs, D.; Martineau, R. C.

    2012-04-01

    Nuclear fuel operates in an environment that induces complex multiphysics phenomena, occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. This multiphysics behavior is often tightly coupled and many important aspects are inherently multidimensional. Most current fuel modeling codes employ loose multiphysics coupling and are restricted to 2D axisymmetric or 1.5D approximations. This paper describes a new modeling tool able to simulate coupled multiphysics and multiscale fuel behavior, for either 2D axisymmetric or 3D geometries. Specific fuel analysis capabilities currently implemented in this tool are described, followed by a set of demonstration problems which include a 10-pellet light water reactor fuel rodlet, three-dimensional analysis of pellet clad mechanical interaction in the vicinity of a defective fuel pellet, coupled heat transfer and fission product diffusion in a TRISO-coated fuel particle, a demonstration of the ability to couple to lower-length scale models to account for material property variation with microstructural evolution, and a demonstration of the tool's ability to efficiently solve very large and complex problems using massively-parallel computing. A final section describes an early validation exercise, comparing simulation results to a light water reactor fuel rod experiment.

  14. Method for inspecting nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A technique for disassembling a nuclear reactor fuel element without destroying the individual fuel pins and other structural components from which the element is assembled is described. A traveling bridge and trolley span a water-filled spent fuel storage pool and support a strongback. The strongback is under water and provides a working surface on which the spent fuel element is placed for inspection and for the manipulation that is associated with disassembly and assembly. To remove, in a non-destructive manner, the grids that hold the fuel pins in the proper relative positions within the element, bars are inserted through apertures in the grids with the aid of special tools. These bars are rotated to flex the adjacent grid walls and, in this way relax the physical engagement between protruding portions of the grid walls and the associated fuel pins. With the grid structure so flexed to relax the physical grip on the individual fuel pins, these pins can be withdrawn for inspection or replacement as necessary without imposing a need to destroy fuel element components

  15. MOX technology for new nuclear fuel fabrication

    International Nuclear Information System (INIS)

    The new nuclear fabrication plant MELOX, at Marcoule in the south of France is the first commercial sized plant to supply a new market for mixed uranium/plutonium oxide (MOX) fuel which will allow the plutonium separated by reprocessing to be recycled profitably in light water reactors. An eighteen-month programme of commissioning has started at the site following completion of construction work in the summer of 1993. The programme envisages completion of testing and certification at the end of 1994, production of about 50 tonnes (heavy metal) of MOX fuel in the first year of commercial operation and achievement of the full capacity of 120 tonnes in 1996. The MELOX plant is described. It has been built to high seismic standards and has been expensive to build. However, there is a demand for mixed oxide fuel assemblies for the French nuclear programme. (Author)

  16. Spent Nuclear Fuel Alternative Technology Decision Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, C.B.

    1999-11-29

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

  17. Summary of nuclear fuel reprocessing activities around the world

    International Nuclear Information System (INIS)

    This review of international practices for nuclear fuel reprocessing was prepared to provide a nontechnical summary of the current status of nuclear fuel reprocessing activities around the world. The sources of information are widely varied

  18. Summary of nuclear fuel reprocessing activities around the world

    Energy Technology Data Exchange (ETDEWEB)

    Mellinger, P.J.; Harmon, K.M.; Lakey, L.T.

    1984-11-01

    This review of international practices for nuclear fuel reprocessing was prepared to provide a nontechnical summary of the current status of nuclear fuel reprocessing activities around the world. The sources of information are widely varied.

  19. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1970-01-01

    Advances in Nuclear Science and Technology, Volume 5 presents the underlying principles and theory, as well as the practical applications of the advances in the nuclear field. This book reviews the specialized applications to such fields as space propulsion.Organized into six chapters, this volume begins with an overview of the design and objective of the Fast Flux Test Facility to provide fast flux irradiation testing facilities. This text then examines the problem in the design of nuclear reactors, which is the analysis of the spatial and temporal behavior of the neutron and temperature dist

  20. Advances in nuclear science and technology

    CERN Document Server

    Greebler, Paul

    1966-01-01

    Advances in Nuclear Science and Technology, Volume 3 provides an authoritative, complete, coherent, and critical review of the nuclear industry. This book presents the advances in the atomic energy field.Organized into six chapters, this volume begins with an overview of the use of pulsed neutron sources for the determination of the thermalization and diffusion properties of moderating as well as multiplying media. This text then examines the effect of nuclear radiation on electronic circuitry and its components. Other chapters consider radiation effects in various inorganic solids, with empha

  1. International guidelines for fire protection at nuclear installations including nuclear fuel plants, nuclear fuel stores, teaching reactors, research establishments

    International Nuclear Information System (INIS)

    The guidelines are recommended to designers, constructors, operators and insurers of nuclear fuel plants and other facilities using significant quantities of radioactive materials including research and teaching reactor installations where the reactors generally operate at less than approximately 10 MW(th). Recommendations for elementary precautions against fire risk at nuclear installations are followed by appendices on more specific topics. These cover: fire protection management and organization; precautions against loss during construction alterations and maintenance; basic fire protection for nuclear fuel plants; storage and nuclear fuel; and basic fire protection for research and training establishments. There are numerous illustrations of facilities referred to in the text. (U.K.)

  2. ADVANCED CERAMIC MATERIALS FOR NEXT-GENERATION NUCLEAR APPLICATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Marra, J.

    2010-09-29

    , and proliferation), the worldwide community is working to develop and deploy new nuclear energy systems and advanced fuel cycles. These new nuclear systems address the key challenges and include: (1) extracting the full energy value of the nuclear fuel; (2) creating waste solutions with improved long term safety; (3) minimizing the potential for the misuse of the technology and materials for weapons; (4) continually improving the safety of nuclear energy systems; and (5) keeping the cost of energy affordable.

  3. Economic viability of innovative nuclear reactor and fuel cycle technologies

    International Nuclear Information System (INIS)

    Full text: Nuclear power has established its position as one of the most stable electricity supply sources in many countries in the world, supplying about 17% of total electricity generated. However, in order to keep that position, there are two important challenges that nuclear energy will face in the coming decades. They are: competition, and social/political acceptance (including non-proliferation and terrorism). There is an increasing concern that existing nuclear technologies may not be able to overcome such tough challenges. It is expected that innovative technologies can be a part of the solutions to overcome such challenges. This paper focuses on economic viability of innovative nuclear reactor and its associated fuel cycle technologies. First, it is important to consider the long term energy paths and potential role of nuclear power under different scenarios. We applied global energy optimization model based on IPCC scenarios. Then, we look at Japan, where electricity market is being liberalized, in order to explore how liberalization will have influence economic viability of nuclear power. The following are our basic conclusions: CO2 constraints as well as power generation cost competitiveness could affect future growth of nuclear power quite significantly. Current trend suggests that nuclear power would not grow much without CO2 constraints, or even face minus growth if its power generation cost became higher. On the other hand, cost reduction with CO2 constraints could accelerate future expansion of nuclear power quite significantly; In addition to life-long average generation cost, other investment criteria (such as asset productivity) may become critically important under the liberalized market. Under the liberalized electricity market, short term investment criteria could become more important than 30 year life time average cost. This suggests that small initial investment is more acceptable than large capital investment. Advanced nuclear reactor

  4. Nuclear reactor fuel element. Kernreaktorbrennelement

    Energy Technology Data Exchange (ETDEWEB)

    Lippert, H.J.

    1985-03-28

    The fuel element box for a BWR is situated with a corner bolt on the inside in one corner of its top on the top side of the top plate. This corner bolt is screwed down with a bolt with a corner part which is provided with leaf springs outside on two sides, where the bolt has a smaller diameter and an expansion shank. The bolt is held captive to the bolt head on the top and the holder on the bottom of the corner part. The holder is a locknut. If the expansion forces are too great, the bolt can only break at the expansion shank.

  5. World nuclear fuel cycle requirements, 1984

    International Nuclear Information System (INIS)

    This report presents projections of the domestic and foreign requirements for uranium and enrichment services, as well as spent nuclear fuel discharges. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity published in a recent Energy Information Administration (EIA) report. Four scenarios (high, middle, low, and no new reactor orders) are included for domestic nuclear power capacity and three (high, middle, and low) for countries in the World Outside Planned Economies (WOCA). In addition, 4 sensitivity cases are presented for the US lower capacity factors, reactor aging, lower tails assay, and higher burnup. Six sensitivity cases are analyzed for the WOCA countries: (1) stable, instead of improving, capacity factors for the United States and for countries in the Other country group; (2) reactor aging; (3) recycling of uranium but not plutonium from spent fuel (the three standard scenarios assume recycling of both uranium and plutonium; (4) no recycling of spent fuels; (5) lower uranium enrichment tails assay; and (6) higher fuel burnup levels. The annual US requirements for uranium and for uranium enrichment service are projected to more than double between 1985 and 2020 in the middle case, and the cumulative amount of spent fuel discharged is projected to increase approximately 10-fold. Annual uranium requirements for the WOCA nations are projected to increase by about 60% between 1985 and 2000. In contrast, a 7- to 8-fold increase in U3O8 and enrichment service requirements is projected for the Other WOCA country group during this time period, as its relatively small existing nuclear power capacity undergoes rapid expansion

  6. A present status for dry storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bang, K. S.; Lee, J. C.; Park, H. Y.; Seo, K. S

    2003-04-01

    National policy for management of a spent nuclear fuel does not establish in Korea yet. A storage capacity of a storage pool that is to store the spent nuclear fuel will be exceeded an amount of accumulation from the first Woljin nuclear power plant in 2007. Therefore it is necessary that dry storage facility is secured to store safely the spent nuclear fuel on site of the nuclear power plant until national policy for a back-end spent nuclear fuel cycle is established. In order to store safely spent nuclear fuel, it is important that the present status and technology on dry storage of spent nuclear fuel is looked over. Therefore, the present status on dry storage of spent nuclear fuel was analyzed so as to develop dry storage system and choose a proper dry storage method domestic.

  7. Robot vision for nuclear advanced robot

    International Nuclear Information System (INIS)

    This paper describes Robot Vision and Operation System for Nuclear Advanced Robot. This Robot Vision consists of robot position detection, obstacle detection and object recognition. With these vision techniques, a mobile robot can make a path and move autonomously along the planned path. The authors implemented the above robot vision system on the 'Advanced Robot for Nuclear Power Plant' and tested in an environment mocked up as nuclear power plant facilities. Since the operation system for this robot consists of operator's console and a large stereo monitor, this system can be easily operated by one person. Experimental tests were made using the Advanced Robot (nuclear robot). Results indicate that the proposed operation system is very useful, and can be operate by only person. (author)

  8. Process for reprocessing a nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    In order to separate the nuclear fuel from the can material, the can is heated evenly in the gastight closed state together with the nuclear fuel contained in it, so that the diameter of the can expands, increasing the gap between the nuclear fuel and the can without cracks occurring in the can. The expanded can is then opened at one end and finally the nuclear fuel from the opened can and is treated separately from the can. (orig./HP)

  9. HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER

    Energy Technology Data Exchange (ETDEWEB)

    BROWN,LC; BESENBRUCH,GE; LENTSCH,RD; SCHULTZ,KR; FUNK,JF; PICKARD,PS; MARSHALL,AC; SHOWALTER,SK

    2003-06-01

    OAK B202 HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER. Combustion of fossil fuels, used to power transportation, generate electricity, heat homes and fuel industry provides 86% of the world's energy. Drawbacks to fossil fuel utilization include limited supply, pollution, and carbon dioxide emissions. Carbon dioxide emissions, thought to be responsible for global warming, are now the subject of international treaties. Together, these drawbacks argue for the replacement of fossil fuels with a less-polluting potentially renewable primary energy such as nuclear energy. Conventional nuclear plants readily generate electric power but fossil fuels are firmly entrenched in the transportation sector. Hydrogen is an environmentally attractive transportation fuel that has the potential to displace fossil fuels. Hydrogen will be particularly advantageous when coupled with fuel cells. Fuel cells have higher efficiency than conventional battery/internal combustion engine combinations and do not produce nitrogen oxides during low-temperature operation. Contemporary hydrogen production is primarily based on fossil fuels and most specifically on natural gas. When hydrogen is produced using energy derived from fossil fuels, there is little or no environmental advantage. There is currently no large scale, cost-effective, environmentally attractive hydrogen production process available for commercialization, nor has such a process been identified. The objective of this work is to find an economically feasible process for the production of hydrogen, by nuclear means, using an advanced high-temperature nuclear reactor as the primary energy source. Hydrogen production by thermochemical water-splitting (Appendix A), a chemical process that accomplishes the decomposition of water into hydrogen and oxygen using only heat or, in the case of a hybrid thermochemical process, by a combination of heat and electrolysis, could meet these goals. Hydrogen produced from

  10. The U.S. Advanced Fuel Cycle Initiative: Development of separations technologies

    International Nuclear Information System (INIS)

    Spent nuclear fuel from 103 operating U.S. commercial nuclear power reactors is accumulating at a rate of about 2,000 metric tons per year. At this rate, the legislated capacity of the Yucca Mountain geologic repository (63,000 tons of commercial spent fuel) will be exceeded by 2015. Accordingly, the U.S. Department of Energy has instituted a new program, the Advanced Fuel Cycle Initiative, which is intended to provide the technologies necessary for the economical and environmentally sound processing of spent fuel. The goal of this technology development program is to preclude or significantly delay the need for a second geologic repository. Separations technologies are being developed that will support the processing of commercial spent fuel as well as the spent fuel arising from the operation of future advanced reactors

  11. PHWR advanced fuel R and D for the 21st Century in Korea

    International Nuclear Information System (INIS)

    As the first CANDU-PHWR nuclear power plant in Korea, Wolsong Unit 1 has been successfully operated since 1983. The CANDU installed electric-generation capacity was about 50 % of the installed electric-generation capacity of nuclear power plants in Korea in 1983 but then decreased to less than 10 % of the total installed nuclear electric-generation capacity by 1996. This CANDU installed electric-generation capacity has recovered to about 20 % of the total installed nuclear electric-generation capacity in 1999, because Wolsong Units 2, 3 and 4 have been placed into commercial operation in 1997, 1998 and 1999, respectively. This indicates that CANDU reactors are not the majority of nuclear power plants in Korea. Since the period of the late 1970s, nuclear fuel design and fabrication technologies have been engaged as one of the important R and D activities in Korea. As one of the early R and D activities leading to nuclear power industrialization in Korea, the project to develop the design and fabrication technology of CANDU-6 37-element fuel had been successfully carried out from 1981 to 1987 by KAERI. Just after the successful completion of the 37-element fuel R and D, KAERI has developed a CANDU-6 advanced fuel. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology. The CANFLEX and DUPIC R and D programs have been conducted under Korea's Nuclear Energy R and D Project as national mid- and long-term programs since 1992. As the second of the CANDU R and D products in Korea, the CANFLEX-NU fuel has been jointly developed by KAERI and AECL.The fuel has demonstrated its irradiation performance in a Canadian commercial power reactor, Pt. Lepreau Generating Station since 1998 September. The RU(SEU) and DUPIC fuels are expected to be developed continuously until about the year 2010 for their use in CANDU reactors. Beside these fuel

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

    International Nuclear Information System (INIS)

    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

  13. Systematic analysis of advanced fusion fuel in inertial fusion energy

    Science.gov (United States)

    Velarde, G.; Eliezer, S.; Henis, Z.; Piera, M.; Martinez-Val, J. M.

    1997-04-01

    Aneutronic fusion reactions can be considered as the cleanest way to exploit nuclear energy. However, these reactions present in general two main drawbacks.—very high temperatures are needed to reach relevant values of their cross sections—Moderate (and even low) energy yield per reaction. This value is still lower if measured in relation to the Z number of the reacting particles. It is already known that bremsstrahlung overruns the plasma reheating by fusion born charged-particles in most of the advanced fuels. This is for instance the case for proton-boron-11 fusion in a stoichiometric plasma and is also so in lithium isotopes fusion reactions. In this paper, the use of deuterium-tritium seeding is suggested to allow to reach higher burnup fractions of advanced fuels, starting at a lower ignition temperature. Of course, neutron production increases as DT contents does. Nevertheless, the ratio of neutron production to energy generation is much lower in DT-advanced fuel mixtures than in pure DT plasmas. One of the main findings of this work is that some natural resources (as D and Li-7) can be burned-up in a catalytic regime for tritium. In this case, neither external tritium breeding nor tritium storage are needed, because the tritium inventory after the fusion burst is the same as before it. The fusion reactor can thus operate on a pure recycling of a small tritium inventory.

  14. Supply Security in Future Nuclear Fuel Markets

    International Nuclear Information System (INIS)

    Previous PNNL work has shown the existing nuclear fuel markets to provide a high degree of supply security, including the ability to respond to supply disruptions that occur for technical and non-technical reasons. It is in the context of new reactor designs - that is, reactors likely to be licensed and market ready over the next several decades - that fuel supply security is most relevant. Whereas the fuel design and fabrication technology for existing reactors are well known, the construction of a new set of reactors could stress the ability of the existing market to provide adequate supply redundancy. This study shows this is unlikely to occur for at least thirty years, as most reactors likely to be built in the next three decades will be evolutions of current designs, with similar fuel designs to existing reactors.

  15. Supply Security in Future Nuclear Fuel Markets

    Energy Technology Data Exchange (ETDEWEB)

    Seward, Amy M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wood, Thomas W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gitau, Ernest T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ford, Benjamin E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-11-18

    Previous PNNL work has shown the existing nuclear fuel markets to provide a high degree of supply security, including the ability to respond to supply disruptions that occur for technical and non-technical reasons. It is in the context of new reactor designs – that is, reactors likely to be licensed and market ready over the next several decades – that fuel supply security is most relevant. Whereas the fuel design and fabrication technology for existing reactors are well known, the construction of a new set of reactors could stress the ability of the existing market to provide adequate supply redundancy. This study shows this is unlikely to occur for at least thirty years, as most reactors likely to be built in the next three decades will be evolutions of current designs, with similar fuel designs to existing reactors.

  16. Practical public acceptance activities in Japan Nuclear Fuel Company

    International Nuclear Information System (INIS)

    JNF PA is characterized by 3 symbolic 'F's if concisely expressed. The first F comes from the Focus F, which stands for JNF's focused or customized attendance to what visitors want. The second F from Friendly F, which symbolizes JNF's simple and easy presentations in an amenable atmosphere without use of specialist language. The last F from First Hand, which means the visitors given the chance to experience themselves in actual touch with uranium. Our nuclear fuel manufacturing facility is one of the limited spots for such an experience. Their encounter with this 'actual reality' is worth a millions of words. Many citizens of Yokosuka have been invited to our facility at every summer holiday season for the past 6 years. 60% of them answered to our survey: 'Yes, I came here with fear in nuclear angry' but 75% admitted 'I now feel easy with it' after they listened to the issues of energy and environment, watched the animated nuclear fuel cycle and toured through our nuclear fuel manufacturing plant. This is an justified encouragement to what we devote in PA. What we should do is two fold. One is to address our PA to younger generations. Another is to support to those PA activities at newly-planned sites for nuclear power generation. For the former case, we now are approaching educational and tutorial personnel to provide pupils with opportunity of touring our facilities. For the latter, we sincerely endeavor to tune up to the visitors from such site areas, with elaboration based on advance information, rather than mere briefings. Like the Japanese tea ceremonial spirit of 'we happen to meet but once for all', why not fulfilling our role from the spot of a nuclear fuel manufacturing facility? (J.P.N.)

  17. Nuclear fuel: sustainable source of energy or burden on society?

    International Nuclear Information System (INIS)

    In the past, the question concerning the sustainability of a resource primarily addressed its finite nature. Accordingly, electricity production using renewable energies was clearly sustainable. Contrasting this are systems based on oil, gas, coal or uranium. However, from the perspective of 'neo-sustainability' being analyzed today, this assessment appears less clear-cut, especially in light of the definition of sustainability as provided by the Brundtland report. Nowadays, the depletion time of fuel resources is thus not the only significant aspect, but factors such as efficiency, ecofriendliness and social responsibility also figure in. The nuclear fuel supply is analyzed from a sustainability perspective. After a short description of the supply chain, each of the most important aspects of sustainability are related to the individual stages of the supply chain and evaluated. This method aims at answering the question concerning to what extent nuclear fuel is a sustainable source of energy. Although the recycling of fissile materials from reprocessing and the deployment of advanced reactors are key factors as regards the issue of sustainability, these topics are deliberately only touched on. The main focus lies on the sustainability of the nuclear fuel cycle as it is currently utilized in light water reactors, without discussing the subject of reprocessing. (orig.)

  18. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1976-01-01

    Advances in Nuclear Science and Technology, Volume 9 provides information pertinent to the fundamental aspects of nuclear science and technology. This book discusses the safe and beneficial development of land-based nuclear power plants.Organized into five chapters, this volume begins with an overview of the possible consequences of a large-scale release of radioactivity from a nuclear reactor in the event of a serious accident. This text then discusses the extension of conventional perturbation techniques to multidimensional systems and to high-order approximations of the Boltzmann equation.

  19. Antineutrino monitoring of spent nuclear fuel

    CERN Document Server

    Brdar, Vedran; Kopp, Joachim

    2016-01-01

    Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries world wide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this letter, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel elements as a function of time, and we illustrate the usefulness of antineutrino detectors in several benchmark scenarios. In particular, we demonstrate how a measurement of the antineutrino flux can help to re-verify the contents of a dry storage cask in case the monitoring chain by conventional means gets disrupted. We then comment on the usefulness of antineutrino detectors at long-term storage facilities such as Yucca mountain. Finally, we put forward antineutrino detection as a tool in locating underground "hot spots" in ...

  20. Technology development of nuclear material safeguards for DUPIC fuel cycle

    International Nuclear Information System (INIS)

    During the second phase of research and development program conducted from 1993 to 1996, nuclear material safeguards studies system were performed on the technology development of DUPIC safeguards system such as nuclear material measurement in bulk form and product form, DUPIC fuel reactivity measurement, near-real-time accountancy, and containment and surveillance system for effective and efficient implementation of domestic and international safeguards obligation. By securing in advance a optimized safeguards system with domestically developed hardware and software, it will contribute not only to the effective implementation of DUPIC safeguards, but also to enhance the international confidence build-up in peaceful use of spent fuel material. (author). 27 refs., 13 tabs., 89 figs

  1. Technology development of nuclear material safeguards for DUPIC fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jong Sook; Kim, Ho Dong; Kang, Hee Young; Lee, Young Gil; Byeon, Kee Ho; Park, Young Soo; Cha, Hong Ryul; Park, Ho Joon; Lee, Byung Doo; Chung, Sang Tae; Choi, Hyung Rae; Park, Hyun Soo

    1997-07-01

    During the second phase of research and development program conducted from 1993 to 1996, nuclear material safeguards studies system were performed on the technology development of DUPIC safeguards system such as nuclear material measurement in bulk form and product form, DUPIC fuel reactivity measurement, near-real-time accountancy, and containment and surveillance system for effective and efficient implementation of domestic and international safeguards obligation. By securing in advance a optimized safeguards system with domestically developed hardware and software, it will contribute not only to the effective implementation of DUPIC safeguards, but also to enhance the international confidence build-up in peaceful use of spent fuel material. (author). 27 refs., 13 tabs., 89 figs.

  2. Method and apparatus for increasing fuel efficiency in nuclear reactors

    International Nuclear Information System (INIS)

    This patent describes an improved method of producing a spectral shift in a nuclear reactor to achieve increased nuclear fuel efficiency, the nuclear reactor containing a fluid moderator juxtaposed with fuel elements containing the nuclear fuel, which comprises disposing within the fluid moderator stationary non-poison displacer rods for achieving the spectral shift, the displacer rods exhibiting a continuous reduction in volume during operation of the nuclear reactor whereby the fluid moderator increases in volume as the nuclear fuel is burned in the nuclear reactor

  3. Ultrasonic decontamination of nuclear fuel. Feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Berg, A.; Libal, A.; Norbaeck, J.; Wegemar, B.

    1995-05-01

    Ultrasonic decontamination of nuclear fuel is an expeditious way to reduce radiation exposures resulting in a minimal volume of waste. The fuel assemblies are set up in the fuel preparation machine one at a time and treated without prior disassemblage. By decontaminating 20% of the BWR fuel assemblies annually, there is a potential to reduce the collective dose by approximately 40-50%. Including also improved reactivity of the fuel, this amounts to an economic benefit of about 4 MSEK per reactor and year. The costs for performing the decontamination can be economically justified if the plants do not plan for short outages each year. The decontamination method could also be used for the purpose of removing tramp Uranium following a fuel failure or minor core accident. An additional benefit is removal of loosely adherent crud. The waste produced will be handled in a closed filtering circuit. The method is suggested to be verified in a test on discharged burnt-up fuel at site. The next step will be to develop the method further in order to be able to remove also tenacious crud. 12 refs, 4 tabs.

  4. Development of nuclear fuel for integrated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

  5. Ultrasonic decontamination of nuclear fuel. Feasibility study

    International Nuclear Information System (INIS)

    Ultrasonic decontamination of nuclear fuel is an expeditious way to reduce radiation exposures resulting in a minimal volume of waste. The fuel assemblies are set up in the fuel preparation machine one at a time and treated without prior disassemblage. By decontaminating 20% of the BWR fuel assemblies annually, there is a potential to reduce the collective dose by approximately 40-50%. Including also improved reactivity of the fuel, this amounts to an economic benefit of about 4 MSEK per reactor and year. The costs for performing the decontamination can be economically justified if the plants do not plan for short outages each year. The decontamination method could also be used for the purpose of removing tramp Uranium following a fuel failure or minor core accident. An additional benefit is removal of loosely adherent crud. The waste produced will be handled in a closed filtering circuit. The method is suggested to be verified in a test on discharged burnt-up fuel at site. The next step will be to develop the method further in order to be able to remove also tenacious crud. 12 refs, 4 tabs

  6. Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options.

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J. (Idaho National Laboratory, Idaho Falls, ID); Parma, Edward J., Jr.; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

    2007-10-01

    The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents.

  7. Nuclear power generation and fuel cycle report 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

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

  8. Comparison of different nuclear fuel cycles for LWR applications

    OpenAIRE

    Winblad von Walter, Tobias

    2008-01-01

    Nuclear power is considered a vital energy source, without greenhouse gas emissions, regarding the commitment towards sustainable energy systems. This is especially on the topic of the present climate debate. A central aspect of nuclear power is nuclear fuel. Presently Uranium dioxide (UOX) is the most common nuclear fuel in the world. However, an increased uranium price, waste and proliferation issues are some of the aspects that have resulted in a growing interest for other nuclear fuels. M...

  9. Nuclear power generation and fuel cycle report 1996

    International Nuclear Information System (INIS)

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

  10. Environmental management in Framatome nuclear fuel

    International Nuclear Information System (INIS)

    Environmental preservation is both a national regulatory requirement and a condition for economic and social development. The various industrial sites belonging to the Framatome Nuclear Fuel Organisation, whose activities range from the processing and transformation of Zirconium alloy products to the fabrication of fuel assemblies, have always demonstrated that protection of the environment was their prime concern by implementing low pollution level processes and reducing and/or recycling industrial waste and effluents. As early as January 1996, a directive issued by the Framatome Group defined its environmental policy and responsibilities in the matter. Within the Framatome Nuclear Fuel Organization, this directive has been applied by implementation of: low level pollution processes; better performance of recycling of effluents, by-products and waste; environmental information policy. In all its plants, the Framatome Nuclear Fuel Organization has decided to pursue and to step up its environmental protection policy by: officializing its action through compliance with ISO standard 14001 and certification of all its industrial sites by 2001 at the latest; launching new actions and extra investment programs. In this context, FBFC has applied for a modification of the decrees concerning the dumping of liquid and gas effluents at the Romans factory. (authors)

  11. Advanced technologies for power and fuel production

    Energy Technology Data Exchange (ETDEWEB)

    Watts, J.U.; Mann, A.N. [US Department of Energy/National Energy Technology Lab., Pittsburgh, PA (United States)

    2001-07-01

    The Clean Coal Technology Program (CCT) being conducted by the United States Department of Energy (DOE) is a government and industry co-funded effort. The program's purpose is to demonstrate new generation of innovative, environmentally friendly processes that enhance the utilization of coal to meet increasing demand for electric power and fuels. Program demonstration areas include environmental control, advanced power generation, fuels production, and industrial applications. The CCT Program has now grown to maturity, with over 50% of the projects selected having successfully completed their demonstration goals and objectives. Under the CCT Program, nine advanced electric power generation projects and five coal processing for clean fuels projects were selected for full scale commercial demonstration. This paper provides the status, accomplishments and results of the most widely accepted technologies currently being commercialized under these two categories. The projects are (1) Atmospheric Fluidized-Bed Combustion (AFBC) at Jacksonville Electric Authority; (2) Integrated Gasification Combined-cycle (IGCC) at Wabash River, Tampa Electric and Kentucky Pioneer; and (3) Eastman Chemical's production of methanol via coal gasification using the LPMEOH{trademark} process. 7 figs., 7 tabs.

  12. IAEA nuclear fuel cycle databases: Relevance to spent nuclear fuel management

    International Nuclear Information System (INIS)

    Full text: Reliable statistical data on spent fuel management would be essential for the global nuclear community, e.g. for approaches related to international cooperation, as well as for the needs of individual countries. Compilation of data on large amounts of spent fuel located at various nuclear facilities around the world is a challenge. It is not a trivial exercise to collect and compile spent fuel inventory data as they are subject to dynamic change. Spent fuel inventory data are important to various national and international spent fuel management activities, especially for planning and regulatory activities. Recently, security issues became an additional factor to be considered in the information management associated with spent fuel or radioactive waste. The specific need for spent fuel inventory data varies depending on the ultimate purpose: International Level - compilation on a gross tonnage (in heavy metal basis) mainly for statistical purposes and global trend analysis both for use by IAEA and at the request of Member States; National Level - compilation for industry and regulatory purposes on either a gross tonnage or individual assembly basis to assist in planning and public awareness; and Operator Level - the origination and maintenance of detailed data on individual assemblies by the utility for operational needs or to meet regulatory requirements. There is, in general, a global trend towards greater transparency of information with the general public which may require more information to be made public on spent fuel management, including data on inventories or transportation. With the increase in the commercialisation of the nuclear industry, the trend is away from national governments operating nuclear facilities, including spent fuel management. This results in the spread of information on spent fuel as it is not concentrated at government level, but is instead held by various organizations . Spent fuel information may also have to be

  13. Outline of policy related to nuclear fuel cycle in Japan

    International Nuclear Information System (INIS)

    In Japan where energy resources are scarce, and energy supply structure is weak, the development and adoption of the substitute energy for petroleum are indispensable for the stable supply of energy. Atomic energy is excellent in its stability of supply and economy and is effective for earth environment problems. Nuclear fuel cycle is composed of the front end, namely ore mining, refining, conversion, enrichment, reconversion and fuel fabrication, and the back end such as the reprocessing of spent fuel and the treatment and disposal of radioactive wastes, and when it is established in Japan and the plutonium and recovered uranium obtained by reprocessing becomes usable, atomic energy becomes semi-home produced energy, and it is very important for ensuring the energy security and effectively utilizing resources in Japan. The present status of the nuclear fuel cycle in Japan is reported. The plan of constructing the facilities for uranium enrichment, spent fuel reprocessing and low level radioactive waste disposal in Ogawara industrial development district, Rokkasho Village, Aomori Prefecture, is advanced. As the measures for supporting the construction, the Ministry of International Trade and Industry carries out the promotion of district development, the promotion of PA measures, the ensuring of construction fund, the promotion of technical development and so on. (K.I.)

  14. Overview of the CANDU fuel handling system for advanced fuel cycles

    International Nuclear Information System (INIS)

    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)

  15. Nuclear fuels policy. Report of the Atlantic Council's Nuclear Fuels Policy Working Group

    International Nuclear Information System (INIS)

    This Policy Paper recommends the actions deemed necessary to assure that future U.S. and non-Communist countries' nuclear fuels supply will be adequate, considering the following: estimates of modest growth in overall energy demand, electrical energy demand, and nuclear electrical energy demand in the U.S. and abroad, predicated upon the continuing trends involving conservation of energy, increased use of electricity, and moderate economic growth (Chap. I); possibilities for the development and use of all domestic resources providing energy alternatives to imported oil and gas, consonant with current environmental, health, and safety concerns (Chap. II); assessment of the traditional energy sources which provide current alternatives to nuclear energy (Chap. II); evaluation of realistic expectations for additional future energy supplies from prospective technologies: enhanced recovery from traditional sources and development and use of oil shales and synthetic fuels from coal, fusion and solar energy (Chap. II); an accounting of established nuclear technology in use today, in particular the light water reactor, used for generating electricity (Chap. III); an estimate of future nuclear technology, in particular the prospective fast breeder (Chap. IV); current and projected nuclear fuel demand and supply in the U.S. and abroad (Chaps. V and VI); the constraints encountered today in meeting nuclear fuels demand (Chap. VII); and the major unresolved issues and options in nuclear fuels supply and use (Chap. VIII). The principal conclusions and recommendations (Chap. IX) are that the U.S. and other industrialized countries should strive for increased flexibility of primary energy fuel sources, and that a balanced energy strategy therefore depends on the secure supply of energy resources and the ability to substitute one form of fuel for another

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  17. DUPIC nuclear fuel manufacturing and process technology development at KAERI

    International Nuclear Information System (INIS)

    DUPIC fuel cycle development project in KAERI of Korea was initiated in 1991 and has advanced in relevant technologies for last 10 years. The project includes five different topics such as nuclear fuel manufacturing, compatibility evaluation, performance evaluation, manufacturing facility management, and safeguards. The contents and results of DUPIC R and D up to now are as follow: - the basic foundation was established for the critically required pelletizing technology and powder treatment technology for DUPIC. - development of DUPIC process line and deployment of 20 each process equipment and examination instruments in DFDF. - powder and pellet characterization study was done at PIEF based on the simfuel study results, and 30 DUPIC pellets were successfully produced. - the manufactured pellets were used for sample fuel rods irradiated in July,2000 in HANARO research reactor in KAERI and has been under post irradiation examination. (Hong, J. S.)

  18. Importance of Advanced Planning of Manufacturing for Nuclear Industry

    Directory of Open Access Journals (Sweden)

    Shykinov Nick

    2016-06-01

    Full Text Available In the context of energy demands by growing economies, climate changes, fossil fuel pricing volatility, and improved safety and performance of nuclear power plants, many countries express interest in expanding or acquiring nuclear power capacity. In the light of the increased interest in expanding nuclear power the supply chain for nuclear power projects has received more attention in recent years. The importance of the advanced planning of procurement and manufacturing of components of nuclear facilities is critical for these projects. Many of these components are often referred to as long-lead items. They may be equipment, products and systems that are identified to have a delivery time long enough to affect directly the overall timing of a project. In order to avoid negatively affecting the project schedule, these items may need to be sourced out or manufactured years before the beginning of the project. For nuclear facilities, long-lead items include physical components such as large pressure vessels, instrumentation and controls. They may also mean programs and management systems important to the safety of the facility. Authorized nuclear operator training, site evaluation programs, and procurement are some of the examples. The nuclear power industry must often meet very demanding construction and commissioning timelines, and proper advanced planning of the long-lead items helps manage risks to project completion time. For nuclear components there are regulatory and licensing considerations that need to be considered. A national nuclear regulator must be involved early to ensure the components will meet the national legal regulatory requirements. This paper will discuss timing considerations to address the regulatory compliance of nuclear long-lead items.

  19. Classification of spent nuclear fuel (SNF)

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. This document discusses the classification of spent nuclear fuels

  20. Nuclear fuel fabrication - developing indigenous capability

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, U.C.; Jayaraj, R.N.; Meena, R.; Sastry, V.S.; Radhakrishna, C.; Rao, S.M.; Sinha, K.K. [Nuclear Fuel Complex, Dept. of Atomic Energy (India)

    1997-07-01

    Nuclear Fuel Complex (NFC), established in early 70's for production of fuel for PHWRs and BWRs in India, has made several improvements in different areas of fuel manufacturing. Starting with wire-wrap type of fuel bundles, NFC had switched over to split spacer type fuel bundle production in mid 80's. On the upstream side slurry extraction was introduced to prepare the pure uranyl nitrate solution directly from the MDU cake. Applying a thin layer of graphite to the inside of the tube was another modification. The Complex has developed cost effective and innovative techniques for these processes, especially for resistance welding of appendages on the fuel elements which has been a unique feature of the Indian PHWR fuel assemblies. Initially, the fuel fabrication plants were set-up with imported process equipment for most of the pelletisation and assembly operations. Gradually with design and development of indigenous equipment both for production and quality control, NFC has demonstrated total self reliance in fuel production by getting these special purpose machines manufactured indigenously. With the expertise gained in different areas of process development and equipment manufacturing, today NFC is in a position to offer know-how and process equipment at very attractive prices. The paper discusses some of the new processes that are developed/introduced in this field and describes different features of a few PLC based automatic equipment developed. Salient features of innovative techniques being adopted in the area Of UO{sub 2} powder production are also briefly indicated. (author)

  1. Advanced nuclear power plants in Korea

    International Nuclear Information System (INIS)

    Full text: Korea Hydro and Nuclear Power Co., Ltd (KHNP) is the largest power company among the six subsidiaries that separated from Korea Electric Power Corporation (KEPCO) in 2001, accounting for approximately 25% of electricity producing facilities, hydro and nuclear combined. KHNP operates 20 nuclear power plants in Kori, Yonggwang, Ulchin and Wolsong site and several hydroelectric power generation facilities, providing approximately 36% of the national power supply. As a major source of electricity generation in Korea, nuclear energy contributes greatly to the stability of national electricity supply and energy security. KHNP's commercial nuclear power plant operation, which started with Kori Unit 1 in 1978, has achieved an average capacity factor more than 90% since 2000 and a high record of 93.4% in 2008. Following the introduction of nuclear power plants in the 1970's, Korea accumulated its nuclear technology in the 1980's, developed OPR 1000(Optimized Power Reactor) and demonstrated advanced level of its nuclear technology capabilities in the 2000's by developing an advanced type reactor, APR 1400(Advanced Power Reactor) which is being constructed at Shin-Kori Unit 3 and 4 for the first time. By 2022, KHNP will construct additional 12 nuclear power plants in order to ensure a stable power supply according to the Government Plan of Long-Term Electricity supply and Demand. 4 units of OPR 1000 reactor model will be commissioned by 2013 and 8 units of APR 1400 are under construction and planned. At the end of 2022, the nuclear capacity will reach 33% share of total generation capacity in Korea and account for 48% of national power generation. (author)

  2. Nuclear fuel cycle requirements in WOCA

    International Nuclear Information System (INIS)

    OECD/NEA will publsih an updated version of its study 'Nuclear Fuel Cycle Requirements and Supply Considerations, Through the Long-Term.' The Nuclear Research Centre Karlsruhe (KfK) was involved in the work necessary to provide this book. Although KfK had only responsiblility for part of the required computations it performed all the calculations for its own documentation interests. This documentation was felt to be a helpful background material for the reader of the second 'Yellow Book'. In this sense the original strategy computer outprints are published now without any discussion of assumptions and results. (orig.)

  3. Seal for objects containing nuclear fuels

    International Nuclear Information System (INIS)

    In order to mark and check the identity of objects - in particular, nuclear fuel elements, these may be sealed. Sealing is required in the context of nuclear safeguards. In the seal proposed here, a multitude of randomly distributed particles with different electromagnetic properties is contained in a hollow space in a body, where they are held by a pin. When the seal is taken off, they enter another, larger hollow space, losing their given order. A seal of this type is easy to check in the undamaged state. (UWI)

  4. Cogema and the nuclear fuel. A clue role in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The present issue of 'Les Cahiers de COGEMAGAZINE' addresses the topics of nuclear fuel production especially for PWR and Breeder Reactors. The papers deal with: the sketchy history of French nuclear industry, the economy and fuel marketing, the situation of the PWR programme, the fuels for breeder and research reactors. In the end prospective and concluding considerations are given. The most significant lines of progress related to the new fuels are estimated to be: high burn-up (by increasing the resistance to fission gas pressure and irradiation), improvement of response to power excursions, fuel matrices of stronger retention, increase in the plutonium content of MOF, 100% MOF-fuelled reactors, optimizing the utilization of consumable poisons (for PWR) and very high burn-up and very long service lifetimes (for breeders)

  5. Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

    Science.gov (United States)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar

    2012-01-01

    A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

  6. Coordinating Space Nuclear Research Advancement and Education

    International Nuclear Information System (INIS)

    The advancement of space exploration using nuclear science and technology has been a goal sought by many individuals over the years. The quest to enable space nuclear applications has experienced many challenges such as funding restrictions; lack of political, corporate, or public support; and limitations in educational opportunities. The Center for Space Nuclear Research (CSNR) was established at the Idaho National Laboratory (INL) with the mission to address the numerous challenges and opportunities relevant to the promotion of space nuclear research and education.1 The CSNR is operated by the Universities Space Research Association and its activities are overseen by a Science Council comprised of various representatives from academic and professional entities with space nuclear experience. Program participants in the CSNR include academic researchers and students, government representatives, and representatives from industrial and corporate entities. Space nuclear educational opportunities have traditionally been limited to various sponsored research projects through government agencies or industrial partners, and dedicated research centers. Centralized research opportunities are vital to the growth and development of space nuclear advancement. Coordinated and focused research plays a key role in developing the future leaders in the space nuclear field. The CSNR strives to synchronize research efforts and provide means to train and educate students with skills to help them excel as leaders.

  7. Survey of nuclear fuel-cycle codes

    International Nuclear Information System (INIS)

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

  8. Survey of nuclear fuel-cycle codes

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-04-01

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

  9. Development of Experimental Facilities for Advanced Spent Fuel Management Technology

    International Nuclear Information System (INIS)

    The Advanced spent fuel Conditioning Process Facility(ACPF) and hotcell system technologies were developed in this program for demonstrating safely and effectively the Advanced spent fuel Conditioning Process(ACP) on a lab-scale. With the analysis of work flow and characteristics of the process, ACP was successively demonstrated on a lab-scale experiments and the performance of process was evaluated. The hotcell system was comprehensively evaluated with those results and the design data for the engineering-scale demonstration was derived to propose the direction for the future research and development. The main items performed in this project were as follows. - The reconstruction of ACPF hotcell and demonstration of the ACP - The design and operation technologies for α-γ type nuclear hot cell facility - The overall evaluation of the performance, safety and operation ability of the hotcell system - The acquisition of the government licences for construction and operation and the IAEA licence for using nuclear materials The results of safety analysis and environmental effects assessment and performance data for ACPF had been used for acquiring the government licence for facility operation. The valuable experiences on pyroprocess facility design and operation knowledges would be applied to new Mock-up Facility being scheduled to be a previous stage facility of Integrated Pyroprocess Facility

  10. Financing Strategies for Nuclear Fuel Cycle Facility

    International Nuclear Information System (INIS)

    To help meet our nation's energy needs, reprocessing of spent nuclear fuel is being considered more and more as a necessary step in a future nuclear fuel cycle, but incorporating this step into the fuel cycle will require considerable investment. This report presents an evaluation of financing scenarios for reprocessing facilities integrated into the nuclear fuel cycle. A range of options, from fully government owned to fully private owned, was evaluated using a DPL (Dynamic Programming Language) 6.0 model, which can systematically optimize outcomes based on user-defined criteria (e.g., lowest life-cycle cost, lowest unit cost). Though all business decisions follow similar logic with regard to financing, reprocessing facilities are an exception due to the range of financing options available. The evaluation concludes that lowest unit costs and lifetime costs follow a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. Other financing arrangements, however, including regulated utility ownership and a hybrid ownership scheme, led to acceptable costs, below the Nuclear Energy Agency published estimates. Overwhelmingly, uncertainty in annual capacity led to the greatest fluctuations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; the annual operating costs dominate the government case. It is concluded that to finance the construction and operation of such a facility without government ownership could be feasible with measures taken to mitigate risk, and that factors besides unit costs should be considered (e.g., legal issues, social effects, proliferation concerns) before making a decision on financing strategy

  11. Management and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    The programme consists of the long-term and short-term programme, the continued bedrock investigations, the underground research laboratory, the decision-making procedure in the site selection process and information questions during the site selection process. The National Board for Spent Nuclear Fuel hereby subunits both the SKB's R and D Programme 86 and the Board's statement concerning the programme. Decisions in the matter have been made by the Board's executive committee. (DG)

  12. Effective economics of nuclear fuel power complex

    International Nuclear Information System (INIS)

    Problems of the economic theory and practice of functioning the nuclear fuel power complex (NFPC) are considered. Using the principle of market equilibrium for optimization of the NFPC hierarchical system is analyzed. The main attention is paid to determining the prices of production and consumption of the NFPC enterprises. Economic approaches on the optimal calculations are described. The ecological safety of NPP and NFPC enterprises is analyzed. A conception of the market socialism is presented

  13. Holdup measurement for nuclear fuel manufacturing plants

    Energy Technology Data Exchange (ETDEWEB)

    Zucker, M.S.; Degen, M.; Cohen, I.; Gody, A.; Summers, R.; Bisset, P.; Shaub, E.; Holody, D.

    1981-07-13

    The assay of nuclear material holdup in fuel manufacturing plants is a laborious but often necessary part of completing the material balance. A range of instruments, standards, and a methodology for assaying holdup has been developed. The objectives of holdup measurement are ascertaining the amount, distribution, and how firmly fixed the SNM is. The purposes are reconciliation of material unbalance during or after a manufacturing campaign or plant decommissioning, to decide security requirements, or whether further recovery efforts are justified.

  14. Recycling in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The nuclear fuel cycle comprises the total scope from uranium mining to reprocessing and/or (direct) final disposal. In all stages there are waste arisings. Depending on the concentration of the activity, various degrees of shieldings are necessary. For many process wastes transport/storage casks are needed and repackaging for final disposal gives an unnecessary dose-rate. Thus it was almost natural to stretch the function of the packages also to final disposal. And since 1983 in Germany, most of the heavy casks are made from recycled scrap metal. For the spent fuel reprocessing gives a high percentage of recycling of energy-containing 'wastes'. However, this is combined with a complicated chemical process and the continuing trend towards higher burn-up is 'replacing' reprocessing and favouring final disposal. This is due to the deteriorating isotopic composition of uranium and plutonium in the spent fuel. (author) 2 figs., 5 refs

  15. Artificial vision in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    The development of artificial vision techniques opens a door to the optimization of industrial processes which the nuclear industry cannot miss out on. Backing these techniques represents a revolution in security and reliability in the manufacturing of a highly technological products as in nuclear fuel. Enusa Industrias Avanzadas S. A. has successfully developed and implemented the first automatic inspection equipment for pellets by artificial vision in the European nuclear industry which is nowadays qualified and is already developing the second generation of this machine. There are many possible applications for the techniques of artificial vision in the fuel manufacturing processes. Among the practices developed by Enusa Industrias Avanzadas are, besides the pellets inspection, the rod sealing drills detection and positioning in the BWR products and the sealing drills inspection in the PWR fuel. The use of artificial vision in the arduous and precise processes of full inspection will allow the absence of human error, the increase of control in the mentioned procedures, the reduction of doses received by the personnel, a higher reliability of the whole of the operations and an improvement in manufacturing costs. (Author)

  16. Report on interim storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-08-10

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

  18. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

  19. Advanced Coal-Fueled Gas Turbine Program

    Energy Technology Data Exchange (ETDEWEB)

    Horner, M.W.; Ekstedt, E.E.; Gal, E.; Jackson, M.R.; Kimura, S.G.; Lavigne, R.G.; Lucas, C.; Rairden, J.R.; Sabla, P.E.; Savelli, J.F.; Slaughter, D.M.; Spiro, C.L.; Staub, F.W.

    1989-02-01

    The objective of the original Request for Proposal was to establish the technological bases necessary for the subsequent commercial development and deployment of advanced coal-fueled gas turbine power systems by the private sector. The offeror was to identify the specific application or applications, toward which his development efforts would be directed; define and substantiate the technical, economic, and environmental criteria for the selected application; and conduct such component design, development, integration, and tests as deemed necessary to fulfill this objective. Specifically, the offeror was to choose a system through which ingenious methods of grouping subcomponents into integrated systems accomplishes the following: (1) Preserve the inherent power density and performance advantages of gas turbine systems. (2) System must be capable of meeting or exceeding existing and expected environmental regulations for the proposed application. (3) System must offer a considerable improvement over coal-fueled systems which are commercial, have been demonstrated, or are being demonstrated. (4) System proposed must be an integrated gas turbine concept, i.e., all fuel conditioning, all expansion gas conditioning, or post-expansion gas cleaning, must be integrated into the gas turbine system.

  20. Proceedings of the Third Scientific Presentation on Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    The proceeding contains papers presented in the Third Scientific Presentation on nuclear Fuel Element Cycle held on 4-5 Nov 1997 in Jakarta, Indonesia. These papers were divided by three groups that are technology of exploration, processing, purification and analysis of nuclear materials; technology of nuclear fuel elements and structures; and technology of waste management, safety and nuclear fuel cycle. There are 38 papers indexed individually. (ID)

  1. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

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

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  2. The use of thorium as an alternative nuclear fuel

    International Nuclear Information System (INIS)

    The use of thorium as an alternative or supplementary nuclear fuel is examined and compared with uranium. A description of various reactor types and their suitability to thorium fuel, and a description of various aspects of the fuel cycle from mining to waste disposal, are included. Comments are made on the safety and economics of each aspect of the fuel cycle and the extension of the lifetime of nuclear fuel

  3. Manufacture of nuclear fuel elements for commercial PWR in China

    International Nuclear Information System (INIS)

    Yibin Nuclear Fuel Element Plant (YFP) under the leadership of China National Nuclear Corporation is sole manufacturer in China to specialize in the production of fuel assemblies and associated core components for commercial PWR nuclear power plant. At the early of 1980's, it began to manufacture fuel assemblies and associated core components for the first core of QINSHAN 300 MW nuclear power plant designed and built by China itself. With the development of nuclear power industry in China and the demand for localization of nuclear fuel elements in the early 1990's, YFP cooperated with FRAMATOME France in technology transfer for design and manufacturing of AFA 2G fuel assembly and successfully supplied the qualified fuel assemblies for the reloads of two units of GUANGDONG Da Ya Bay 900 MW nuclear power plant (Da Ya Bay NPP), and has achieved the localization of fuel assemblies and nuclear power plants. Meanwhile, it supplied fuel assemblies and associated core components for the first core and further reloads of Pakistan CHASHMA 300 MW nuclear power plant which was designed and built by China, and now it is manufacturing AFA 2G fuel assemblies and associated core components for the first core of two units of NPQJVC 600 MW nuclear power plant. From 2001 on, YFP will be able to supply Da Ya Bay NPP with the third generation of fuel assembly-AFA 3G which is to realize a strategy to develop the fuel assembly being of long cycle reload and high burn-up

  4. World nuclear fuel cycle requirements 1985

    International Nuclear Information System (INIS)

    Projections of uranium requirements (both yellowcake and enrichment services) and spent fuel discharges are presented, corresponding to the nuclear power plant capacity projections presented in ''Commercial Nuclear Power 1984: Prospects for the United States and the World'' (DOE/EIA-0438(85)) and the ''Annual Energy Outlook 1984:'' (DOE/EIA-0383(84)). Domestic projections are provided through the year 2020, with foreign projections through 2000. The domestic projections through 1995 are consistent with the integrated energy forecasts in the ''Annual Energy Outlook 1984.'' Projections of capacity beyond 1995 are not part of an integrated energy foreccast; the methodology for their development is explained in ''Commercial Nuclear Power 1984.'' A range of estimates is provided in order to capture the uncertainty inherent in such forward projections. The methodology and assumptions are also stated. A glossary is provided. Two appendixes present additional material. This report is of particular interest to analysts involved in long-term planning for the disposition of radioactive waste generated from the nuclear fuel cycle. 14 figs., 18 tabs

  5. Ultrasonic spectral analysis for nuclear fuel characterization

    Energy Technology Data Exchange (ETDEWEB)

    Baroni, Douglas B.; Bittencourt, Marcelo S.Q.; Leal, Antonio M.M., E-mail: douglasbaroni@ien.gov.b, E-mail: bittenc@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    Ceramic materials have been widely used for various purposes in many different industries due to certain characteristics, such as high melting point and high resistance to corrosion. Concerning the areas of applications, automobile, aeronautics, naval and even nuclear, the characteristics of these materials should be strictly controlled. In the nuclear area, ceramics are of great importance once they are the nuclear fuel pellets and must have, among other features, a well controlled porosity due to mechanical strength and thermal conductivity required by the application. Generally, the techniques used to characterize nuclear fuel are destructive and require costly equipment and facilities. This paper aims to present a nondestructive technique for ceramic characterization using ultrasound. This technique differs from other ultrasonic techniques because it uses ultrasonic pulse in frequency domain instead of time domain, associating the characteristics of the analyzed material with its frequency spectrum. In the present work, 40 Alumina (Al{sub 2}O{sub 3}) ceramic pellets with porosities ranging from 5% to 37%, in absolute terms measured by Archimedes technique, were tested. It can be observed that the frequency spectrum of each pellet varies according to its respective porosity and microstructure, allowing a fast and non-destructive association of the same characteristics with the same spectra pellets. (author)

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

    Science.gov (United States)

    Palmer, David J.

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

  7. Advances in nuclear science and technology

    CERN Document Server

    Greebler, Paul

    1968-01-01

    Advances in Nuclear Science and Technology Volume 4 provides information pertinent to the fundamental aspects of advanced reactor concepts. This book discusses the advances in various areas of general applicability, including modern perturbation theory, optimal control theory, and industrial application of ionizing radiations.Organized into seven chapters, this volume begins with an overview of the technology of sodium-cooled fast breeder power reactors and gas-cooled power reactors. This text then examines the key role of reactor safety in the development of fast breeder reactors. Other chapt

  8. Thermal phenomenae in nuclear fuel rods

    International Nuclear Information System (INIS)

    Thermal phenomenae occurring in a nuclear fuel rod under irradiation are studied. The most important parameters of either steady or transient thermal states are determined. The validity of applying the Fourier's approximation equations to these problems is also studied. A computer program TRANS is developed in order to study the transient cases. This program solves a system of coupled, non-linear partial differential equations, of parabolic type, in cylindrical coordinates with various boundary conditions. The benchmarking of the TRANS program is done by comparing its predictions with the analytical solution of some simplified transient cases. Complex transient cases such as those corresponding to characteristic reactor accidents are studied, in particular for typical pressurized heavy water reactor (PHWR) fuel rods, such as those of Atucha I. The Stefan problem emerging in the case of melting of the fuel element is solved. Qualitative differences between the classical Stefan problem, without inner sources, and that one, which includes sources are discussed. The MSA program, for solving the Stefan problem with inner sources is presented; and furthermore, it serves to predict thermal evolution, when the fuel element melts. Finally a model for fuel phase change under irradiation is developed. The model is based on the dimensional invariants of the percolation theory when applied to the connectivity of liquid spires nucleated around each fission fragment track. Suggestions for future research into the subject are also presented. (autor)

  9. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    International Nuclear Information System (INIS)

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  10. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean [Texas A & M Univ., College Station, TX (United States); Shao, Lin [Texas A & M Univ., College Station, TX (United States); Tsvetkov, Pavel [Texas A & M Univ., College Station, TX (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Kennedy, Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  11. Tensile Hoop Behavior of Irradiated Zircaloy-4 Nuclear Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jaramillo, Roger A [ORNL; Hendrich, WILLIAM R [ORNL; Packan, Nicolas H [ORNL

    2007-03-01

    A method for evaluating the room temperature ductility behavior of irradiated Zircaloy-4 nuclear fuel cladding has been developed and applied to evaluate tensile hoop strength of material irradiated to different levels. The test utilizes a polyurethane plug fitted within a tubular cladding specimen. A cylindrical punch is used to compress the plug axially, which generates a radial displacement that acts upon the inner diameter of the specimen. Position sensors track the radial displacement of the specimen outer diameter as the compression proceeds. These measurements coupled with ram force data provide a load-displacement characterization of the cladding response to internal pressurization. The development of this simple, cost-effective, highly reproducible test for evaluating tensile hoop strain as a function of internal pressure for irradiated specimens represents a significant advance in the mechanical characterization of irradiated cladding. In this project, nuclear fuel rod assemblies using Zircaloy-4 cladding and two types of mixed uranium-plutonium oxide (MOX) fuel pellets were irradiated to varying levels of burnup. Fuel pellets were manufactured with and without thermally induced gallium removal (TIGR) processing. Fuel pellets manufactured by both methods were contained in fuel rod assemblies and irradiated to burnup levels of 9, 21, 30, 40, and 50 GWd/MT. These levels of fuel burnup correspond to fast (E > 1 MeV) fluences of 0.27, 0.68, 0.98, 1.4 and 1.7 1021 neutrons/cm2, respectively. Following irradiation, fuel rod assemblies were disassembled; fuel pellets were removed from the cladding; and the inner diameter of cladding was cleaned to remove residue materials. Tensile hoop strength of this cladding material was tested using the newly developed method. Unirradiated Zircaloy-4 cladding was also tested. With the goal of determining the effect of the two fuel types and different neutron fluences on clad ductility, tensile hoop strength tests were

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

    Energy Technology Data Exchange (ETDEWEB)

    David E. Shropshire

    2009-05-01

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

  13. In-Pile Thermal Conductivity Measurement Method for Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Joy L. Rempe; Brandon Fox; Heng Ban; Joshua E. Daw; Darrell L. Knudson; Keith G. Condie

    2009-08-01

    Thermophysical properties of advanced nuclear fuels and materials during irradiation must be known prior to their use in existing, advanced, or next generation reactors. Thermal conductivity is one of the most important properties for predicting fuel and material performance. A joint Utah State University (USU) / Idaho National Laboratory (INL) project, which is being conducted with assistance from the Institute for Energy Technology at the Norway Halden Reactor Project, is investigating in-pile fuel thermal conductivity measurement methods. This paper focuses on one of these methods – a multiple thermocouple method. This two-thermocouple method uses a surrogate fuel rod with Joule heating to simulate volumetric heat generation to gain insights about in-pile detection of thermal conductivity. Preliminary results indicated that this method can measure thermal conductivity over a specific temperature range. This paper reports the thermal conductivity values obtained by this technique and compares these values with thermal property data obtained from standard thermal property measurement techniques available at INL’s High Test Temperature Laboratory. Experimental results and material properties data are also compared to finite element analysis results.

  14. TVO's new encapsulation method for spent nuclear fuel

    International Nuclear Information System (INIS)

    Teollisuuden Voima Oy has developed a new encapsulation method for spent nuclear fuel s.c. cold process. Instead of casting (400 deg C) molten lead the new canister is filled with cold granulated material like quartz sand, lead shots or glassa beads. The new canister concept ACPC (Advanced Cold Process Canister) consists of the outer oxygen free copper canister and of the inner steel canister. The function of the steel canister is merely to give mechanical strength. The copper canister acts as corrosion protection guaranteeing practically lifetime of millions of years for the ACPC concept

  15. 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.

  16. The Advanced BWR Nuclear Plant: Safe, economic nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Redding, J.R. [GE Nuclear Energy, San Jose, CA (United States)

    1994-12-31

    The safety and economics of Advanced BWR Nuclear Power Plants are outlined. The topics discussed include: ABWR Programs: status in US and Japan; ABWR competitiveness: safety and economics; SBWR status; combining ABWR and SBWR: the passive ABWR; and Korean/GE partnership.

  17. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. 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. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  18. Transportation capabilities study of DOE-owned spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1994-10-01

    This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

  19. Seismic design considerations for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    During the last few decades, there have been considerable advances in the field of a seismic design of nuclear structures and components housed inside a Nuclear power Plant (NPP). The seismic design and qualification of theses systems and components are carried out through the use of well proven and established theoretical as well as experimental means. Many of the related research works pertaining to these methods are available in the published literature, codes, guides etc. Contrary to this, there is very little information available with regards to the seismic design aspects of the nuclear fuel cycle facilities. This is probably on account of the little importance attached to these facilities from the point of view of seismic loading. In reality, some of these facilities handle a large inventory of radioactive materials and, therefore, these facilities must survive during a seismic event without giving rise to any sort of undue radiological risk to the plant personnel and the public at large. Presented herein in this paper are the seismic design considerations which are adopted for the design of nuclear fuel cycle facilities in India. (author)

  20. Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors. Report of the collaborative project COOL of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000, based on a resolution by the IAEA General Conference (GC(44)/RES/21). INPRO aims at helping to ensure that nuclear energy is available in the twenty-first century in a sustainable manner, and seeks to bring together all interested Member States, both technology holders and technology users, to jointly consider actions to achieve desired innovations. INPRO is taking care of the specific needs of developing countries. One of the aims of INPRO is to develop options for enhanced sustainability through promotion of technical and institutional innovations in nuclear energy technology through collaborative projects among IAEA Member States. Collaboration among INPRO members is fostered on selected innovative nuclear technologies to bridge technology gaps. Collaborative projects have been selected so that they complement other national and international R and D activities. The INPRO Collaborative Project COOL on Investigation of Technological Challenges Related to the Removal of Heat by Liquid Metal and Molten Salt Coolants from Reactor Cores Operating at High Temperatures investigated the technological challenges of cooling reactor cores that operate at high temperatures in advanced fast reactors, high temperature reactors and accelerator driven systems by using liquid metals and molten salts as coolants. The project was initiated in 2008 and was led by India; experts from Brazil, China, Germany, India, Italy and the Republic of Korea participated and provided chapters of this report. The INPRO Collaborative Project COOL addressed the following fields of research regarding liquid metal and molten salt coolants: (i) survey of thermophysical properties; (ii) experimental investigations and computational fluid dynamics studies on thermohydraulics, specifically pressure drop and heat transfer under different operating conditions; (iii) monitoring and control of coolant

  1. Nuclear fuel burn-up credit for criticality safety justification of spent nuclear fuel storage systems

    International Nuclear Information System (INIS)

    Burn-up credit analysis of RBMK-1000 an WWER-1000 spent nuclear fuel accounting only for actinides is carried out and a method is proposed for actinide burn-up credit. Two burn-up credit approaches are analyzed, which consider a system without and with the distribution of isotopes along the height of the fuel assembly. Calculations are performed using SCALE and MCNP computer codes

  2. Shielding considerations for advanced space nuclear reactor systems

    International Nuclear Information System (INIS)

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO2) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications

  3. Shielding considerations for advanced space nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Angelo, J.P. Jr.; Buden, D.

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

  4. Disposal of Canada's nuclear fuel waste

    International Nuclear Information System (INIS)

    In 1978, the governments of Canada and Ontario established the Nuclear Fuel Waste Management program. As of the time of the conference, the research performed by AECL was jointly funded by AECL and Ontario Hydro through the CANDU owners' group. Ontario Hydro have also done some of the research on disposal containers and vault seals. From 1978 to 1992, AECL's research and development on disposal cost about C$413 million, of which C$305 was from funds provided to AECL by the federal government, and C$77 million was from Ontario Hydro. The concept involves the construction of a waste vault 500 to 1000 metres deep in plutonic rock of the Canadian Precambrian Shield. Used fuel (or possibly solidified reprocessing waste) would be sealed into containers (of copper, titanium or special steel) and emplaced (probably in boreholes) in the vault floor, surrounded by sealing material (buffer). Disposal rooms might be excavated on more than one level. Eventually all excavated openings in the rock would be backfilled and sealed. Research is organized under the following headings: disposal container, waste form, vault seals, geosphere, surface environment, total system, assessment of environmental effects. A federal Environmental Assessment Panel is assessing the concept (holding public hearings for the purpose) and will eventually make recommendations to assist the governments of Canada and Ontario in deciding whether to accept the concept, and how to manage nuclear fuel waste. 16 refs., 1 tab., 3 figs

  5. Advanced fuel system technology for utilizing broadened property aircraft fuels

    Science.gov (United States)

    Reck, G. M.

    1980-01-01

    Possible changes in fuel properties are identified based on current trends and projections. The effect of those changes with respect to the aircraft fuel system are examined and some technological approaches to utilizing those fuels are described.

  6. Assessing the proliferation resistance of innovative nuclear fuel cycles

    International Nuclear Information System (INIS)

    on the system rather than on the characteristics of the proliferant. A top-level measure of proliferation resistance for a fuel cycle system can be developed from a hierarchy of metrics. At the lowest level, intrinsic and extrinsic barriers to proliferation can be defined. These barriers, which were documented in the report 'Technological Opportunities to Increase the Resistance of Global Civilian Nuclear Power Systems', are recommended as a means to characterize the proliferation characteristics of a fuel cycle. From these characteristics, intermediate measures can be evaluated of the resistance of barriers to the acquisition of material and of resistance of the barriers to successfully using the material to make a weapon. Because of the complexity of nonproliferation assessments, the problem must be decomposed into manageable elements. The proposed decomposition involves the definition of a finite set of threats, definition of barriers to proliferation, development of metrics, and segmentation of the system being evaluated. The spectrum of potential threats of nuclear proliferation is complex and ranges from small terrorist cells to industrialized countries with advanced nuclear fuel cycles. Adding to this complexity, the potential objectives of these threats are highly multidimensional. This paper identifies three categories of proliferant: subnational, non-industrialized state, and developed state. These three categories are further subdivided into a total of eight threats depending on the aspirations of the proliferant regarding number of weapons sought, weapon yield, weapon reliability, and delivery vehicle. Because the proliferation resistance of a fuel cycle is highly-dependent on the specific threat being evaluated, the final product of the assessment may be a vector of results with a calculated value of proliferation resistance for each threat, rather than an aggregation of results where the proliferation resistance is averaged over the different threats

  7. Nuclear energy center site survey: fuel cycle studies

    Energy Technology Data Exchange (ETDEWEB)

    Harty, H.

    1976-05-01

    Background information for the Nuclear Regulatory Commission Nuclear Energy Center Site Survey is presented in the following task areas: economics of integrated vs. dispersed nuclear fuel cycle facilities, plutonium fungibility, fuel cycle industry model, production controls and failure contingencies, environmental impact, waste management, emergency response capability, and feasibility evaluations. (DG)

  8. Strategies for a competitive nuclear fuel

    International Nuclear Information System (INIS)

    In the new framework of electricity generation, Nuclear Power Plants are operated by the electric utilities based on the competition required by an increasingly deregulated and liberalized market so that there is frequently a competition between the strategies of innovation and standardisation. On one had, innovation promotes the use of new technologies, products and/or processes locking for a reduction of costs based on the increase of the operating margins, while, on the other hand, standardisation promotes the use of well known and consolidated technologies, products and/or processes looking for getting the maximum benefit from the accumulated previous operating experience. In order to evaluate the standardisation versus the innovation an analysis of risks and opportunities of each of these strategies applied to the industry of the nuclear fuel has been suggested. As a results of it, a combined strategy innovation + standardisation based on the integration of both basic strategies in a complementary mode, has been suggested, the disadvantages of each of these strategies being compensated with the advantages of the other one. in this way, the total risk is minimized, the global opportunities are maximized and the main overall objective of getting the maximum benefit of the combination of both strategies looking for a competitive nuclear fuel is guaranteed. (Author)

  9. Advanced membrane electrode assemblies for fuel cells

    Science.gov (United States)

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  10. Spent Nuclear Fuel Vibration Integrity Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Jiang, Hao [ORNL; Yan, Yong [ORNL; Bevard, Bruce Balkcom [ORNL

    2016-01-01

    The objective of this research is to collect dynamic experimental data on spent nuclear fuel (SNF) under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT), the hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The collected CIRFT data will be utilized to support ongoing spent fuel modeling activities, and support SNF transportation related licensing issues. Recent testing to understand the effects of hydride reorientation on SNF vibration integrity is also being evaluated. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance, Fuel structure contributes to the SNF system stiffness, There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response, and that data will also be presented.

  11. Recycling as an option of used nuclear fuel management strategy

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, Tomaz, E-mail: tomaz.zagar@gen-energija.s [GEN energija, d.o.o., Cesta 4. julija 42, 8270 Krsko (Slovenia); Institute Jozef Stefan, Jamova 39, 1000 Ljubljana (Slovenia); Bursic, Ales; Spiler, Joze [GEN energija, d.o.o., Cesta 4. julija 42, 8270 Krsko (Slovenia); Kim, Dana; Chiguer, Mustapha; David, Gilles; Gillet, Philippe [AREVA, 33 rue La Fayette, 75009 Paris (France)

    2011-04-15

    The paper presents recycling as an option of used nuclear fuel management strategy with specific focus on the Slovenia. GEN energija is an independent supplier of integral and competitive electricity for Slovenia. In response to growing energy needs, GEN has conducted several feasibility and installation studies of a new nuclear power plant in Slovenia. With sustainable development, the environment, and public acceptance in mind, GEN conducted a study with AREVA concerning the options for the management of its' new plant's used nuclear fuel. After a brief reminder of global political and economic context, solutions for used nuclear fuel management using current technologies are presented in the study as well as an economic assessment of a closed nuclear fuel cycle. The paper evaluates and proposes practical solutions for mid-term issues on used nuclear fuel management strategies. Different scenarios for used nuclear fuel management are presented, where used nuclear fuel recycling (as MOX, for mixed oxide fuel, and ERU, for enriched reprocessed uranium) are considered. The study concludes that closing the nuclear fuel cycle will allow Slovenia to have a supplementary fuel supply for its new reactor via recycling, while reducing the radiotoxicity, thermal output, and volume of its wastes for final disposal, reducing uncertainties, gaining public acceptance, and allowing time for capitalization on investments for final disposal.

  12. Framatome advanced nuclear power-benefits for our clients from the new company

    International Nuclear Information System (INIS)

    Framatome ANP (Advanced Nuclear Power) merges the complementary strengths of two global nuclear industry leaders Framatome and Siemens - offering clients the best technological solutions for safe, reliable and economical plant performance. With a combined workforce of 13,300 skilled individuals, Framatome ANP is now the nuclear industry's leading supplier. Serving as Original Equipment Manufacturer (OEM) for more than 90 reactors that provide about 30% of the world's total installed nuclear power capacity, our experienced resources remain focused on the local needs of individual clients, wherever in the world they may be. The Company main business used to be turnkey construction of complete Nuclear Power plants, BWR and PWR capabilities, heavy equipment manufacturing, comprehensive I and C capabilities, and also expertise and knowledge of VVER. Framatome ANP will benefit in all of its fields of activity of the experience gained through Framatome and Siemens' collaboration on the next generation reactor, the EPR, as well as on steam generators replacements and or modernization of VVER. Framatome ANP nuclear fuel designs for both PWR and BWR plants provide innovative features and world-leading performance. Framatome ANP is organized according a matrix organization with: - 4 Business Groups (Project and Engineering, Service, Nuclear Fuel, Mechanical Equipment) - 3 Regional Divisions (Framatome Advanced Nuclear Power S.A.S., France; Framatome Advanced Nuclear Power GmbH, Germany; Framatome Advanced Nuclear Power Inc., USA). By 30th January 2001 Siemens Nuclear Power GmbH, founded in 2000 as successor of the Nuclear Division of Siemens Power Generation Group (KWU), was renamed to Framatome Advanced Nuclear Power GmbH forming the German part of the world wide acting company. Over the past 40 years 23 nuclear power plants all around the world - not only pressurized and boiling water reactors, but also two heavy-watermoderated reactors have been designed, constructed and

  13. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ilas, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Spencer, B. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14C, 3H, 129I, and 85Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is

  14. Passive neutron assay of irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Passive neutron assay of irradiated nuclear fuel has been investigated by calculations and experiments as a simple, complementary technique to the gamma assay. From the calculations it is found that the neutron emission arises mainly from the curium isotopes, the neutrons exhibit very good penetrability of the assemblies, and the neutron multiplication is not affected by the burnup. From the experiments on BWR and PWR assemblies, it is found that the neutron emission rate is proportional to burnup raised to 3.4 power. Recent investigations indicate that the passive neutron assay is a simple and useful technique to determine the consistency of burnups between assemblies. 10 refs

  15. Passive neutron assay of irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Passive neutron assay of irradiated nuclear fuel has been investigated by calculations and experiments as a simple, complementary technique to the gamma assay. From the calculations it was found that the neutron emission arises mainly from the curium isotopes, the neutrons exhibit very good penetrability of the assemblies, and the neutron multiplication is not affected by the burnup. From the experiments on BWR and PWR assemblies, the neutron emission rate is proportional to burnup raised to 3.4 power. The investigations indicate that the passive neutron assay is a simple and useful technique to determine the consistency of burnups between assemblies

  16. The management of nuclear fuel waste

    International Nuclear Information System (INIS)

    A Select Committee of the Legislature of Ontario was established to examine the affairs of Ontario Hydro, the provincial electrical utility. The Committee's terms of reference included examination of the waste management program being carried out jointly by the Ontario provincial government and the Canadian federal government. Public hearings were held which included private citizens as well as officials of organizations in the nuclear field and independent experts. Recommendations were made concerning the future direction of the Canadian fuel waste management program. (O.T.)

  17. Seal for an object containing nuclear fuel

    International Nuclear Information System (INIS)

    This seal which cannot be counterfeited, specially for sealing nuclear objects, e.g. fuel rods, not only makes any damage which has taken place obvious, but makes identification according to a key possible. For this purpose a minimum number of 'particles' or small bodies, which are identical but of different permeability, are fixed inside a short tube during 'loading' of the seal in a certain or an accidental sequence. The sequence of the spheres, which represents a key, can only be determined by special electromagnetic measuring equipment. On first opening the seal, this key sequence is irrevocably destroyed. (HP)

  18. Fluidization calculation on nuclear fuel kernel coating

    International Nuclear Information System (INIS)

    The fluidization of nuclear fuel kernel coating was calculated. The bottom of the reactor was in the from of cone on top of the cone there was a cylinder, the diameter of the cylinder for fluidization was 2 cm and at the upper part of the cylinder was 3 cm. Fluidization took place in the cone and the first cylinder. The maximum and the minimum velocity of the gas of varied kernel diameter, the porosity and bed height of varied stream gas velocity were calculated. The calculation was done by basic program

  19. Fuel element storage pond for nuclear installations

    International Nuclear Information System (INIS)

    In a fuel element storage pond for nuclear installations, with different water levels, radioactive particles are deposited at the points of contact of the water surface with the pond wall. So that this deposition will not occur, a metal apron is provided in the area of the points of contact of the water surface with the bond wall. The metal apron consists of individual sheets of metal which are suspended by claws in wall hooks. To clean the sheets, these are moved to a position below the water level. The sheets are suspended from the wall hooks during this process. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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