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Sample records for htr fuel cycle

  1. Actinide production in different HTR-fuel cycle concepts

    International Nuclear Information System (INIS)

    Filges, D.; Hecker, R.; Mirza, N.; Rueckert, M.

    1978-01-01

    At the 'Institut fuer Reaktorentwicklung der Kernforschungsanlage Juelich' the production of α-activities in the following HTR-OTTO cycle concepts were studied: 1. standard HTR cycle (U-Th); 2. low enriched HTR cycle (U-Pu); 3. near breeder HTR cycle (U-Th); 4. combined system (conventional and near breeder HTR). The production of α-activity in HTR Uranium-Thorium fuel cycles has been investigated and compared with the standard LWR cycles. The production of α-activity in HTR Uranium-Thorium fuel cycles has been investigated and compared with the standard LWR cycles. The calculations were performed by the short depletion code KASCO and the well-known ORIGEN program

  2. Fuel cycle studies for the Dragon HTR

    Energy Technology Data Exchange (ETDEWEB)

    Desoisa, J A; Nunn, R M; Twitchin, A E

    1971-02-15

    This note reports the progress made at B.N.L. in the study of the fuel cycle for the HTR design described by Daub (1970). The primary purpose of the study is to examine the special problems of the approach to equilibrium fuel cycle.

  3. Prospective studies of HTR fuel cycles involving plutonium

    International Nuclear Information System (INIS)

    Bonin, B.; Greneche, D.; Carre, F.; Damian, F.; Doriath, J.Y.

    2002-01-01

    High Temperature Gas Cooled reactors (HTRs) are able to accommodate a wide variety of mixtures of fissile and fertile materials without any significant modification of the core design. This flexibility is due to an uncoupling between the parameters of cooling geometry, and the parameters which characterize neutronic optimisation (moderation ratio or heavy nuclide concentration and distribution). Among other advantageous features, an HTR core has a better neutron economy than a LWR because there is much less parasitic capture in the moderator (capture cross section of graphite is 100 times less than the one of water) and in internal structures. Moreover, thanks to the high resistance of the coated particles, HTR fuels are able to reach very high burn-ups, far beyond the possibilities offered by other fuels (except the special case of molten salt reactors). These features make HTRs especially interesting for closing the nuclear fuel cycle and stabilizing the plutonium inventory. A large number of fuel cycle studies are already available today, on 3 main categories of fuel cycles involving HTRs : i) High enriched uranium cycle, based on thorium utilization as a fertile material and HEU as a fissile material; ii) Low enriched uranium cycle, where only LEU is used (from 5% to 12%); iii) Plutonium cycle based on the utilization of plutonium only as a fissile material, with (or without) fertile materials. Plutonium consumption at high burnups in HTRs has already been tested with encouraging results under the DRAGON project and at Peach Bottom. To maximize plutonium consumption, recent core studies have also been performed on plutonium HTR cores, with special emphasis on weapon-grade plutonium consumption. In the following, we complete the picture by a core study for a HTR burning reactor-grade plutonium. Limits in burnup due to core neutronics are investigated for this type of fuel. With these limits in mind, we study in some detail the Pu cycle in the special case of a

  4. HTR fuel research in the HTR-TN network on the high flux reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; Conrad, R.; Sevini, P.; Burghartz, M. [HFR Unit, Institute for Advanced Materials, European Commission, Joint Research Centre, Petten (Netherlands); Languille, A. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Guillermier, P. [FRAMATOME ANP, 69 - Lyon (France); Bakker, K. [Nuclear Research and Consultancy Group, Petten (Netherlands); Nabielek, H. [Forschungszentrum Juelich (Germany)

    2001-07-01

    Foremost, this paper explains the economic and strategic reasons for the comeback of the HTR reactor as one of the most promising reactors in the future. To study all the points related to HTR technology, a European network called HTR-TN was created in April 2000, with actually twenty European companies involved. This paper explains the organisation of the network and the related task-groups. In the field of fuel, one of these task-groups works on the fuel cycle and another works on the fuel itself in order to validate by testing HTR fuel possibilities. To this aim, an experimental loop is under construction in the HFR reactor to test full-size pebble type fuel elements and another under study to test compact fuel possibilities. These loops are based on all the experience accumulated by the High Flux Reactor in the years 70-90, when a lot of test were performed for fuel and material for the HTR technology and the facility design uses all the existing HFR knowledge. In conclusion, a host of research work, co-ordinated in the frame of a European network HTR-TN has begun. and should allow in the near future a substantial progress in the knowledge of this very promising fuel. (author)

  5. HTR fuel research in the HTR-TN network on the high flux reactor

    International Nuclear Information System (INIS)

    Guidez, J.; Conrad, R.; Sevini, P.; Burghartz, M.; Languille, A.; Guillermier, P.; Bakker, K.; Nabielek, H.

    2001-01-01

    Foremost, this paper explains the economic and strategic reasons for the comeback of the HTR reactor as one of the most promising reactors in the future. To study all the points related to HTR technology, a European network called HTR-TN was created in April 2000, with actually twenty European companies involved. This paper explains the organisation of the network and the related task-groups. In the field of fuel, one of these task-groups works on the fuel cycle and another works on the fuel itself in order to validate by testing HTR fuel possibilities. To this aim, an experimental loop is under construction in the HFR reactor to test full-size pebble type fuel elements and another under study to test compact fuel possibilities. These loops are based on all the experience accumulated by the High Flux Reactor in the years 70-90, when a lot of test were performed for fuel and material for the HTR technology and the facility design uses all the existing HFR knowledge. In conclusion, a host of research work, co-ordinated in the frame of a European network HTR-TN has begun. and should allow in the near future a substantial progress in the knowledge of this very promising fuel. (author)

  6. New Developments in Actinides Burning with Symbiotic LWR-HTR-GCFR Fuel Cycles

    International Nuclear Information System (INIS)

    Bomboni, Eleonora

    2008-01-01

    The long-term radiotoxicity of the final waste is currently the main drawback of nuclear power production. Particularly, isotopes of Neptunium and Plutonium along with some long-lived fission products are dangerous for more than 100000 years. 96% of spent Light Water Reactor (LWR) fuel consists of actinides, hence it is able to produce a lot of energy by fission if recycled. Goals of Generation IV Initiative are reduction of long-term radiotoxicity of waste to be stored in geological repositories, a better exploitation of nuclear fuel resources and proliferation resistance. Actually, all these issues are intrinsically connected with each other. It is quite clear that these goals can be achieved only by combining different concepts of Gen. IV nuclear cores in a 'symbiotic' way. Light-Water Reactor - (Very) High Temperature Reactor ((V)HTR) - Fast Reactor (FR) symbiotic cycles have good capabilities from the viewpoints mentioned above. Particularly, HTR fuelled by Plutonium oxide is able to reach an ultra-high burn-up and to burn Neptunium and Plutonium effectively. In contrast, not negligible amounts of Americium and Curium build up in this core, although the total mass of Heavy Metals (HM) is reduced. Americium and Curium are characterised by an high radiological hazard as well. Nevertheless, at least Plutonium from HTR (rich in non-fissile nuclides) and, if appropriate, Americium can be used as fuel for Fast Reactors. If necessary, dedicated assemblies for Minor Actinides (MA) burning can be inserted in Fast Reactors cores. This presentation focuses on combining HTR and Gas Cooled Fast Reactor (GCFR) concepts, fuelled by spent LWR fuel and depleted uranium if need be, to obtain a net reduction of total mass and radiotoxicity of final waste. The intrinsic proliferation resistance of this cycle is highlighted as well. Additionally, some hints about possible Curium management strategies are supplied. Besides, a preliminary assessment of different chemical forms of

  7. French programme for HTR fuel

    International Nuclear Information System (INIS)

    Gillet, R.M.

    1991-01-01

    It is reported that in the frameworks of the French HTR research program, stopped in 1979 the HTR coated particle fuel, fuel rod and prismatic fuel element design have been successfully developed and irradiation tested in France and specific examination methods for irradiated fuel particles, rods and graphite blocks have been developed. Currently CEA is involved in fission product transport experiments sponsored by the US Department of Energy and performed in the COMEDIE loop. Finally the CEA follows progress and developments in HTR fuel research and development throughout the world. 1 tab

  8. Plutonium re-cycle in HTR

    Energy Technology Data Exchange (ETDEWEB)

    Desoisa, J. A.

    1974-03-15

    The study of plutonium cycles in HTRs using reprocessed plutonium from Magnox and AGR fuel cycles has shown that full core plutonium/uranium loadings are in general not feasible, burn-up is limited due the need for lower loadings of plutonium to meet reload core reactivity limits, on-line refueling is not practicable due to the need for higher burnable poison loadings, and low conversion rates in the plutonium-uranium cycles cannot be mitigated by axial loading schemes so that fissile make-up is needed if HTR plutonium recycle is desired.

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

    International Nuclear Information System (INIS)

    Gray S. Chang

    2005-01-01

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

  10. The properties of spherical fuel elements and its behavior in the modular HTR

    International Nuclear Information System (INIS)

    Lohnert, G.H.; Ragoss, H.

    1985-01-01

    The reference fuel element for all future HTR applications in the Federal Republic of Germany as developed by NUKEM/HOBEG in the framework of the 'High temperature Fuel-Cycle Project' had to be scrutinised for its compatibility with all the other design principles of the modular HTR, or possibly for restrictions forced upon reactor layout. This reference fuel element can be characterized by the following features: moulded spherical fuel element of 60 mm in diameter with fuel free shell of 5 mm thickness, based on carbon matrix; low enriched uranium (U/Pu fuel cycle); UO 2 fuel kernels; TRISO coating (pyrocarbon and additional SiC layers)

  11. Operational requirements of spherical HTR fuel elements and their performance

    International Nuclear Information System (INIS)

    Roellig, K.; Theymann, W.

    1985-01-01

    The German development of spherical fuel elements with coated fuel particles led to a product design which fulfils the operational requirements for all HTR applications with mean gas exit temperatures from 700 deg C (electricity and steam generation) up to 950 deg C (supply of nuclear process heat). In spite of this relatively wide span for a parameter with strong impact on fuel element behaviour, almost identical fuel specifications can be used for the different reactor purposes. For pebble bed reactors with relatively low gas exit temperatures of 700 deg C, the ample design margins of the fuel elements offer the possibility to enlarge the scope of their in-service duties and, simultaneously, to improve fuel cycle economics. This is demonstrated for the HTR-500, an electricity and steam generating 500 MWel eq plant presently proposed as follow-up project to the THTR-300. Due to the low operating temperatures of the HTR-500 core, the fuel can be concentrated in about 70% of the pebbles of the core thus saving fuel cycle costs. Under all design accident conditions fuel temperatures are maintained below 1250 deg C. This allows a significant reduction in the engineered activity barriers outside the primary circuit, in particular for the loss of coolant accident. Furthermore, access to major primary circuit components and the reuse of the fuel elements after any design accident are possible. (author)

  12. A synthesis on the HTR scenario studies at CEA - HTR2008-58059

    International Nuclear Information System (INIS)

    Boucher, L.; Greneche, D.

    2008-01-01

    The aim of the studies is to assess the impact of the deployment of an HTR park replacing one part of the current PWR reactors. The other part of the current park is replaced by EPRs. In these scenarios, the annual electricity production is constant at 400 TWhe. This value corresponds roughly to the present nuclear electricity production in France. From 2002 to 2007, an important program study on HTR has been carried out by CEA and AREVA NC under the joint CEA - AREVA NC project 'prospective studies on the management of Plutonium and the back end of the cycle'. This program addresses core physic and scenario studies, and also the back end of the fuel cycle : reprocessing of spent fuel and HTR waste management. Some core physic studies have already been presented in the reference [1]. This paper presents the results of the scenario studies using two concepts: either the standard core of the Gas Turbine Modular Helium Reactor concept (GTMHR) with Uranium or Plutonium fuel, or the Multiple Fuel Rows Core (MFRC) dedicated to the actinide burning. The insertion of a new concept (fuel, reactor, process) must be evaluated in the global electronuclear system with an analysis of the impact on the fuel cycle (Enrichment, Fuel Fabrication, Reactor, Processing, Interim Storage, Waste storage). The scenario studies are used to evaluate different solutions to manage nuclear materials (uranium, plutonium) and wastes (minor actinides and fission products), from the present situation in France (closed cycle with storage of used MOX fuels) until the final equilibrium: mixed nuclear park with EPR and HTR. These studies allow to calculate material flows and inventories of these elements in each step of the fuel cycle. The simulation of transient scenarios from the present situation to the future situation is performed with the COSI code. HTR reactors feature a high flexibility with regard to fuel cycle options. Several versions of core have been investigated, with different type of

  13. 'Once through' cycles in the pebble bed HTR

    International Nuclear Information System (INIS)

    Teuchert, E.

    1977-12-01

    In the pebble bed HTR the 'Once Through' cycles achieve a favorable conservation of uranium resources due to their high burnup and due to the relatively low fissile inventory. A detailed study is given for cycles with highly enriched uranium and thorium, 20% enriched uranium and thorium, and for the low (approximately 8%) enriched cycle. The recommended cycle is based on the known THTR fuel element in the Th/U (93%) cycle. The variant with separate Seed elements and Breed elements presents the best pioneer in view of later recycling and thermal breeding. The minimum proliferation risk is achieved in the Th/U (20%) cycle basing on the fuel element type of the AVR, due to the low amount and high denaturization of the disloaded plutonium. (orig.) [de

  14. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  15. Fuel management of HTR-10

    International Nuclear Information System (INIS)

    Wu Zongxin; Jing Xingqing

    2001-01-01

    The 10 MW high temperature cooled reactor (HTR-10) built in Tsinghua University is a pebble bed type of HTGR. The continuous recharge and multiple-pass of spherical fuel elements are used for fuel management. The initiative stage of core is composed of the mix of spherical fuel elements and graphite elements. The equilibrium stage of core is composed of identical spherical fuel elements. The fuel management during the transition from the initiative stage to the equilibrium stage is a key issue for HTR-10 physical design. A fuel management strategy is proposed based on self-adjustment of core reactivity. The neutron physical code is used to simulate the process of fuel management. The results show that the graphite elements, the recharging fuel elements below the burn-up allowance, and the discharging fuel elements over the burn-up allowance could be identified by burn-up measurement. The maximum of burn-up fuel elements could be controlled below the burn-up limit

  16. Introduction of HTR-PM Operation and Fuel Management System

    International Nuclear Information System (INIS)

    Liu Fucheng; Luo Yong; Gao Qiang

    2014-01-01

    There is a big difference between High Temperature Gas-cooled Reactor Pebble-modules Demonstration Project(HTR-PM) and PWR in operation mode. HTR-PM is a continually refuelled reactor, and the operation and fuel management of it, which affect each other, are inseparable. Therefore, the analysis of HTR-PM fuel management needs to be carried out “in real time”. HTR-PM operation and fuel management system is developed for on-power refuelling mode of HTR-PM. The system, which calculates the core neutron flux and power distribution, taking high-temperature reactor physics analysis software-VSOP as a basic tool, can track and predict the core state online, and it has the ability to restructure core power distribution online, making use of ex-core detectors to correct and check tracking calculation. Based on the ability to track and predict, it can compute the core parameters to provide support for the operation of the reactor. It can also predict the operation parameters of the reactor to provide reference information for the fuel management.The contents of this paper include the development purposes, architecture, the main function modules, running process, and the idea of how to use the system to carry out HTR-PM fuel management. (author)

  17. Preparation of spherical fuel elements for HTR-PM in INET

    International Nuclear Information System (INIS)

    Xiangwen, Zhou; Zhenming, Lu; Jie, Zhang; Bing, Liu; Yanwen, Zou; Chunhe, Tang; Yaping, Tang

    2013-01-01

    Highlights: • Modifications and optimizations in the manufacture of spherical fuel elements (SFE) for HTR-PM are presented. • A newly developed overcoater exhibits good stability and high efficiency in the preparation of overcoated particles. • The optimized carbonization process reduces the process time from 70 h in the period of HTR-10 to 20 h. • Properties of the prepared SFE and matrix graphite balls meet the design specifications for HTR-PM. • In particular the mean free uranium fraction of 5 consecutive batches is only 8.7 × 10 −6 . -- Abstract: The spherical fuel elements were successfully manufactured in the period of HTR-10. In order to satisfy the mass production of fuel elements for HTR-PM, several measures have been taken in modifying and optimizing the manufacture process of fuel elements. The newly developed overcoater system and its corresponding parameters exhibited good stability and high efficiency in the preparation of overcoated particles. The optimized carbonization process could reduce the carbonization time from more than 70 h to 20 h and improve the manufacturing efficiency. Properties of the manufactured spherical fuel elements and matrix graphite balls met the design specifications for HTR-PM. The mean free uranium fraction of 5 consecutive batches was 8.7 × 10 −6 . The optimized fuel elements manufacturing process could meet the requirements of design specifications of spherical fuel elements for HTR-PM

  18. HTR fuel development for advanced application

    International Nuclear Information System (INIS)

    Nickel, H.; Balthesen, E.; Graham, L.W.; Hick, H.

    1975-01-01

    The advantages of the HTR for nuclear steam supply systems are briefly outlined. Due to its great design flexibility a number of different designs have evolved and the main characteristics of existing experimental prototype and power reactor HTR designs are summarized. The present state of coated particle fuel, particularly with regard to performance, is considered. Some implications of producing higher temperatures are discussed. Finally some of the developments in progress such as minimising the temperature drop between fuel and coolant, and of improving fuel performance by better fission product retention, better chemical stability, and the use of alternative coated materials, are discussed. (U.K.)

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

    Institute of Scientific and Technical Information of China (English)

    CHANG Hong; YANG Yongwei; JING Xingqing; XU Yunlin

    2006-01-01

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

  20. The high temperature reactor and its fuel cycle options

    International Nuclear Information System (INIS)

    1979-07-01

    The status of the HTR system in the Federal Republic of Germany as well as the consecutive steps and the probable cost of further development are presented. The considerations are based on a recycling Th/highly enriched uranium (HEU) fuel cycle which has been chosen as the main line of the German HTR R and D efforts. Alternative fuel cycles such as medium-enriched uranium (MEU) and low-enriched uranium (LEU) are discussed as well

  1. Waste arisings from a high-temperature reactor with a uranium-thorium fuel cycle

    International Nuclear Information System (INIS)

    1979-09-01

    This paper presents an equilibrium-recycle condition flow sheet for a high-temperature gas-cooled reactor (HTR) fuel cycle which uses thorium and high-enriched uranium (93% U-235) as makeup fuel. INFCE Working Group 7 defined percentage losses to various waste streams are used to adjust the heavy-element mass flows per gigawatt-year of electricity generated. Thorium and bred U-233 are recycled following Thorex reprocessing. Fissile U-235 is recycled one time following Purex reprocessing and then is discarded to waste. Plutonium and other transuranics are discarded to waste. Included are estimates of volume, radioactivity, and heavy-element content of wastes arising from HTR fuel element fabrication; HTR operation, maintenance, and decommissioning; and reprocessing spent fuel where the waste is unique to the HTR fuel cycle

  2. AREVA HTR concept for near-term deployment

    Energy Technology Data Exchange (ETDEWEB)

    Lommers, L.J., E-mail: lewis.lommers@areva.com [AREVA Inc., 2101 Horn Rapids Road, Richland, WA 99354 (United States); Shahrokhi, F. [AREVA Inc., Lynchburg, VA (United States); Mayer, J.A. [AREVA Inc., Marlborough, MA (United States); Southworth, F.H. [AREVA Inc., Lynchburg, VA (United States)

    2012-10-15

    This paper introduces AREVA's High Temperature Reactor (HTR) steam cycle concept for near-term industrial deployment. Today, nuclear power primarily impacts only electricity generation. The process heat and transportation fuel sectors are completely dependent on fossil fuels. In order to impact this energy sector as rapidly as possible, AREVA has focused its HTR development effort on the steam cycle HTR concept. This reduces near-term development risk and minimizes the delay before a useful contribution to this sector of the energy economy can be realized. It also provides a stepping stone to longer term very high temperature concepts which might serve additional markets. A general description of the current AREVA steam cycle HTR concept is provided. This concept provides a flexible system capable of serving a variety of process heat and cogeneration markets in the near-term.

  3. AREVA HTR concept for near-term deployment

    International Nuclear Information System (INIS)

    Lommers, L.J.; Shahrokhi, F.; Mayer, J.A.; Southworth, F.H.

    2012-01-01

    This paper introduces AREVA's High Temperature Reactor (HTR) steam cycle concept for near-term industrial deployment. Today, nuclear power primarily impacts only electricity generation. The process heat and transportation fuel sectors are completely dependent on fossil fuels. In order to impact this energy sector as rapidly as possible, AREVA has focused its HTR development effort on the steam cycle HTR concept. This reduces near-term development risk and minimizes the delay before a useful contribution to this sector of the energy economy can be realized. It also provides a stepping stone to longer term very high temperature concepts which might serve additional markets. A general description of the current AREVA steam cycle HTR concept is provided. This concept provides a flexible system capable of serving a variety of process heat and cogeneration markets in the near-term.

  4. HTR-TN a European network for the development of HTR technology

    International Nuclear Information System (INIS)

    Von Lensa, W.

    2001-01-01

    A network called High-temperature reactor technology network (HTR-TN) has been created at a European level to coordinate works and knowledge on the subject with a long-term perspective and to serve as a channel for international collaboration. An analysis confirmed that the obvious economic penalty of HTR due to its low density power could be compensated by the combination of recent advances that may completely change the positioning of HTR on the energy market: -) the modular concept allowed to get a reactor free from core melt risk without intervention of any active safety system, implying a drastic simplification of the design of the reactor and the safety systems as well as a standardisation and potential for shop fabrication in series; -) the development of gas turbines, the efficiency of which increased, in 10 years, from 35% till 50% and more, enabling to consider suppression of the secondary system; -) the ultra high burn-up potential of HTR fuel and the possibility for direct disposal of spent HTR fuel elements that may reduce cost of the fuel cycle and contribute to the reduction of civil and military plutonium stockpiles. (A.C.)

  5. HTR characteristics affecting reactor physics

    International Nuclear Information System (INIS)

    Ehlers, K.

    1980-01-01

    A physical description of high-temperature has-cooled reactors is given, followed by an overview of HTR characteristics. The emphasis is placed on the HTR fuel cycle alternatives and thermohydraulics of pebble bed core. Some prospects of HTRs in the Federal Republic of Germany are also presented

  6. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope

    2011-10-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  7. Fabrication technology of spherical fuel element for HTR-10

    International Nuclear Information System (INIS)

    He Jun; Zou Yanwen; Liang Tongxiang; Qiu Xueliang

    2002-01-01

    R and D on the fabrication technology of the spherical fuel elements for the 10 MW HTR Test Module (HTR-10) began from 1986. Cold quasi-isostatic molding with a silicon rubber die is used for manufacturing the spherical fuel elements.The fabrication technology and the graphite matrix materials were investigated and optimized. Twenty five batches of fuel elements, about 11000 of the fuel elements, have been produced. The cold properties of the graphite matrix materials satisfied the design specifications. The mean free uranium fraction of 25 batches was 5 x 10 -5

  8. The Renewal of HTR Development in Europe

    International Nuclear Information System (INIS)

    Hittner, Dominique

    2002-01-01

    The European HTR-Technology Network (HTR-TN), created in 2000, presently groups 20 organisations from European nuclear research and industry for developing the technologies of direct-cycle modular HTRs, which presently raise a large world-wide interest, because of their high potential for economic competitiveness, natural resource sparing, safety and minimisation of the waste impacts, in line with the goals of sustainable development of Generation IV. All aspects of HTR technologies are addressed by HTR-TN, from the reactor physics to the development of materials, fuel and components. Most of this activity is supported by the European Commission in the frame of its 5. EURATOM Framework Programme. The first results of HTR-TN programme are given: the analysis of the reactor physics international benchmark on the commissioning tests of HTTR (Japan), the long term behaviour of spent HTR fuel in geologic disposal conditions, the preparation of a very high burnup fuel irradiation and the development of fabrication processes for producing high performance coated particles, etc. (authors)

  9. Review on characterization methods applied to HTR-fuel element components

    International Nuclear Information System (INIS)

    Koizlik, K.

    1976-02-01

    One of the difficulties which on the whole are of no special scientific interest, but which bear a lot of technical problems for the development and production of HTR fuel elements is the proper characterization of the element and its components. Consequently a lot of work has been done during the past years to develop characterization procedures for the fuel, the fuel kernel, the pyrocarbon for the coatings, the matrix and graphite and their components binder and filler. This paper tries to give a status report on characterization procedures which are applied to HTR fuel in KFA and cooperating institutions. (orig.) [de

  10. Instrumentation of steam cycle HTR's up to 900 MWe

    International Nuclear Information System (INIS)

    Leithner, D.E.; Winkenbach, B.

    1982-06-01

    Due to basic design features and inherent safety qualities in-core instrumentation is not needed in an HTR. Reactor safety requirements can be met by integral measurements. A modest spatial resolving power of the out-of-core instrumentation is sufficient for all operational purposes in small and medium sized steam cycle HTR's. Thus, the instrumentation concept of the THTR 300 MWe prototype reactor can be adopted without major changes for the HTR 450 MWe reactor project, as is demonstrated here for the neutron flux and temperature measurements. (author)

  11. Experimental Study of Fuel Element Motion in HTR-PM Conveying Pipelines

    International Nuclear Information System (INIS)

    Wang Xin; Zhang Haiquan; Nie Junfeng; Li Hongke; Liu Jiguo; He Ayada

    2014-01-01

    The motion action of sphere fuel element (FE) inside fuel pipelines in HTR-PM is indeterminate. Fuel motion is closely connected with the interaction of FE and inner surface of fuel conveying pipe. In this paper, motion method of fuel elements in its conveying pipe is Experimental studied. Combined with the measurement of the fuel passing speed in stainless steel pipe and the track left by sphere ball for experiment, interaction modes of fuel and inner-surface of pipe, which is sliding friction, rolling friction and Collision, has been found. The modes of interaction can affect the speed of fuel conveying, amount of sphere waste and operation stability of fuel handling of high temperature reactor-pebble bed modules (HTR-PM). Furthermore, the motion process of fuel passing a big-elbow which is lying on the top of fuel pneumatic hoisting pipe were experimented. The result shows that the speed before and the speed after the elbow is positive correlation. But with the increase of speed before the elbow, the speed after the elbow increase less. Meanwhile the fuel conveying mode changes from friction to collision. And the conveying process is still steady. The effect can be used to controlling the speed of fuel conveying in fuel handling process of HTR-PM. (author)

  12. Development of a Reliable Fuel Depletion Methodology for the HTR-10 Spent Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kiwhan [Los Alamos National Laboratory; Beddingfield, David H. [Los Alamos National Laboratory; Geist, William H. [Los Alamos National Laboratory; Lee, Sang-Yoon [unaffiliated

    2012-07-03

    A technical working group formed in 2007 between NNSA and CAEA to develop a reliable fuel depletion method for HTR-10 based on MCNPX and to analyze the isotopic inventory and radiation source terms of the HTR-10 spent fuel. Conclusions of this presentation are: (1) Established a fuel depletion methodology and demonstrated its safeguards application; (2) Proliferation resistant at high discharge burnup ({approx}80 GWD/MtHM) - Unfavorable isotopics, high number of pebbles needed, harder to reprocess pebbles; (3) SF should remain under safeguards comparable to that of LWR; and (4) Diversion scenarios not considered, but can be performed.

  13. Actual characteristics study on HTR-10GT coupling with direct gas turbine cycle

    International Nuclear Information System (INIS)

    Peng Xuechuang; Zhu Shutang; Wang Jie

    2005-01-01

    HTR-10GT is a testing project coupling the reactor HTR-10 with direct gas turbine cycle. Its thermal cycle can be taken as a closed, recuperated and inter-cooled Brayton cycle. The present study is focused on the thermal cycle performance of HTR-10GT under practical conditions of leakage, pressure losses, etc.. Through thermodynamic analysis, the expression of cycle efficiency for actual thermal cycle is derived. By establishing a physical model with friction loss and leakage, a set of governing equation are constructed based on some reasonable assumptions. The results of actual cycle efficiency have been calculated for different leakage amount at different locations while the effects of leakage under different power level have also been calculated and analyzed. (authors)

  14. An evaluation of the results of the HTR fuel programme conducted in the Dragon reactor experiment

    International Nuclear Information System (INIS)

    Shepherd, L.R.

    1982-01-01

    The Dragon Reactor Experiment was used over a period of ten years to investigate the behaviour of HTR fuel elements under realistic service conditions. The purpose of the work was to develop fuel capable of meeting the requirements of commercial power reactors. The studies divided into areas concerned with the mechanical behaviour of the graphite core structure under fast neutron irradiation and the ability of the coated particle fuel to retain fissile products over commercially viable life-cycles. (author)

  15. The failure mechanisms of HTR coated particle fuel and computer code

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Shao Youlin; Liang Tongxiang; Tang Chunhe

    2010-01-01

    The basic constituent unit of fuel element in HTR is ceramic coated particle fuel. And the performance of coated particle fuel determines the safety of HTR. In addition to the traditional detection of radiation experiments, establishing computer code is of great significance to the research. This paper mainly introduces the structure and the failure mechanism of TRISO-coated particle fuel, as well as a few basic assumptions,principles and characteristics of some existed main overseas codes. Meanwhile, this paper has proposed direction of future research by comparing the advantages and disadvantages of several computer codes. (authors)

  16. Burner and dissolver off-gas treatment in HTR fuel reprocessing

    International Nuclear Information System (INIS)

    Barnert-Wiemer, H.; Heidendael, M.; Kirchner, H.; Merz, E.; Schroeder, G.; Vygen, H.

    1979-01-01

    In the reprocessing of HTR fuel, essentially all of the gaseous fission products are released during the heat-end tratment, which includes burning of the graphite matrix and dissolving of the heavy metallic residues in THOREX reagent. Three facilities for off-gas cleaning are described, the status of the facility development and test results are reported. Hot tests with a continuous dissolver for HTR-type fuel (throughput 2 kg HM/d) with a closed helium purge loop have been carried out. Preliminary results of these experiments are reported

  17. Solution of multiple circuits of steam cycle HTR system

    International Nuclear Information System (INIS)

    Li, Fu; Wang, Dengying; Hao, Chen; Zheng, Yanhua

    2014-01-01

    In order to analyze the dynamic operation performance and safety characteristics of the steam cycle high temperature gas cooled reactor (HTR) systems, it is necessary to find the solution of the whole HTR systems with all coupled circuits, including the primary circuit, the secondary circuit, and the residual heat removal system (RHRS). Considering that those circuits have their own individual fluidity and characteristics, some existing code packages for independent circuits themselves have been developed, for example THEMRIX and TINTE code for the primary circuit of the pebble bed reactor, BLAST for once through steam generator. To solve the coupled steam cycle HTR systems, a feasible way is to develop coupling method to integrate these independent code packages. This paper presents several coupling methods, e.g. the equivalent component method between the primary circuit and steam generator which reflect the close coupling relationship, the overlapping domain decomposition method between the primary circuit and the passive RHRS which reflects the loose coupling relationship. Through this way, the whole steam cycle HTR system with multiple circuits can be easily and efficiently solved by integration of several existing code packages. Based on this methodology, a code package TINTE–BLAST–RHRS was developed. Using this code package, some operation performance of HTR–PM was analyzed, such as the start-up process of the plant, and the depressurized loss of forced cooling accident when different number of residual heat removal trains is operated

  18. Nonproliferation and safeguard considerations: Pebble Bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, conpare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  19. Status of Research on Pebble Bed HTR Fuel Fabrication Technology in Indonesia

    International Nuclear Information System (INIS)

    Rachmawati, M.; Sarjono; Ridwan; Langenati, R.

    2014-01-01

    Research on pebble bed HTR fuel fabrication is conducted in Indonesia. One of the aims is to build a knowledge base on pebble bed HTR fuel element fabrication technology for fuel procurement. The steps of research strategies are firstly to understand the basic design research of TRISO fuel, properties, and requirements, and secondly to understand the TRISO fuel manufacturing technology, which comprises fabrication and quality control, including its facility. Both steps are adopted from research and experiences of the countries with HTR fuel element fabrication technology. From the knowledge gained in the research, an experimental design of the process and a set of prototype process equipment for fabrication are developed, namely kernels production using external gelation process, TRISO coating of the kernel, and pebble compacting. Experiments using the prototypes have been conducted. Characterization of the kernel product, i.e. diameter, sphericity, density and O/U ratio, shows that the kernel product is still not in compliance with the specification requirements. These are deemed to be caused mainly by the selected vibrating system and the viscosity adjustment. Another major cause is the selected NH3 and air feeding method for both NH3 and air layer in the preparation for spherical droplets of liquid. The FB-CVD TRISO coating of the kernel has been experimented but unsuccessful by using an FB-CVD once‐through continuous coating process. For the pebble compacting, the process is still in the early stage of setting-up compaction equipment. This paper summarizes the current status of research on HTR fuel fabrication technology in Indonesia, the proposed process and its equipment setting-up for improvement of the kernel production. The knowledge and lessons learned gained from the research is useful and can be an assistance in planning for fuel development laboratory facilities procurement, formulating User Requirement Document and Bid Invitation Specification for

  20. The HTR modular power reactor system. Qualification of fuel elements and materials

    International Nuclear Information System (INIS)

    Heidenreich, U.; Breitling, H.; Nieder, R.; Ohly, W.; Mittenkuehler, A.; Ragoss, H.; Seehafer, H.J.; Wirtz, K.; Serafin, N.

    1989-01-01

    For further development of the HTR modular power reactor system (HTR-M-KW), the project activities for 'Qualification of fuel elements and materials' reported here cover the work for specifying the qualifications to be met by metallic and ceramic materials, taking into account the design-based requirements and the engineered safety requirements. The fission product retention data determined for the HTR modular reactor fuel elements could be better confirmed by evaluation of the experiments, and have been verified by various calculation methods for different operating conditions. The qualification of components was verified by strength analyses including a benchmark calculation for specified normal operation and emergencies; the results show a convenient behaviour of the components and their materials. In addition, a fuel element burnup measuring system was designed that applies Cs-137 gamma spectroscopy; its feasibility was checked by appropriate analyses, and qualification work is in progress. The installation of a prototype measurement system is the task for project No. 03 IAT 211. (orig.) [de

  1. The conceptual flowsheet of effluent treatment during preparing spherical fuel elements of HTR

    Energy Technology Data Exchange (ETDEWEB)

    Ying, Quan, E-mail: quanying@tsinghua.edu.cn; Xiao-tong, Chen; Bing, Liu; Gen-na, Fu; Yang, Wang; You-lin, Shao; Zhen-ming, Lu; Ya-ping, Tang; Chun-he, Tang

    2014-05-01

    High temperature gas-cooled reactor (HTR) is one of the advanced nuclear reactors owing to its inherent safety and broad applications. For HTR, one of the key components is the ceramic fuel element. During the preparation of spherical fuel elements, the radioactive effluent treatment is necessary. Referring to the current treatment technologies and methods, the conceptual flowsheet of low-level radioactive effluent treatment during preparing spherical fuel elements was established. According to the above treatment process, the uranium concentration was decreased from 200 mg/l to the level of discharged standard.

  2. Long-term testing of HTR fuel elements in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Nickel, H.

    1986-12-01

    The extensive results from irradiation experiments carried out on coated particles, on graphitic matrices of different composition and on integral fuel elements have shown that the spherical fuel elements with high-enriched uranium/thorium mixed-oxide particles and optimized graphitic matrix are available for use in the planned HTR facilities. A concentrated qualification programme is on the way in order to bring the fuel elements with particles from low-enriched uranium dioxide (LEU) and TRISO coating to a comparable level of experience and knowledge, i.e. to make them licensable for the planned HTR facilities. (orig.) [de

  3. Postirradiation examination of HTR fuel

    International Nuclear Information System (INIS)

    Nabielek, H.; Reitsamer, G.; Kania, M.J.

    1986-01-01

    Fuel for the High Temperature Reactor (HTR) consists of 1 mm diameter coated particles uniformly distributed in a graphite matrix within a cold-molded 60 mm diameter spherical fuel element. Fuel performance demonstrations under simulated normal operation conditions are conducted in accelerated neutron environments available in Material Test Reactors and in real-time environments such as the Arbeitsgemeinschaft Versuchsreaktor (AVR) Juelich. Postirradiation examinations are then used to assess fuel element behavior and the detailed performance of the coated particles. The emphasis in postirradiation examination and accident testing is on assessment of the capability for fuel elements and individual coated particles to retain fission products and actinide fuel materials. To accomplish this task, techniques have been developed which measures fission product and fuel material distributions within or exterior to the particle: Hot Gas Chlorination - provides an accurate method to measure total fuel material concentration outside intact particles; Profile Electrolytic Deconsolidation - permits determination of fission product distribution along fuel element diameter and retrieval of fuel particles from positions within element; Gamma Spectrometry - provides nondestructive method to measure defect particle fractions based on retention of volatile metallic fission products; Particle Cracking - permits a measure of the partitioning of fission products between fuel kernel and particle coatings, and the derivation of diffusion parameters in fuel materials; Micro Gas Analysis - provides gaseous fission product and reactive gas inventory within free volume of single particles; and Mass-spectrometric Burnup Determination - utilizes isotope dilution for the measurement of heavy metal isotope abundances

  4. Symbiosis of near breeder HTR's with hybrid fusion reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1978-07-01

    In this contribution to INFCE a symbiotic fusion/fission reactor system, consisting of a hybrid beam-driven micro-explosion fusion reactor (HMER) and associated high-temperature gas-cooled reactors (HTR) with a coupled fuel cycle, is proposed. This system is similar to the well known Fast Breeder/Near Breeder HTR symbiosis except that the fast fission breeder - running on the U/Pu-cycle in the core and the axial blankets and breeding the surplus fissile material as U-233 in its radial thorium metal or thorium oxide blankets - is replaced by a hybrid micro-explosion DT fusion reactor

  5. Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  6. Prediction calculation of HTR-10 fuel loading for the first criticality

    International Nuclear Information System (INIS)

    Jing Xingqing; Yang Yongwei; Gu Yuxiang; Shan Wenzhi

    2001-01-01

    The 10 MW high temperature gas cooled reactor (HTR-10) was built at Institute of Nuclear Energy Technology, Tsinghua University, and the first criticality was attained in Dec. 2000. The high temperature gas cooled reactor physics simulation code VSOP was used for the prediction of the fuel loading for HTR-10 first criticality. The number of fuel element and graphite element was predicted to provide reference for the first criticality experiment. The prediction calculations toke into account the factors including the double heterogeneity of the fuel element, buckling feedback for the spectrum calculation, the effect of the mixture of the graphite and the fuel element, and the correction of the diffusion coefficients near the upper cavity based on the transport theory. The effects of impurities in the fuel and the graphite element in the core and those in the reflector graphite on the reactivity of the reactor were considered in detail. The first criticality experiment showed that the predicted values and the experiment results were in good agreement with little relative error less than 1%, which means the prediction was successful

  7. HTR-TN achievements and prospects for future developments

    International Nuclear Information System (INIS)

    Hittner, D.; Angulo, C.; Basini, V.; Bogusch, E.; Breuil, E.; Buckthorpe, D.; Chauvet, V.; Futterer, M.A.; Van Heek, A.; Von Lensa, W.; Yvon, P.

    2011-01-01

    It is already 10 years since the (European) High Temperature Reactor Technology Network (HTR-TN) launched a program for development of HTR technology, which expanded through three successive Euratom framework programs, with many projects in line with the network strategy. Widely relying in the beginning on the legacy of the former European HTR developments (DRAGON, AVR, THTR, etc.) that it contributed to safeguard, this program led to advances in HTR/VHTR technologies and produced significant results, which can contribute to the international cooperation through Euratom involvement in the Generation IV International Forum (GIF). the main achievements of the European program, performed in complement to efforts made in several European countries and other GIF partners, are presented: they concern the validation of computer codes (reactor physics, as well as system transient analysis from normal operation to air ingress accident and fuel performance in normal and accident conditions), materials (metallic materials for vessel, direct cycle turbines and intermediate heat exchanger, graphite, etc.), component development, fuel manufacturing and irradiation behavior, and specific HTR waste management (fuel and graphite). Key experiments have been performed or are still ongoing, like irradiation of graphite and of fuel material (PYCASSO experiment), high burn-up fuel PIE, safety test and isotopic analysis, IHX mock-up thermohydraulic test in helium atmosphere, air ingress experiment for a block type core, etc. Now HTR-TN partners consider that it is time for Europe to go a step forward toward industrial demonstration. In line with the orientations of the 'Strategic Energy Technology Plan (SET-Plan)' recently issued by the European Commission that promotes a strategy for development of low-carbon energy technologies and mentions Generation IV nuclear systems as part of key technologies, HTR-TN proposes to launch a program for extending the contribution of nuclear energy to

  8. Accident situations tests HTR fuel with the device Kufa

    International Nuclear Information System (INIS)

    Kellerbauer, A. I.; Freis, D.

    2010-01-01

    The ceramic and ceramic-like coating materials in modern high-temperature reactor fuel are designed to ensure mechanical stability and retention of fission products under normal and transient conditions, regardless of the radiation damage sustained in-pile. In hypothetical depressurization and loss-of-forced-circulation (D LOFC) accidents, fuel elements of modular high-temperate reactors are exposed to temperatures several hundred degrees higher than during normal operation, causing increased thermo-mechanical stress on the coating layers. At the Institute for Transuranium Elements of the European Commission, a vigorous experimental program is being pursued with the aim of characterizing the performance of irradiated HTR fuel under such accident conditions. A cold finger device (Kufa), operational in ITUs hot cells since 2006, has been used to perform heating experiments on eight irradiated HTR fuel pebbles from the AVR experimental reactor and from dedicated irradiation campaigns at the High-Flux Reactor in Petten, the Netherlands. Gaseous fission products are collected in a cryogenic charcoal trap, while volatiles,are plated out on a water-cooled condensate plate. A quantitative measurement of the release is obtained by gamma spectroscopy. We highlight experimental results from the Kufa testing as well as the on-going development of new experimental facilities. (Author) 9 refs.

  9. Experiments in MARIUS on HTR tubular fuel with loose particles

    Energy Technology Data Exchange (ETDEWEB)

    Bosser, R; Langlet, G

    1972-06-15

    The work described on HTR tubular fuel with loose particles is the first part of a program in three points. The cell is the same in the three experiments, only particles in the fuel container are changed. The aim of the experiment is to achieve the buckling in a critical facility. A description of the techniques of measurements, calculations, and results are presented.

  10. Performance limits of coated particle fuel. Part III. Fission product migration in HTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nabielek, H.; Hick, H.; Wagner-Loffler, M.; Voice, E. H.

    1974-06-15

    A general introduction and literature survey to the physics and mathematics of fission product migration in HTR fuel is given as well as a review of available experimental results and their evaluation in terms of models and materials data.

  11. The HTR-10 project and its further development

    International Nuclear Information System (INIS)

    Xu Yuanhui

    2002-01-01

    The 10 MW High Temperature Gas-cooled Reactor-Test Module (termed as HTR-10) is one of key project in the National High Technology Research and Development Program (1986-2000). Main objectives for the HTR-10 are: (1). To acquire know-how to design, construct and operate the HTGRs, (2). To establish an experimental facility, (3). To demonstrate the inherent safety features of the Modular HTGR, (4). To test electricity and heat co-generation and closed cycle gas turbine technology and (5). To do research and development work for high temperature process heat application. The Institute of Nuclear Energy Technology (INET) of Tsinghua University was appointed as the leading institute to be responsible for design, license applications, construction and operation of the HTR-10. The HTR-10 technical design represents the features of HTR-Module design. After five years construction, installation and pre-operation the HTR-10 reached the criticality in December 2000. Up to now all of results on zero point experiments and fuel elements irradiation test are fine. China will continue to develop the high temperature gas-cooled reactor in the future using the HTR-10 base

  12. Maw and spent HTR Fuel Element Test storage in Boreholes in rock salt

    International Nuclear Information System (INIS)

    Barnert, E.; Brucher, P.H.; Kroth, K.; Merz, E.; Niephaus, D.

    1986-01-01

    The Budesminister fur Forschung und Technolgie (BMFT, Federal Ministry for Research and Technology) is sponsoring a project at the Kernforschungsanlage Julich (KFA, Juelich Nuclear Research Centre) entitled ''MAW and HTR Fuel Element Test disposal in Boreholes.'' The aim of this project is to develop a technique for the final disposal of (1) dissolver sludge, (2) cladding hulls/structural components and (3) spent HTR fuels elements in salt, and to test this technique in the abandoned Asse salt mine, including safety calculations and safety engineering demonstrations. The project is divided into the sub-projects I ''Disposal/sealing technique'' and II ''Retrievable disposal test.''

  13. Nuclear power and the possibility of alternative fuel cycles

    International Nuclear Information System (INIS)

    Engelmann, P.

    1979-01-01

    Concern about the societal implications, potential risks and the possibility of nuclear weapons proliferation has slowed down the growth of nuclear energy. Assuming a further moderate growth of nuclear power in the Federal Republic of Germany several fuel cycle and reactor strategies can the followed without exhausting the nuclear the resources before the year 2100. The uranium demand of various reactor strategies with LWR's FBR's and HTR's is compared for two demand cases in the FRG. While recycling of spent fuel seems necessary in any case, it is shown that the Th/U cycle can provide a realistic alternative to the U/Pu cycle. The parallel introduction of both cycles appears as the best solution, as it reduces the overall risks and leads to minimum uranium demand. The risk of nuclear proliferation does not vary considerably with the fuel cycle applied; it can, however, be reduced to acceptable levels by safeguards methods and institutional means. (orig.) [de

  14. Studies on equilibrium fuel management schemes on the Dragon HTR core design

    Energy Technology Data Exchange (ETDEWEB)

    Daub, J; Pedersen, J

    1971-02-03

    The Dragon Project has recently started investigations on fuel management in HTR's with the assumed Dragon design. The study covers the results of investigations into a number of equilibrium fuel management schemes with the 1-dimensional FLATTER code and calculations of the corresponding total power generating costs with the programme TECO.

  15. Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

    International Nuclear Information System (INIS)

    Reitsamer, G.; Proksch, E.; Stolba, G.; Strigl, A.; Falta, G.; Zeger, J.

    1985-01-01

    Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e. fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations

  16. Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

    Energy Technology Data Exchange (ETDEWEB)

    Reitsamer, G; Proksch, E; Stolba, G; Strigl, A; Falta, G; Zeger, J [Department of Chemistry, Austrian Research Centre Seibersdorf Ltd., Seibersdorf (Austria)

    1985-07-01

    Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e., fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations.

  17. Testing of HTR UO{sub 2} TRISO fuels in AVR and in material test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kania, Michael J., E-mail: MichaelJKania@googlemail.com [Retired from Lockheed Martin Corp, 20 Beach Road, Averill Park, NY 12018 (United States); Nabielek, Heinz, E-mail: heinznabielek@me.com [Retired from Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Verfondern, Karl [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Allelein, Hans-Josef [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, 52072 Aachen (Germany)

    2013-10-15

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO{sub 2} TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO{sub 2} TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO{sub 2} TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C.

  18. HTR-2002: Proceedings of the conference on high temperature reactors

    International Nuclear Information System (INIS)

    2002-01-01

    High temperature reactors are considered as future inherently safe and efficient energy sources. The presentations covered all the relevant aspects of the existing HTGRs and/or helium cooled pebble bed reactors. They were sorted into 7 sessions: HTR Projects and Programmes; Fuel and Fuel Cycle; Physics and Neutronics; Thermohydraulic Calculation; Engineering, Design and Applications; Materials and Components; Safety and Licensing

  19. Automatic X-ray inspection for the HTR-PM spherical fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Yi, DU, E-mail: duyi11@mails.tsinghua.edu.cn [Institute of Nuclear and New Energy Technology (INET), Tsinghua University, Energy Science Building A309, Haidian District, Beijing 100084 (China); Xiangang, WANG, E-mail: wangxiangang@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology (INET), Tsinghua University, Energy Science Building A309, Haidian District, Beijing 100084 (China); Xincheng, XIANG, E-mail: inetxxc@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology (INET), Tsinghua University, Energy Science Building, Haidian District, Beijing 100084 (China); Bing, LIU, E-mail: bingliu@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology (INET), Tsinghua University, Energy Science Building, Haidian District, Beijing 100084 (China)

    2014-12-15

    Highlights: • An automatic X-ray inspection method is established to characterize HTR pebbles. • The method provides physical characterization and the inner structure of pebbles. • The method can be conducted non-destructively, quickly and automatically. • Sample pebbles were measured with this AXI method for validation. • The method shows the potential to be applied in situ. - Abstract: Inefficient quality assessment and control (QA and C) of spherical fuel elements for high temperature reactor-pebblebed modules (HTR-PM) has been a long-term problem, since conventional methods are labor intensive and cannot reveal the inside information nondestructively. Herein, we proposed a nondestructive, automated X-ray inspection (AXI) method to characterize spherical fuel elements including their inner structures based on X-ray digital radiography (DR). Briefly, DR images at different angles are first obtained and then the chosen important parameters such as spherical diameters, geometric and mass centers, can be automatically extracted and calculated via image processing techniques. Via evaluating sample spherical fuel elements, we proved that this AXI method can be conducted non-destructively, quickly and automatically. This method not only provides accurate physical characterization of spherical fuel elements but also reveals their inner structure with good resolution, showing great potentials to facilitate fast QA and C in HTM-PM spherical fuel element development and production.

  20. Automatic X-ray inspection for the HTR-PM spherical fuel elements

    International Nuclear Information System (INIS)

    Yi, DU; Xiangang, WANG; Xincheng, XIANG; Bing, LIU

    2014-01-01

    Highlights: • An automatic X-ray inspection method is established to characterize HTR pebbles. • The method provides physical characterization and the inner structure of pebbles. • The method can be conducted non-destructively, quickly and automatically. • Sample pebbles were measured with this AXI method for validation. • The method shows the potential to be applied in situ. - Abstract: Inefficient quality assessment and control (QA and C) of spherical fuel elements for high temperature reactor-pebblebed modules (HTR-PM) has been a long-term problem, since conventional methods are labor intensive and cannot reveal the inside information nondestructively. Herein, we proposed a nondestructive, automated X-ray inspection (AXI) method to characterize spherical fuel elements including their inner structures based on X-ray digital radiography (DR). Briefly, DR images at different angles are first obtained and then the chosen important parameters such as spherical diameters, geometric and mass centers, can be automatically extracted and calculated via image processing techniques. Via evaluating sample spherical fuel elements, we proved that this AXI method can be conducted non-destructively, quickly and automatically. This method not only provides accurate physical characterization of spherical fuel elements but also reveals their inner structure with good resolution, showing great potentials to facilitate fast QA and C in HTM-PM spherical fuel element development and production

  1. The low enriched fuel cycle in the GA 1160 MW design and the switch-over to thorium

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, H.

    1974-03-15

    Calculations for the GA 1160 MW HTR are presented. The aim of these investigations was to compare the Low Enriched Uranium (LEU) cycle and the Thorium cycle for the GA 1160 MW HTR both using the same GA designed integral block fuel element. The total fuel cycle cost for the equilibrium cycle comes out to be about 16% cheaper for the Thorium cycle than for the Low-Enriched cycle. However, these favorable results for the thorium cycle are completely dependent on the availability of reprocessing and refabrication facilities, for costs comparable with the costs used for these investigations. The possibility of starting the reactor on a LEU 3 year cycle and later switching over to a thorium 4 year cycle was investigated. No cost penalties were found to be paid during the switch-over. The problems of local power peaks and age factors were not investigated in greater detail as only integral physical quantities were obtained from the neutron physics calculations. However, no indications of any problem in the switch-over phase were given. Elaborate 3-dimensional methods are necessary for further investigation of these types of problems.

  2. Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Verfondern, K.; Mueller, D.

    1991-01-01

    Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

  3. Temperature Analysis and Failure Probability of the Fuel Element in HTR-PM

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Tang Chunhe

    2014-01-01

    Spherical fuel element is applied in the 200-MW High Temperature Reactor-Pebble-bed Modular (HTR-PM). Each spherical fuel element contains approximately 12,000 coated fuel particles in the inner graphite matrix with a diameter of 50mm to form the fuel zone, while the outer shell with a thickness of 5mm is a fuel-free zone made up of the same graphite material. Under high burnup irradiation, the temperature of fuel element rises and the stress will result in the damage of fuel element. The purpose of this study is to analyze the temperature of fuel element and to discuss the stress and failure probability. (author)

  4. Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010

    International Nuclear Information System (INIS)

    Snead, Lance Lewis; Besmann, Theodore M.; Collins, Emory D.; Bell, Gary L.

    2010-01-01

    The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

  5. Present status of research and development for HTR in China

    Energy Technology Data Exchange (ETDEWEB)

    Dazhong, Wang; Daxin, Zhong; Yuanhul, Xu [Institute of Nuclear Energy Technology, Tsinghua University, Beijing (China)

    1990-07-01

    The HTR R and D Project is being carried out in the relevant institutions in China. Some topics are covered such as, fuel element technology, graphite development, fuel element handling system, helium technology, fuel reprocessing technology as well as HTR design study. Some results of HTR research work are described. In addition, to provide a test facility for investigation of HTR Module reactor safety and process heat application of HTR, a joint project on building a 10 MW test HTR with Siemens-Interatom, KFA Juelich and INET is going on. The conceptual design of 10 MW test HTR has been completed by the joint group. In parallel the application study of HTR Module is being carried out for the oil industry, petrochemical industry as well as power generation. Some preliminary results of the application study, for example, for heavy oil recovery on Shengli oil field and process heat application in Yan shan petroleum company, are described. (author)

  6. Fuel cycle economics of HTRs

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U.

    1975-06-15

    The High Temperature Reactor commands a unique fuel cycle flexibility and alternative options are open to the utilities. The reference thorium reactor operating in the U-233 recycle mode is 10 to 20% cheaper than the low-enriched reactor; however, the thorium cycle depends on the supply of 93% enriched uranium and the availability of reprocessing and refabrication facilities to utilize its bred fissile material. The economic landscape towards the end of the 20th Century will presumably be dominated by pronounced increases in the costs of natural resources. In the case of nuclear energy, resource considerations are reflected in the price of uranium, which is expected Lo have reached 50 $/lbm U3O8 in the early 1990s and around 100 $/lbm U3O8 around 2010. In this economic environment the fuel cycle advantage of the thorium system amounts to some 20% and is capable of absorbing substantial expenses in bringing about the closing of the out-of-pile cycle. A most attractive aspect of the HTR fuel cycle flexibility is for the utility to start operating the reactor on the low enriched uranium cycle and at a later date switch over to the thorium cycle as this becomes economically more and more attractive. The incentive amounts to some 50 M$ in terms of present worth money at the time of decision making, assumed to take place 10 years after start-up. The closing of the thorium cycle is of paramount importance and a step to realize this objective lies in simplifying the head-end reprocessing technology by abandoning the segregation concept of feed and breed coated particles in the reference cycle. A one-coated-particle scheme in which all discharged uranium isotopes are recycled in mixed oxide particles is feasible and suffers a very minor economic penalty only.

  7. Assessment of nuclear fuel cycles with respect to assurance of energy supply; economic aspects; environmental aspects; non-proliferation

    International Nuclear Information System (INIS)

    1979-01-01

    This paper, which was presented to all INFCE Working Groups gives a broad qualitative assessment in tabular form of the following five fuel cycles: LWR once-through, LWR with thermal recycle, HWR once-through, HTR with uranium recycle, fast breeder reactor. The assessment is given of the assurance of supply aspects, the macro- and micro-economic aspects, the environmental aspects, and the non-proliferation, including safeguards, aspects of each fuel cycle

  8. Irradiation behaviour of advanced fuel elements for the helium-cooled high temperature reactor (HTR)

    International Nuclear Information System (INIS)

    Nickel, H.

    1990-05-01

    The design of modern HTRs is based on high quality fuel. A research and development programme has demonstrated the satisfactory performance in fuel manufacturing, irradiation testing and accident condition testing of irradiated fuel elements. This report describes the fuel particles with their low-enriched UO 2 kernels and TRISO coating, i.e. a sequence of pyrocarbon, silicon carbide, and pyrocarbon coating layers, as well as the spherical fuel element. Testing was performed in a generic programme satisfying the requirements of both the HTR-MODUL and the HTR 500. With a coating failure fraction less than 2x10 -5 at the 95% confidence level, the results of the irradiation experiments surpassed the design targets. Maximum accident temperatures in small, modular HTRs remain below 1600deg C, even in the case of unrestricted core heatup after depressurization. Here, it was demonstrated that modern TRISO fuels retain all safety-relevant fission products and that the fuel does not suffer irreversible changes. Isothermal heating tests have been extended to 1800deg C to show performance margins. Ramp tests to 2500deg C demonstrate the limits of present fuel materials. A long-term programm is planned to improve the statistical significance of presently available results and to narrow remaining uncertainty limits. (orig.) [de

  9. CEA and AREVA R and D on V/HTR fuel fabrication with the CAPRI experimental manufacturing line

    International Nuclear Information System (INIS)

    Charollais, Francois; Fonquernie, Sophie; Perrais, Christophe; Perez, Marc; Cellier, Francois; Vitali, Marie-Pierre

    2006-01-01

    In the framework of the French V/HTR fuel development and qualification program, the Commissariat a l'Energie Atomique (CEA) and AREVA through its program called ANTARES (Areva New Technology for Advanced Reactor Energy Supply) conduct R and D projects covering the mastering of UO 2 coated particle and fuel compact fabrication technology. To fulfill this task, a review of past knowledge, of existing technologies and a preliminary laboratory scale work program have been conducted with the aim of retrieving the know-how on HTR coated particle and compact manufacture: - The different stages of UO 2 kernel fabrication GSP Sol-Gel process have been reviewed, reproduced and improved; - The experimental conditions for the chemical vapour deposition (CVD) of coatings have been defined on dummy kernels and development of innovative characterization methods has been carried out; - Former CERCA compacting process has been reviewed and updated. In parallel, an experimental manufacturing line for coated particles, named GAIA, and a compacting line based on former CERCA compacting experience have been designed, constructed and are in operation since early 2005 at CEA Cadarache and CERCA Romans, respectively. These two facilities constitute the CAPRI line (CEA and AREVA PRoduction Integrated line). The major objectives of the CAPRI line are: - to recover and validate past knowledge; - to permit the optimisation of reference fabrication processes for kernels and coatings and the investigation of alternative and innovative fuel design (UCO kernel, ZrC coating); - to test alternative compact process options; - to fabricate and characterize fuel required for irradiation and qualification purpose; - to specify needs for the fabrication of representative V/HTR TRISO fuel meeting industrial standards. This paper presents the progress status of the R and D conducted on V/HTR fuel particle and compact manufacture by mid 2005. (authors)

  10. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brian Boer; Abderrafi M. Ougouag

    2011-03-01

    significant failure is to be expected for the reference fuel particle during normal operation. It was found, however, that the sensitivity of the coating stress to the CO production in the kernel was large. The CO production is expected to be higher in DB fuel than in UO2 fuel, but its exact level has a high uncertainty. Furthermore, in the fuel performance analysis transient conditions were not yet taken into account. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high transuranic [TRU] content and high burn-up). Accomplishments of this work include: •Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel. •Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel and Uranium. •Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel and Modified Open Cycle Components. •Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel and Americium targets.

  11. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    International Nuclear Information System (INIS)

    Boer, Brian; Ougouag, Abderrafi M.

    2011-01-01

    failure is to be expected for the reference fuel particle during normal operation. It was found, however, that the sensitivity of the coating stress to the CO production in the kernel was large. The CO production is expected to be higher in DB fuel than in UO2 fuel, but its exact level has a high uncertainty. Furthermore, in the fuel performance analysis transient conditions were not yet taken into account. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high transuranic (TRU) content and high burn-up). Accomplishments of this work include: (1) Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel. (2) Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel and Uranium. (3) Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel and Modified Open Cycle Components. (4) Core analysis of a HTR-MODULE design loaded with Deep-Burn fuel and Americium targets.

  12. Oxidation of carbon based material for innovative energy systems (HTR, fusion reactor): status and further needs

    International Nuclear Information System (INIS)

    Moormann, R.; Hinssen, H.K.; Latge, Ch.; Dumesnil, J.; Veltkamp, A.C.; Grabon, V.; Beech, D.; Buckthorpe, D.; Dominguez, T.; Krussenberg, A.K.; Wu, C.H.

    2000-01-01

    Following an overview on kinetics of carbon/gas reactions, status and further needs in selected safety relevant fields of graphite oxidation in high temperature reactors (HTRs) and fusion reactors are outlined. Kinetics was detected due to the presence of such elements as severe air ingress, lack of experimental data on Boudouard reaction and a similar lack of data in the field of advanced oxidation. The development of coatings which protect against oxidation should focus on stability under neutron irradiation and on the general feasibility of coatings on HTR pebble fuel graphite. Oxidation under normal operation of direct cycle HTR requires examinations of gas atmospheres and of catalytic effects. Advanced carbon materials like CFCs and mixed materials should be developed and tested with respect to their oxidation resistance in a common HTR/fusion task. In an interim HTR, fuel storage radiolytic oxidation under normal operation and thermal oxidation in accidents have to be considered. Plans for future work in these fields are described. (authors)

  13. Costs of head-end incineration with respect to Kr separation in the reprocessing of HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Barnert-Wiemer, H.; Boehnert, R.

    1976-07-15

    The C-incinerations and the Kr-separations during head-end incineration in the reprocessing of HTR fuel elements are described. The costs for constructing an operating a head-end incineration of reprocessing capacities with 5,000 to 50,000 MW(e)-HTR power have been determined. The cost estimates are divided into investment and operating costs, further after the fraction of the N/sub 2/-content in the incineration exhaust gas, which strongly affects costs. It appears that, in the case of Kr-separation from the incineration exhaust gas, the investment costs as well as the operating costs of the head-end for N/sub 2/-containing exhaust gas are considerably greater than those for gas without N/sub 2/. The C-incineration of the graphite of the HTR fuel elements should therefore only be performed with influx gas that is free of N/sub 2/.

  14. Assessment of the thorium and uranium fuel cycle in the fast breeder and the high temperature reactor

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    1977-01-01

    This report assesses the fissile fuel economy of the uranium and thorium cycle in the advanced reactors currently under development, the fast breeder reactor (FBR) and the high temperature reactor (HTR). It is shown by means of detailed burnup calculations that replacing UO 2 with ThO 2 or Th-metal as the radial blanket breeding material will not have any significant imapct on the breeding and burnup properties of the FBR. A global, analytical investigation is performed to study the fissile fuel economy of the many fissile fuel cycles possible in the HTR. Here it is demonstrated that the optimum conversion ratio of CR 3 O 8 ) demands are evaluated for a country such as the FRG under the assumptions of different future reactor strategy scenarios. Here it is demonstrated that the employement of both HTRs and FBRs can lead to a practically resource independent energy supply system within the next 40 to 60 years. However only through the large scale employement of the fast breeder can the future nuclear resource requirements be assured. (orig.) [de

  15. MCNP qualification on the HTR critical configurations: HTTR, HTR10 and PROTEUS results

    Energy Technology Data Exchange (ETDEWEB)

    TRAKAS, Christos; STOVEN, Gilles [AREVA NP, Tour Areva, 92084 Paris La Defence Cedex (France)

    2008-07-01

    Recent critical experiments, including PROTEUS, HTTR and HTR-10 provide a reliable qualification base for HTR criticality predictions. The fuel tested in these experiments, be it hexagonal block or pebble type, is irradiated in a spectrum comparable to that of the HTR planned by AREVA NP. The neutron spectrum is comparable in all three cases; the mean C/M value for all critical cases is less than +350 pcm (JEF2.2), +250 pcm (JEFF3.1) and +60 pcm (ENDF BVI). The C/M obtained for the rods worth, the reaction rates and the isothermal coefficient are very satisfactory. (authors)

  16. Weighting comparison of the production of α-emitters in HTR-fuel cycles

    International Nuclear Information System (INIS)

    Rueckert, M.; Hecker, R.; Migenda, J.; Mirza, N.

    1976-03-01

    This study compares three fuel-cycles of the OTTO-Pebble-Bed High-Temperature Reactor to each other and as a reference to a Light-Water Reactor with regard to the 'Theoretical Environmental Risk' of their heavy-metal production. For the weighting of the α-activity those water- and air masses are used, which according to the recommendations of the 'International Committee for Radiation Protection' are necessary for the dilution of radioactive substances in the cycle. The Th/U-cycles are better than the U/Pu-cycles; one has to put into account, however, that in the Th/U-cycles the Ra which comes up after long time period increases the 'Theoretical Environmental Risk'. (orig.) [de

  17. Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

    International Nuclear Information System (INIS)

    Reitsamer, G.; Proksch, E.; Stolba, G.; Strigl, A.; Falta, G.; Zeger, J.

    1984-02-01

    This paper describes methods and measurements developed at the Austrian Research Centre Seibersdorf for the evaluation of the irradiation performance of HTR fuel. Main interest is concentrated on particle failure rates, fission product release, burn-up and inventory measurements (solid and gaseous fission products, uranium inventory). (Author) [de

  18. Advanced Characterization Techniques for Silicon Carbide and Pyrocarbon Coatings on Fuel Particles for High Temperature Reactors (HTR)

    Energy Technology Data Exchange (ETDEWEB)

    Basini, V.; Charollais, F. [CEA Cadarache, DEN/DEC/SPUA, BP 1, 13108 St Paul Lez Durance (France); Dugne, O. [CEA Marcoule, DEN/DTEC/SCGS BP 17171 30207 Bagnols sur Ceze (France); Garcia, C. [Laboratoire des Composites Thermostructuraux (LCTS), UMR CNRS 5801, 3 allee de La Boetie, 33600 Pessac (France); Perez, M. [CEA Grenoble DRT/DTH/LTH, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)

    2008-07-01

    Cea and AREVA NP have engaged an extensive research and development program on HTR (high temperature reactor) fuel. The improving of safety of (very) high temperature reactors (V/HTR) is based on the quality of the fuel particles. This requires a good knowledge of the properties of the four-layers TRISO particles designed to retain the uranium and fission products during irradiation or accident conditions. The aim of this work is to characterize exhaustively the structure and the thermomechanical properties of each unirradiated layer (silicon carbide and pyrocarbon coatings) by electron microscopy (SEM, TEM), selected area electronic diffraction (SEAD), thermo reflectance microscopy and nano-indentation. The long term objective of this study is to define pertinent parameters for fuel performance codes used to better understand the thermomechanical behaviour of the coated particles. (authors)

  19. HTR plus modern turbine technology for higher efficiencies

    International Nuclear Information System (INIS)

    Barnert, H.; Kugeler, K.

    1996-01-01

    The recent efficiency race for natural gas fired power plants with gas-plus steam-turbine-cycle, is shortly reviewed. The question 'can the HTR compete with high efficiencies?' is answered: Yes, it can - in principle. The gas-plus steam-turbine cycle, also called combi-cycle, is proposed to be taken into consideration here. A comparative study on the efficiency potential is made; it yields 54.5% at 1,050 deg. C gas turbine-inlet temperature. The mechanisms of release versus temperature in the HTR are summarized from the safety report of the HTR MODUL. A short reference is made to the experiences from the HTR-Helium Turbine Project HHT, which was performed in the Federal Republic of Germany in 1968 to 1981. (author). 8 figs,. 1 tab

  20. HTR plus modern turbine technology for higher efficiencies

    Energy Technology Data Exchange (ETDEWEB)

    Barnert, H; Kugeler, K [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

    1996-08-01

    The recent efficiency race for natural gas fired power plants with gas-plus steam-turbine-cycle, is shortly reviewed. The question `can the HTR compete with high efficiencies?` is answered: Yes, it can - in principle. The gas-plus steam-turbine cycle, also called combi-cycle, is proposed to be taken into consideration here. A comparative study on the efficiency potential is made; it yields 54.5% at 1,050 deg. C gas turbine-inlet temperature. The mechanisms of release versus temperature in the HTR are summarized from the safety report of the HTR MODUL. A short reference is made to the experiences from the HTR-Helium Turbine Project HHT, which was performed in the Federal Republic of Germany in 1968 to 1981. (author). 8 figs,. 1 tab.

  1. On Power Refueling Management of HTR-PM

    International Nuclear Information System (INIS)

    Sun Furui; Luo Yong; Gao Qiang

    2014-01-01

    The refueling management is an important work of nuclear power plant , directly affecting its safety and economy. At present, the ordinary commercial pressurized water reactor (PWR) nuclear power plant has developed more mature in the refueling management, and formed a set of relatively complete system and methods.The High Temperature Gas-cooled Reactor Pebble-modules Demonstration Project(HTR-PM) has significant differences with the ordinary PWR nuclear power plant in the fuel morphology and the refueling mode. It adopts the spherical fuel element and the on-power refueling. Therefore, the HTR-PM refueling management has its own unique characteristics, but currently there is no mature experience to use for reference across the world. This paper gives a brief introduction to the HTR-PM on power refueling management, including the refueling management system construction, the refueling strategy, the fuel element internal transportation,charging and discharging, etc. It aims at finding the befitting HTR-PM refueling management methods in view of its own unique characteristics in order to ensure the orderly development of the refueling management and the refueling safety. (author)

  2. Licensing experience of the HTR-10 test reactor

    International Nuclear Information System (INIS)

    Sun, Y.; Xu, Y.

    1996-01-01

    A 10MW high temperature gas-cooled test reactor (HTR-10) is now being projected by the Institute of Nuclear Energy Technology within China's National High Technology Programme. The Construction Permit of HTR-10 was issued by the Chinese nuclear licensing authority around the end of 1994 after a period of about one year of safety review of the reactor design. HTR-10 is the first high temperature gas-cooled reactor (HTGR) to be constructed in China. The purpose of this test reactor project is to test and demonstrate the technology and safety features of the advanced modular high temperature reactor design. The reactor uses spherical fuel elements with coated fuel particles. The reactor unit and the steam generator unit are arranged in a ''side-by-side'' way. Maximum fuel temperature under the accident condition of a complete loss of coolant is limited to values much lower than the safety limit set for the fuel element. Since the philosophy of the technical and safety design of HTR-10 comes from the high temperature modular reactor design, the reactor is also called the Test Module. HTR-10 represents among others also a licensing challenge. On the one side, it is the first helium reactor in China, and there are less licensing experiences both for the regulator and for the designer. On the other side, the reactor design incorporates many advanced design features in the direction of passive or inherent safety, and it is presently a world-wide issue how to treat properly the passive or inherent safety design features in the licensing safety review. In this presentation, the licensing criteria of HTR-10 are discussed. The organization and activities of the safety review for the construction permit licensing are described. Some of the main safety issues in the licensing procedure are addressed. Among these are, for example, fuel element behaviour, source term, safety classification of systems and components, containment design. The licensing experiences of HTR-10 are of

  3. DEM simulation of particle mixing for optimizing the overcoating drum in HTR fuel fabrication

    Science.gov (United States)

    Liu, Malin; Lu, Zhengming; Liu, Bing; Shao, Youlin

    2013-06-01

    The rotating drum was used for overcoating coated fuel particles in HTR fuel fabrication process. All the coated particles should be adhered to equal amount of graphite powder, which means that the particle should be mixed quickly in both radial and axial directions. This paper investigated the particle flow dynamics and mixing behavior in different regimes using the discrete element method (DEM). By varying the rotation speed, different flow regimes such as slumping, rolling, cascading, cataracting, centrifuging were produced. The mixing entropy based on radial and axial grid was introduced to describe the radial and axial mixing behaviors. From simulation results, it was found that the radial mixing can be achieved in the cascading regime more quickly than the slumping, rolling and centrifuging regimes, but the traditional rotating drum without internal components can not achieve the requirements of axial mixing and should be improved. Three different structures of internal components are proposed and simulated. The new V-shaped deflectors were found to achieve a quick axial mixing behavior and uniform axial distribution in the rotating drum based on simulation results. At last, the superiority was validated by experimental results, and the new V-shaped deflectors were used in the industrial production of the overcoating coated fuel particles in HTR fuel fabrication process.

  4. Thorium utilisation in a small long-life HTR. Part III: Composite-rod fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Verrue, Jacques, E-mail: jacques.verrue@polytechnique.org [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); École Polytechnique (Member of ParisTech), 91128 Palaiseau Cedex (France); Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-02-15

    Highlights: • Composite-rod fuel blocks are proposed for a small block-type HTR. • An axial separation of fuel compacts is the most important feature. • Three patterns are presented to analyse the effects of the spatial distribution. • The spatial distribution has a large influence on the neutron spectrum. • Composite-rod fuel blocks reach a reactivity swing less than 4%. - Abstract: The U-Battery is a small long-life high temperature gas-cooled reactor (HTR) with power of 20 MWth. In order to increase its lifetime and diminish its reactivity swing, the concept of composite-rod fuel blocks with uranium and thorium was investigated. Composite-rod fuel blocks feature a specific axial separation between UO{sub 2} and ThO{sub 2} compacts in fuel rods. The design parameters, investigated by SCALE 6, include the number and spatial distribution of fuel compacts within the rods, the enrichment of uranium, the radii of fuel kernels and fuel compacts, and the packing fractions of uranium and thorium TRISO particles. The analysis shows that a lower moderation ratio and a larger inventory of heavy metals results in a lower reactivity swing. The optimal atomic carbon-to-heavy metal ratio depends on the mass fraction of U-235 and is commonly in the 160–200 range. The spatial distribution of the fuel compacts within the fuel rods has a large influence on the energy spectrum in each fuel compact and thus on the beginning-of-life reactivity and the reactivity swing. At end-of-life, the differences caused by the spatial distribution of the fuel compacts are smaller due to the fissions of U-233 in the ThO{sub 2} fuel compacts. This phenomenon enables to design fuel blocks with a very low reactivity swing, down to less than 4% in a 10-year lifetime. Among three types of thorium fuelled U-Battery blocks, the composite-rod fuel block achieves the highest end-of-life reactivity and the lowest reactivity swing.

  5. Gas reactor international cooperative program. HTR-synfuel application assessment

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H/sub 2/ +CO/sub 2/) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000/sup 0/F steam is generated at the industrial user sites. The products of methanation (CH/sub 4/ + H/sub 2/O) are piped back to the reformer at the central station HTR.

  6. Gas reactor international cooperative program. HTR-synfuel application assessment

    International Nuclear Information System (INIS)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H 2 +CO 2 ) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000 0 F steam is generated at the industrial user sites. The products of methanation (CH 4 + H 2 O) are piped back to the reformer at the central station HTR

  7. A subroutine for the calculation of resonance cross sections of U-238 in HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Cuniberti, R; Marullo, G C

    1971-02-15

    In this paper, a survey of the codes used at Ispra for the calculations of resonance absorption in HTR fuel elements is presented and a subroutine for the calculation of resonance cross-sections, in a seven groups energy structure, for a HTR lattice of annular type is described. A library of homogeneous resonance integrals and a wide tabulation of lump and kernel Bell factors, and moderators efficiency is given. This paper deals mainly with the problem of taking into account the correct slowing down of neutrons in the graphite and with the derivation of Bell factors to be used in a multigroup calculation scheme.

  8. HTR-10 severe accident management

    International Nuclear Information System (INIS)

    Xu Yuanhui; Sun Yuliang

    1997-01-01

    The High Temperature Gas-cooled Reactor (HTR-10) is under construction at the Institute of Nuclear Energy Technology site northwest of Beijing. This 10 MW thermal plant utilizes a pebble bed high temperature gas cooled reactor for a large range of applications such as electricity generation, steam and district heat generation, gas turbine and steam turbine combined cycle and process heat for methane reforming. The HTR-10 is the first high temperature gas cooled reactor to be licensed in China. This paper describes the safety characteristics and design criteria for the HTR-10 as well as the accident management and analysis required for the licensing process. (author)

  9. Thorium utilization in a small long-life HTR. Part II: Seed-and-blanket fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Ming, E-mail: dingming@hrbeu.edu.cn [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-02-15

    Highlights: • Seed-and-blanket (S and B) fuel blocks are proposed for a small block-type HTR. • S and B fuel blocks consist of a seed region (UO{sub 2}) and a blanket region (ThO{sub 2}). • The neutronic performance of S and B fuel blocks are analyzed using SCALE 6. • Three S and B fuel blocks with a reactivity swing of 0.1 Δk are recommended. • S and B fuel blocks are compared with thorium MOX fuel blocks. - Abstract: In order to utilize thorium in high temperature gas-cooled reactors (HTRs), the concept of seed-and-blanket (S and B) fuel block is introduced into the U-Battery, which is a long-life block-type HTR with a thermal power of 20 MWth. A S and B fuel block consists of a seed region with uranium in the center, and a blanket region with thorium. The neutronic performance, such as the multiplication factor, conversion ratio and reactivity swing, of a typical S and B fuel block was investigated by SCALE 6.0 by parametric analysis of the composition parameters and geometric parameters of the fuel block for the U-Battery application. Since the purpose of U-235 in the S and B fuel block is to ignite the fission reactions in the fuel block, 20% enriched uranium is recommended for the S and B fuel block. When the ratio of the number of carbon to heavy metal atoms changes with the geometric parameters of the fuel block in the range of 200–250, the reactivity swing reaches very small values. Furthermore, for a reactivity swing of 0.1 Δk during 10 effective full power years, three configurations with 36, 54 and 78 UO{sub 2} fuel rods are recommended for the application of the U-Battery. The comparison analysis of the S and B fuel block with the Th/U MOX fuel block shows that the former has a longer lifetime and a lower reactivity swing.

  10. Comparison of the thorium- and low-enriched uranium fuel cycle in the OTTO pebble bed HTR

    Energy Technology Data Exchange (ETDEWEB)

    Teuchert, E; Maly, V

    1973-03-15

    From the study of the physical, technical and economical properties of the LOTTO and TOTTO fuel cycle the authors draw the conclusion that no fundamental reason can be found for a definite preference of one of the two cycles. Either of them can be developed to become a very attractive concept. The flexibility of this reactor allows the design performance in that way that the reactor becomes a safe system with reasonable economy. The decision for the preference of the LOTTO or TOTTO can be allowed to be governed by the requirements of the outer fuel cycle or by politics.

  11. Potentialities of high temperature reactors (HTR)

    International Nuclear Information System (INIS)

    Hittner, D.

    2001-01-01

    This articles reviews the assets of high temperature reactors concerning the amount of radioactive wastes produced. 2 factors favors HTR-type reactors: high thermal efficiency and high burn-ups. The high thermal efficiency is due to the high temperature of the coolant, in the case of the GT-MHR project (a cooperation between General Atomic, Minatom, Framatome, and Fuji Electric) designed to burn Russian military plutonium, the expected yield will be 47% with an outlet helium temperature of 850 Celsius degrees. The high temperature of the coolant favors a lot of uses of the heat generated by the reactor: urban heating, chemical processes, or desalination of sea water.The use of a HTR-type reactor in a co-generating way can value up to 90% of the energy produced. The high burn-up is due to the technology of HTR-type fuel that is based on encapsulation of fuel balls with heat-resisting materials. The nuclear fuel of Fort-Saint-Vrain unit (Usa) has reached values of burn-ups from 100.000 to 120.000 MWj/t. It is shown that the quantity of unloaded spent fuel can be divided by 4 for the same amount of electricity produced, in the case of the GT-MHR project in comparison with a light water reactor. (A.C.)

  12. Thorium utilization in a small long-life HTR. Part I: Th/U MOX fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands)

    2014-02-15

    Highlights: • We propose thorium MOX (TMOX) fuel blocks for a small block-type HTR. • The TMOX fuel blocks with low-enriched uranium are recommended. • More thorium decreases the reactivity swing of the TMOX fuel blocks. • Thorium reduces the negative temperature coefficient of the TMOX fuel blocks. • Thorium increases the conversion ratio of the TMOX fuel blocks. - Abstract: The U-Battery is a small, long-life and transportable high temperature gas-cooled reactor (HTR). The neutronic features of a typical fuel block with uranium and thorium have been investigated for a application of the U-Battery, by parametrically analyzing the composition and geometric parameters. The type of fuel block is defined as Th/U MOX fuel block because uranium and thorium are assumed to be mixed in each fuel kernel as a form of (Th,U)O{sub 2}. If the initially loaded mass of U-235 is mostly consumed in the early period of the lifetime of Th/U MOX fuel block, low-enriched uranium (LEU) as ignited fuel will not largely reduce the neutronic performance of the Th/U MOX fuel block, compared with high-enriched uranium. The radii of fuel kernels and fuel compacts and packing fraction of TRISO particles determine the atomic ratio of the carbon to heavy metal. When the ratio is smaller than 400, the difference among them due to double heterogeneous effects can be neglected for the Th/U MOX fuel block. In the range between 200 and 400, the reactivity swing of the Th/U MOX fuel block during 10 years is sufficiently small. The magnitude of the negative reactivity temperature coefficients of the Th/U MOX fuel block decreases by 20–45%, which is positive to reduce temperature defect of the Th/U MOX fuel block. The conversion ratio (CR) of the fuel increases from 0.48 (typical CR of the LEU-fueled U-Battery) to 0.78. The larger conversion ratio of the Th/U MOX fuel block reduces the reactivity swing during 10 years for the U-Battery.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    International Nuclear Information System (INIS)

    Tavron, Barak; Shwageraus, Eugene

    2016-01-01

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

  15. Axial temperatures and fuel management models for a HTR system

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1971-11-12

    In the HTR system, there is a large difference in temperature between different parts of the reactor core. The softer neutron spectrum in the upper colder core regions tends to shift the power productions in the fresh fuel upwards. As uranium 235 depletes and plutonium with its higher cross sections in the lower hot regions is built-up, an axial power flattening takes place. These effects have been studied in detail for a single column in an equilibrium environment. The aim of this paper is to relate these findings to a whole reactor core and to investigate the influence of axial temperatures on the overall performance and in particular, the fuel management scheme chosen for the reference design. A further objective has been to calculate the reactivity requirements for different part load conditions and for various daily and weekly load diagrams. As the xenon cross section changes significantly with temperature these investigations are performed for an equilibrium core with due representation of axial temperature zones.

  16. Actual characteristics study on HTR-10GT coupling with direct gas turbine cycle

    International Nuclear Information System (INIS)

    Peng Xuechuang; Zhu Shutang; Wang Jie

    2005-01-01

    Compared with a plant of steam turbine cycle, a HTGR plant with direct gas turbine cycle has a higher thermal efficiency. A lot of investigations on the characteristics of HTR-10GT, which is the reactor studying project of Tsinghua University, have been carried out, however, all of them are based on the theoretical Brayton Cycle which neglects many actual conditions, such as leakage, pressure loss and so on. For engineering practices, leakage is an unavoidable problem. The difference of the location and capacity of leakage will directly influence the working medium's thermoparameters and lead to fall of the cycle efficiency. The present study is focused on the performance of an actual Brayton cycle with practical conditions of leakage. The present study which based on building the physical and mathematical model of the leakage, aims to study the actual characteristics of the direct gas turbine circle. (authors)

  17. KüFA safety testing of HTR fuel pebbles irradiated in the High Flux Reactor in Petten

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O., E-mail: oliver.seeger@rwth-aachen.de [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Laurie, M., E-mail: mathias.laurie@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Abjani, A. El; Ejton, J.; Boudaud, D.; Freis, D.; Carbol, P.; Rondinella, V.V. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Fütterer, M. [European Commission, Joint Research Centre (JRC), Institute for Energy and Transport (IET), Nuclear Reactor Integrity Assessment and Knowledge Management Unit, PO Box 2, 1755 ZG Petten (Netherlands); Allelein, H.-J. [Lehrstuhl für Reaktorsicherheit und -technik an der RWTH Aachen, Kackertstraße 9, 52072 Aachen (Germany)

    2016-09-15

    The Cold Finger Apparatus (KühlFinger-Apparatur—KüFA) in operation at JRC-ITU is designed to experimentally scrutinize the effects of Depressurization LOss of Forced Circulation (D-LOFC) accident scenarios on irradiated High Temperature Reactor (HTR) fuel pebbles. Up to 1600 °C, the reference maximum temperature for these accidents, high-quality German HTR fuel pebbles have already demonstrated a small fission product release. This paper discusses and compares the releases obtained from KüFA-testing the pebbles HFR-K5/3 and HFR-EU1/3, which were both irradiated in the High Flux Reactor (HFR) in Petten. We present the time-dependent fractional release of the volatile fission product {sup 137}Cs as well as the fission gas {sup 85}Kr for both pebbles. For HFR-EU1/3 the isotopes {sup 134}Cs and {sup 154}Eu as well as the shorter-lived {sup 110m}Ag have also been measured. A detailed description of the experimental setup and its accuracy is given. The data for the recently tested pebbles is discussed in the context of previous results.

  18. The behaviour of spherical HTR fuel elements under accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schenk, W; Naoumidis, A [Institute for Reactor Material, KFA Juelich (Germany)

    1985-07-01

    Hypothetical accidents may lead to significantly higher temperatures in HTR fuel than during normal operation. In order to obtain meaningful statements on fission product behaviour and release, irradiated spherical fuel elements containing a large number of coated particles (20,000-40,000) with burnups between 6 and 16% FIMA were heated at temperatures between 1400 and 2500 deg. C. HTI-pyrocarbon coating retains the gaseous fission products (e.g. Kr) very well up to about 2400 deg. C if the burnup does not exceed the specified value for THTR (11.5%). Cs diffuses through the pyrocarbon significantly faster than Kr and the diffusion is enhanced at higher fuel burnups because of irradiation induced kernel microstructure changes. Below about 1800 deg. C the Cs release rate is controlled by diffusion in the fuel kernel; above this temperature the diffusion in the pyrocarbon coating is the controlling parameter. An additional SiC coating interlayer (TRISO) ensures Cs retention up to 1600 deg. C. However, the release obtained in the examined fuel elements was only by a factor of three lower than through the HTI pyrocarbon. Solid fission products added to UO{sub 2}-TRISO particles to simulate high burnup behave in various ways and migrate to attack the SiC coating. Pd migrates fastest and changes the SiC microstructure making it permeable.

  19. Mechanical Properties and Structures of Pyrolytic Carbon Coating Layer in HTR Coated Particle Fuel

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, Young Min; Kim, Woong Ki; Cho, Moon Sung

    2009-01-01

    The TRISO(tri-isotropic)-coated fuel particle for a HTR(High Temperature gas-cooled Reactor) has a diameter of about 1 mm, composed of a nuclear fuel kernel and four different outer coating layers, consisting of a buffer PyC (pyrolytic carbon) layer, inner PyC layer, SiC layer, and outer PyC layer with different coating thicknesses following a specific fuel design. While the fuel kernel is a source for a heat generation by a nuclear fission of fissile uranium, each of the four coating layers acts as a different role in view of retaining the generated fission products and the other interactions during an in-reactor service. Among these coating layers, PyC properties are scarcely in agreement among various investigators and the dependency of their changes upon the deposition condition is comparatively large due to their additional anisotropic properties. Although a recent review work has contributed to an establishment of relationship between the material properties and QC measurements, the data on the mechanical properties and structural parameters of PyC coating layers remain still unclearly evaluated. A review work on dimensional changes of PyC by neutron irradiation was one of re-evaluative works recently attempted by the authors. In this work, an attempt was made to analyze and re-evaluate the existing data of the experimental results of the mechanical properties, i.e., Young's modulus and fracture stress, in relation with the coating conditions, density and the BAF (Bacon Anisotropy Factor), an important structural parameter, of PyC coating layers obtained from various experiments performed in the early periods of the HTR coated particle development

  20. State of the Art of helium heat exchanger development for future HTR-projects - HTR2008-58146

    International Nuclear Information System (INIS)

    Esch, M.; Juergens, B.; Hurtado, A.; Knoche, D.; Tietsch, W.

    2008-01-01

    In Germany two HTR nuclear power plants had been built and operated, the AVR-15 and the THTR-300. Also various projects for different purposes in a large power range had been developed, The AVR-15, an experimental reactor with a power output of 15 MWel was operated for more than 20 years with excellent results. The THTR-300 was designed as a prototype demonstration plant with 300 MWel and should be the technological basis for the entire future reactor line. The THTR-300 was prematurely shut down and decommissioned because of political reasons. But because of the accompanying comprehensive R and D program and the operation time of about 5 years, the technology was proved and essential operational results were gained. The AVR steam generator was installed above the reactor core. The six THTR heat exchangers were arranged circularly around the reactor core, Both heat exchanger systems have been operated successfully and furthermore acted as a residual heat removal system. The technology knowledge and experience gained on these existing HTR plants is still available at Westinghouse Electric Germany GmbH since Westinghouse is one of the legal successors of the former German HTR companies. As a follow-up project of THTR, the HTR-500 was developed and designed up to the manufacturing stage. For this plant additionally to the 8 steam generators, two residual heat removal heat exchangers were foreseen. These were to be installed in a ring around the reactor core. All these HTRs were designed for the generation of electricity using a steam cycle. Extensive research work has also been done for advanced applications of HTR technology e.g. using a direct cycle within the HHT project or generating process heat within the framework of the PNP project, Because of the critical attitude of the German government to the nuclear power in the past 20 years in Germany there was only a very limited interest in the further development of the HTR technology. As a consequence of the German

  1. Establishment of quality control technology for HTR fuel in Korea

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Kim, Woong Ki; Kim, Yeon Ku; Cho, Moon Sung

    2009-01-01

    Korea is currently developing the HTR coated particle fuel technology in view of its long-term Nuclear Hydrogen Production Technology Development and Demonstration (NHDD) Project, which was launched in 2004, of an extensive R and D program on technology development for a hydrogen production by a VHTR. The current NHDD Project essentially covers the R and D works on the core and reactor system analysis, thermo-hydraulics and safety, coated particle fuel technology, material and component aspects and the hydrogen production technology by using the so-called Sulfur-Iodine Process (S-I Process). As a part of the NHDD Project, the fundamental technology for the coated particle fuel has been being developed, which consist of UO 2 kernel fabrication, pyrolytic carbon (PyC) and silicon carbide (SiC) coating technology, an in-reactor performance model development of a coated particle fuel and a preliminary preparative study for the irradiation tests of the coated particle fuel specimens in the HANARO reactor. In parallel with the development of fabrication process technology of the coated particle fuel, namely, kernel fabrication and coating processes, the characterization techniques for the important characteristics and quality control (QC) methods of the products after each process step were established. This paper deals with the works carried out for the development of the characterization technologies and establishment of the QC techniques for the coated fuel particles. Emphasis is given to the selection and development of the laboratory equipment and apparatus for the development of the methods of the characterizations and relevant QC methods

  2. HTR-PM Safety requirement and Licensing experience

    International Nuclear Information System (INIS)

    Li Fu; Zhang Zuoyi; Dong Yujie; Wu Zongxin; Sun Yuliang

    2014-01-01

    HTR-PM is a 200MWe modular pebble bed high temperature reactor demonstration plant which is being built in Shidao Bay, Weihai, Shandong, China. The main design parameters of HTR-PM were fixed in 2006, the basic design was completed in 2008. The review of Preliminary Safety Analysis Report (PSAR) of HTR-PM was started in April 2008, completed in September 2009. In general, HTR- PM design complies with the current safety requirement for nuclear power plant in China, no special standards are developed for modular HTR. Anyway, Chinese Nuclear Safety Authority, together with the designers, developed some dedicated design criteria for key systems and components and published the guideline for the review of safety analysis report of HTR-PM, based on the experiences from licensing of HTR-10 and new development of nuclear safety. The probabilistic safety goal for HTR-PM was also defined by the safety authority. The review of HTR-PM PSAR lasted for one and a half years, with 3 dialogues meetings and 8 topics meetings, with more than 2000 worksheets and answer sheets. The heavily discussed topics during the PSAR review process included: the requirement for the sub-atmospheric ventilation system, the utilization of PSA in design process, the scope of beyond design basis accidents, the requirement for the qualification of TRISO coating particle fuel, and etc. Because of the characteristics of first of a kind for the demonstration plant, the safety authority emphasized the requirement for the experiment and validation, the PSAR was licensed with certain licensing conditions. The whole licensing process was under control, and was re-evaluated again after Fukushima accident to be shown that the design of HTR-PM complies with current safety requirement. This is a good example for how to license a new reactor. (author)

  3. Potential Applications for Nuclear Energy besides Electricity Generation: AREVA Global Perspective of HTR Potential Market

    International Nuclear Information System (INIS)

    Soutworth, Finis; Gauthier, Jean-Claude; Lecomte, Michel; Carre, Franck

    2007-01-01

    Energy supply is increasingly showing up as a major issue for electricity supply, transportation, settlement, and process heat industrial supply including hydrogen production. Nuclear power is part of the solution. For electricity supply, as exemplified in Finland and France, the EPR brings an immediate answer; HTR could bring another solution in some specific cases. For other supply, mostly heat, the HTR brings a solution inaccessible to conventional nuclear power plants for very high or even high temperature. As fossil fuels costs increase and efforts to avoid generation of Greenhouse gases are implemented, a market for nuclear generated process heat will develop. Following active developments in the 80's, HTR have been put on the back burner up to 5 years ago. Light water reactors are widely dominating the nuclear production field today. However, interest in the HTR technology was renewed in the past few years. Several commercial projects are actively promoted, most of them aiming at electricity production. ANTARES is today AREVA's response to the cogeneration market. It distinguishes itself from other concepts with its indirect cycle design powering a combined cycle power plant. Several reasons support this design choice, one of the most important of which is the design flexibility to adapt readily to combined heat and power applications. From the start, AREVA made the choice of such flexibility with the belief that the HTR market is not so much in competition with LWR in the sole electricity market but in the specific added value market of cogeneration and process heat. In view of the volatility of the costs of fossil fuels, AREVA's choice brings to the large industrial heat applications the fuel cost predictability of nuclear fuel with the efficiency of a high temperature heat source free of greenhouse gases emissions. The ANTARES module produces 600 MWth which can be split into the required process heat, the remaining power drives an adapted prorated

  4. Rework of process effluents from the fabrication of HTR fuel

    International Nuclear Information System (INIS)

    Lasberg, Ingo; Braehler, Georg; Boyes, David

    2008-01-01

    HTR fuel facilities require the application of several liquid chemicals and accordingly they produce significant amounts of Uranium contaminated/potentially contaminated effluents. The main effluents are (amounts for a 3 t Uranium/a plant): aqueous solutions including tetrahydrofurfuryl alcohol THFA, ammonium hydroxide NH4OH, and ammonium nitrate NH4NO3 (180 m 3 /a), isopropanol IPA/water mixtures (130 m 3 /a); Non-Process Water NPW (300 m 3 /a); methanol (7m 3 /a); additionally off-gas streams, containing ammonia (9 t/a) have to be treated. In an industrial scale facility all such effluents/gases need to be processed for recycling, decontamination prior to release to the environment (as waste or as valuable material). Thermal decomposition is applied to dispose of burnable residues.

  5. Strengths, weaknesses, opportunities and threats for HTR deployment in Europe

    International Nuclear Information System (INIS)

    Bredimas, Alexandre; Kugeler, Kurt; Fütterer, Michael A.

    2014-01-01

    High temperature nuclear reactors are a technology, of which early versions were demonstrated in the 1960s–1980s in Germany (AVR, THTR) and the United States (Peach Bottom, Fort St. Vrain). HTRs were initially designed for high temperature, high efficiency electricity generation but the technology, the market and the targeted applications have evolved since then to address industrial cogeneration and new operational conditions (in particular new safety regulations). This paper intends to analyse the latest status of HTR today, as regards their intrinsic strengths and weaknesses and their external context, whether positive (opportunities) or negative (threats). Different dimensions are covered by the analysis: technology status, results from R and D programmes (especially in Europe), competences and skills, licensing aspects, experience feedback from demonstrator operation (in particular in Germany), economic conditions and other non-technical aspects. Europe has a comprehensive experience in the field of HTR with capabilities in both pebble bed and prismatic designs (R and D, engineering, manufacturing, operation, dismantling, and the full fuel cycle). Europe is also a promising market for HTR as the process heat market is large with good industrial and cogeneration infrastructures. The analysis of the European situation is to a good deal indicative for the global potential of this technology

  6. HTR core physics analysis at NRG

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

    2002-01-01

    Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

  7. Assessment Of Possible Cycle Lengths For Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  8. Calculation of HTR-10 first criticality with MVP

    International Nuclear Information System (INIS)

    Xie Jiachun; Yao Lianying

    2015-01-01

    The first criticality of 10 MW pebble-bed high temperature gas-cooled reactor-test module (HTR-10) was calculated with MVP. According to the characteristics of HTR-10, the Statistical Geometry Model of MVP was employed to describe the random arrangement of coated fuel particles in the fuel pebbles and the random distribution of the fuel and dummy pebbles in the core. Compared with previous results from VSOP and MCNP, the MVP results with JENDL-3.3 library were little more different, but the results with ENDF/B-Ⅵ.8 library were very close. The relative errors were less than 0.7%, compared with the first criticality experimental results. The study shows that MVP could be used in the physics calculations for pebble bed high temperature gas-cooled reactors. (authors)

  9. Post-irradiation examination of a 13000C-HTR fuel experiment Project J 96.M3

    International Nuclear Information System (INIS)

    Bueger, J. de; Roettger, H.

    1977-01-01

    A large variety of loose coated fuel particles have been irradiated in the BR2 at Mol/Belgium at temperatures between 1200 0 C and 1400 0 C and up to a fast neutron fluence of 1.2x1022 cm -2 (E>0.1 MeV) as a Euratom sponsored experiment for the advanced testing of HTR fuel. The specimens have been provided by Belgonucleaire and the Dragon Project. A short description of the experiment as well as the results of post-irradiation examination mainly carried out at Petten (N.H.), The Netherlands, are presented here. The post-irradiation examination has shown that the required performance can be achieved by a number of the tested fuel specimens without serious damage

  10. Rework of process effluents from the fabrication of HTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lasberg, Ingo; Braehler, Georg [NUKEM Technologies GmbH (Germany); Boyes, David [Pebble Bed Modular Reactor (Pty) Ltd., Centurion (South Africa)

    2008-07-01

    HTR fuel facilities require the application of several liquid chemicals and accordingly they produce significant amounts of Uranium contaminated/potentially contaminated effluents. The main effluents are (amounts for a 3 t Uranium/a plant): aqueous solutions including tetrahydrofurfuryl alcohol THFA, ammonium hydroxide NH4OH, and ammonium nitrate NH4NO3 (180 m{sup 3}/a), isopropanol IPA/water mixtures (130 m{sup 3}/a); Non-Process Water NPW (300 m{sup 3}/a); methanol (7m{sup 3}/a); additionally off-gas streams, containing ammonia (9 t/a) have to be treated. In an industrial scale facility all such effluents/gases need to be processed for recycling, decontamination prior to release to the environment (as waste or as valuable material). Thermal decomposition is applied to dispose of burnable residues.

  11. The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident

    Energy Technology Data Exchange (ETDEWEB)

    Zuhair, S.; Setiadipura, Topan [National Nuclear Energy Agency of Indonesia, Serpong Tagerang Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Su' ud, Zaki [Bandung Institute of Technology (Indonesia). Dept. of Physics

    2017-03-15

    Graphite is used as the moderator, fuel barrier material, and core structure in High Temperature Reactors (HTRs). However, despite its good thermal and mechanical properties below the radiation and high temperatures, it cannot avoid corrosion as a consequence of an accident of water/air ingress. Degradation of graphite as a main HTR material and the formation of dangerous CO gas is a serious problem in HTR safety. One of the several steps that can be adopted to avoid or prevent the corrosion of graphite by the water/air ingress is the application of a thin layer of silicon carbide (SiC) on the surface of the fuel element. This study investigates the effect of applying SiC coating on the fuel surfaces of pebble-bed HTR in water ingress accident from the reactivity points of view. A series of reactivity calculations were done with the Monte Carlo transport code MCNPX and continuous energy nuclear data library ENDF/B-VII at temperature of 1200 K. Three options of UO{sub 2}, PuO{sub 2}, and ThO{sub 2}/UO{sub 2} fuel kernel were considered to obtain the inter comparison of the core reactivity of pebble-bed HTR in conditions of water/air ingress accident. The calculation results indicated that the UO{sub 2}-fueled pebble-bed HTR reactivity was slightly reduced and relatively more decreased when the thickness of the SiC coating increased. The reactivity characteristic of ThO{sub 2}/UO{sub 2}-fueled pebble-bed HTR showed a similar trend to that of UO{sub 2}, but did not show reactivity peak caused by water ingress. In contrast with UO{sub 2}- and ThO{sub 2}-fueled pebble-bed HTR, although the reactivity of PuO{sub 2}-fueled pebble-bed HTR was the lowest, its characteristics showed a very high reactivity peak (0.33 Δk/k) and this introduction of positive reactivity is difficult to control. SiC coating on the surface of the plutonium fuel pebble has no significant impact. From the comparison between reactivity characteristics of uranium, thorium and plutonium cores with 0

  12. Waste heat of HTR power stations for district heating

    International Nuclear Information System (INIS)

    Bonnenberg, H.; Schlenker, H.V.

    1975-01-01

    The market situation, the applied techniques, and the transport, for district heating in combination with HTR plants are considered. Analysis of the heat market indicates a high demand for heat at temperatures between 100 and 150 0 C in household and industry. This market for district heating can be supplied by heat generated in HTR plants using two methods: (1) the combined heat and power generation in steam cycle plants by extracting steam from the turbine, and (2) the use of waste heat of a closed gas turbine cycle. The heat generation costs of (2) are negligible. The cost for transportation of heat over the average distance between existing plant sites and consumer regions (25 km) are between 10 and 20% of the total heat price, considering the high heat output of nuclear power stations. Comparing the price of heat gained by use of waste heat in HTR plants with that of conventional methods, considerable advantages are indicated for the combined heat and power generation in HTR plants. (author)

  13. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  14. Digital Distributed Control System Design: Control Policy for Shared Objects in HTR-PM

    International Nuclear Information System (INIS)

    Zhou Shuqiao; Huang Xiaojin

    2014-01-01

    HTR-PM is an HTR demonstration plant with a structure of two modules feeding one steam turbine. Compared with the structure of one single reactor feeding one turbine, there are more devices shared between these two modules. When they are operated, the shared components are prone to introduce collisions or even logical deadlocks for different technical processes. The future commercial HTR-PM plants are supposed to comprise more modules for a larger turbine, thus the collision problem introduced by the shared components may become severer. Therefore, how to design suitable policies in the distributed control system (DCS) to relieve the collisions during using these shared devices is a new and also a very important problem. In this paper, the classifications of the shared devices are first addressed, and then how to identify the shared objects of an NPP is proposed. Furthermore, a general model for the control logic design is proposed, taking into consideration the collision avoidance, time delay and fairness. The example of how to apply the schemes to relieve the conflicts and deadlocks in the processes of using the shared devices in fuel element cycling system is illustrated. (author)

  15. HTR core physics and transient analyses by the Panthermix code system

    International Nuclear Information System (INIS)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

    2005-01-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

  16. Modelling of HTR (High Temperature Reactor Pebble-Bed 10 MW to Determine Criticality as A Variations of Enrichment and Radius of the Fuel (Kernel With the Monte Carlo Code MCNP4C

    Directory of Open Access Journals (Sweden)

    Hammam Oktajianto

    2014-12-01

    Full Text Available Gas-cooled nuclear reactor is a Generation IV reactor which has been receiving significant attention due to many desired characteristics such as inherent safety, modularity, relatively low cost, short construction period, and easy financing. High temperature reactor (HTR pebble-bed as one of type of gas-cooled reactor concept is getting attention. In HTR pebble-bed design, radius and enrichment of the fuel kernel are the key parameter that can be chosen freely to determine the desired value of criticality. This paper models HTR pebble-bed 10 MW and determines an effective of enrichment and radius of the fuel (Kernel to get criticality value of reactor. The TRISO particle coated fuel particle which was modelled explicitly and distributed in the fuelled region of the fuel pebbles using a Simple-Cubic (SC lattice. The pebble-bed balls and moderator balls distributed in the core zone using a Body-Centred Cubic lattice with assumption of a fresh fuel by the fuel enrichment was 7-17% at 1% range and the size of the fuel radius was 175-300 µm at 25 µm ranges. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP4C. The details of model are discussed with necessary simplifications. Criticality calculations were conducted by Monte Carlo transport code MCNP4C and continuous energy nuclear data library ENDF/B-VI. From calculation results can be concluded that an effective of enrichment and radius of fuel (Kernel to achieve a critical condition was the enrichment of 15-17% at a radius of 200 µm, the enrichment of 13-17% at a radius of 225 µm, the enrichments of 12-15% at radius of 250 µm, the enrichments of 11-14% at a radius of 275 µm and the enrichment of 10-13% at a radius of 300 µm, so that the effective of enrichments and radii of fuel (Kernel can be considered in the HTR 10 MW. Keywords—MCNP4C, HTR, enrichment, radius, criticality 

  17. The Energy Conversion Analysis of HTR Gas Turbine System

    International Nuclear Information System (INIS)

    Utaja

    2000-01-01

    The energy conversion analysis of HTR gas turbine system by hand calculation is tedious work and need much time. This difficulty comes from the repeated thermodynamic process calculation, both on compression or expansion of the cycle. To make the analysis faster and wider variable analyzed, HTR-1 programme is used. In this paper, the energy conversion analysis of HTR gas turbine system by HTR-1 will be described. The result is displayed as efficiency curve and block diagram with the input and output temperature of the component. This HTR-1 programme is developed by Basic language programming and be compiled by Visual Basic 5.0 . By this HTR-1 programme, the efficiency, specific power and effective compression of the amount of gas can be recognized fast. For example, for CO 2 gas between 40 o C and 700 o C, the compression on maximum efficiency is 4.6 and the energy specific is 18.9 kcal/kg, while the temperature changing on input and output of the component can be traced on monitor. This process take less than one second, while the manual calculation take more than one hour. It can be concluded, that the energy conversion analysis of the HTR gas turbine system by HTR-1 can be done faster and more variable analyzed. (author)

  18. Equilibrium core layout for the 1000 MW direct cycle HTR (HHT) with hexagonal monolith moulded fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Dworak, A

    1973-03-15

    The aim of this survey is to calculate an equilibrium Thorium fuel cycle for a 1000 MW HHT-core in off-load refuelling with hexagonal monolith moulded fuel blocks. It was tried to achieve an axial power distribution similar to the advanced pebble-bed reactors (OTTO) by introducing three axial core zones with different heavy metal content and initial enrichment.

  19. HTR fuel modelling with the ATLAS code. Thermal mechanical behaviour and fission product release assessment

    International Nuclear Information System (INIS)

    Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent

    2009-01-01

    To support AREVA NP in its design on HTR reactor and its HTR fuel R and D program, the Commissariat a l'Energie Atomique developed the ATLAS code (Advanced Thermal mechanicaL Analysis Software) with the objectives: - to quantify, with a statistical approach, the failed particle fraction and fission product release of a HTR fuel core under normal and accidental conditions (compact or pebble design). - to simulate irradiation tests or benchmark in order to compare measurements or others code results with ATLAS evaluation. These two objectives aim at qualifying the code in order to predict fuel behaviour and to design fuel according to core performance and safety requirements. A statistical calculation uses numerous deterministic calculations. The finite element method is used for these deterministic calculations, in order to be able to choose among three types of meshes, depending on what must be simulated: - One-dimensional calculation of one single particle, for intact particles or particles with fully debonded layers. - Two-dimensional calculations of one single particle, in the case of particles which are cracked, partially debonded or shaped in various ways. - Three-dimensional calculations of a whole compact slice, in order to simulate the interactions between the particles, the thermal gradient and the transport of fission products up to the coolant. - Some calculations of a whole pebble, using homogenization methods are being studied. The temperatures, displacements, stresses, strains and fission product concentrations are calculated on each mesh of the model. Statistical calculations are done using these results, taking into account ceramic failure mode, but also fabrication tolerances and material property uncertainties, variations of the loads (fluence, temperature, burn-up) and core data parameters. The statistical method used in ATLAS is the importance sampling. The model of migration of long-lived fission products in the coated particle and more

  20. Future Development of Modular HTGR in China after HTR-PM

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Wang, Haitao; Dong Yujie; Li Fu

    2014-01-01

    The modular high temperature gas-cooled reactor (MHTGR) is an inherently safe nuclear energy technology for efficient electricity generation and process heat applications. The MHTGR is promising in China as it may replace fossil fuels in broader energy markets. In line with China’s long-term development plan of nuclear power, the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University developed and designed a MHTGR demonstration plant, named high-temperature gas-cooled reactor-pebble bed module (HTR-PM). The HTR-PM came into the construction phase at the end of 2012. The HTR-PM aims to demonstrate safety, economic potential and modularization technologies towards future commercial applications. Based on experiences obtained from the HTR-PM project with respect to design, manufacture, construction, licensing and project management, a further step aiming to promote commercialization and market applications of the MHTGR is expected. To this purpose, INET is developing a commercialized MHTGR named HTR-PM600 and a conceptual design is under way accordingly. HTR-PM600 is a pebble-bed MHTGR power generation unit with a six-pack of 250MWth reactor modules. The objective is to cogenerate electricity and process heat flexibly and economically in order to meet a variety of market needs. The design of HTR-PM600 closely follows HTR-PM with respect to safety features, system configuration and plant layout. HTR-PM600 has the six modules feeding one steam turbine to generate electricity with capacity to extract high temperature steam from various interfaces of the turbine for further process heat applications. A standard plant consists of two HTR-PM600 units. Based on the economic information of HTR-PM, a preliminary study is carried out on the economic prospect of HTR-PM600. (author)

  1. Fuel handling, reprocessing, and waste and related nuclear data aspects

    International Nuclear Information System (INIS)

    Kuesters, H.; Lalovic, M.; Wiese, H.W.

    1979-06-01

    The essential processes in the out-of-pile nuclear fuel cycle are described, i.e. mining and milling of uranium ores, enrichment, fuel fabrication, storage, transportation, reprocessing of irradiated fuel, waste treatment and waste disposal. The aspects of radiation (mainly gammas and neutrons) and of heat production, as well as special safety considerations are outlined with respect to their potential operational impacts and long-term hazards. In this context the importance of nuclear data for the out-of-pile fuel cycle is discussed. Special weight is given to the LWR fuel cycle including recycling; the differences of LMFBR high burn-up fuel with large PuO 2 content are described. The HTR fuel cycle is discussed briefly as well as some alternative fuel cycle concepts. (orig.) [de

  2. Nuclear power fuel cycle

    International Nuclear Information System (INIS)

    Havelka, S.; Jakesova, L.

    1982-01-01

    Economic problems are discussed of the fuel cycle (cost of the individual parts of the fuel cycle and the share of the fuel cycle in the price of 1 kWh), the technological problems of the fuel cycle (uranium ore mining and processing, uranium isotope enrichment, the manufacture of fuel elements, the building of long-term storage sites for spent fuel, spent fuel reprocessing, liquid and gaseous waste processing), and the ecologic aspects of the fuel cycle. (H.S.)

  3. Thorium fuel performance assessment in HTRs

    Energy Technology Data Exchange (ETDEWEB)

    Allelein, H.-J. [Forschungszentrum Jülich, D-52425 Jülich (Germany); RWTH Aachen, D-52072 Aachen (Germany); Kania, M.J.; Nabielek, H. [Forschungszentrum Jülich, D-52425 Jülich (Germany); Verfondern, K., E-mail: k.verfondern@fz-juelich.de [Forschungszentrum Jülich, D-52425 Jülich (Germany)

    2014-05-01

    Thorium as a nuclear fuel is receiving renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTR development employed thorium together with high-enriched uranium. After 1980, most HTR fuel systems switched to low-enriched uranium. After completing fuel development for AVR and THTR with BISO coated particles, the German program expanded efforts on a new program utilizing thorium and high-enriched uranium TRISO coated particles for advanced HTR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of LTI inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with HTI-BISO coatings. The improved performance of the HEU (Th,U)O{sub 2} TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 °C in normal operations and 1600 °C in accidents, with burnups up to 13% FIMA and fast fluences to 8 × 10{sup 25} m{sup −2} (E > 16 fJ), the results exceed the design limits on manufacturing and operational requirements for the German HTR Modul concept, which were: <6.5 × 10{sup −5} for manufacturing; <2 × 10{sup −4} for normal operating conditions; and <5 × 10{sup −4} for accident conditions. These

  4. HTR core physics and transient analyses by the Panthermix code system

    Energy Technology Data Exchange (ETDEWEB)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

  5. Effects of homogeneous geometry models in simulating the fuel balls in HTR-10

    International Nuclear Information System (INIS)

    Wang Mengjen; Liang Jenqhorng; Peir Jinnjer; Chao Dersheng

    2012-01-01

    In this study, the core geometry of HTR-10 was simulated using four different models including: (1) model 1 - an explicit double heterogeneous geometry, (2) model 2 - a mixing of UO 2 kernel and four layers in each TRISO particle into one, (3) model 3 - a mixing of 8,335 TRISO particles and the inner graphite matrix in each fuel ball into one, and (4) model 4 - a mixing of the outer graphite shell, 8,335 TRISO particles, and the inner graphite matrix in each fuel ball into one. The associated initial core computations were performed using the MCNP version 1.51 computer code. The experimental fuel loading height of 123 cm was employed for each model. The results revealed that the multiplication factors ranged from largest to smallest with model 1, model 2, model 3, and model 4. The neutron spectrum in the fuel region of each models varied from the hardest to the softest are model 1, model 2, model 3, and model 4 while the averaged neutron spectrum in fuel ball from hardest to softest are model 4, model 3, model 2, and model 1. In addition, the CPU execution times extended from longest to shortest with model 1, model 2, model 3, and model 4. (author)

  6. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

    1977-01-01

    Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

  7. The Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    2011-08-01

    This brochure describes the nuclear fuel cycle, which is an industrial process involving various activities to produce electricity from uranium in nuclear power reactors. The cycle starts with the mining of uranium and ends with the disposal of nuclear waste. The raw material for today's nuclear fuel is uranium. It must be processed through a series of steps to produce an efficient fuel for generating electricity. Used fuel also needs to be taken care of for reuse and disposal. The nuclear fuel cycle includes the 'front end', i.e. preparation of the fuel, the 'service period' in which fuel is used during reactor operation to generate electricity, and the 'back end', i.e. the safe management of spent nuclear fuel including reprocessing and reuse and disposal. If spent fuel is not reprocessed, the fuel cycle is referred to as an 'open' or 'once-through' fuel cycle; if spent fuel is reprocessed, and partly reused, it is referred to as a 'closed' nuclear fuel cycle.

  8. Nuclear power plant types and the management of plutonium and minor actinides - in search of fuel cycle flexibility

    International Nuclear Information System (INIS)

    Thomas, J.B.

    2002-01-01

    Transuranics management concerns all NPP types, because of the specifications for sustainable development. Multiple recycling is mandatory. Neutronic abundance can be obtained in fast spectrum, or by adding external neutrons or (temporarily) with additional 235 U. The LWRs can control the plutonium inventory and significantly reduce the amount of transuranics transferred to the geological repository, thanks to the use of innovative nuclear fuel in a limited part of the NPP fleet. HTR adapted to transuranics burning can help. In the future, in addition to the liquid metal FBR, a strategy based on a gas cooled technological line and advanced fuel opens a second path towards fast spectra. Strategies for defining the optimal mix of reactor types in the nuclear fleet at a given time and demonstrating the fuel cycle flexibility are under study. (author)

  9. Current status and technical description of Chinese 2 x 250 MWth HTR-PM demonstration plant

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Wu Zongxin; Wang Dazhong; Xu Yuanhui; Sun Yuliang; Li Fu; Dong Yujie

    2009-01-01

    After the nuclear accidents of Three Mile Island and Chernobyl the world nuclear community made great efforts to increase research on nuclear reactors and to develop advanced nuclear power plants with much improved safety features. Following the successful construction and a most gratifying operation of the 10 MW th high-temperature gas-cooled test reactor (HTR-10), the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University has developed and designed an HTR demonstration plant, called the HTR-PM (high-temperature-reactor pebble-bed module). The design, having jointly been carried out with industry partners from China and in collaboration of experts worldwide, closely follows the design principles of the HTR-10. Due to intensive engineering and R and D efforts since 2001, the basic design of the HTR-PM has been finished while all main technical features have been fixed. A Preliminary Safety Analysis Report (PSAR) has been compiled. The HTR-PM plant will consist of two nuclear steam supply system (NSSS), so called modules, each one comprising of a single zone 250 MW th pebble-bed modular reactor and a steam generator. The two NSSS modules feed one steam turbine and generate an electric power of 210 MW. A pilot fuel production line will be built to fabricate 300,000 pebble fuel elements per year. This line is closely based on the technology of the HTR-10 fuel production line. The main goals of the project are two-fold. Firstly, the economic competitiveness of commercial HTR-PM plants shall be demonstrated. Secondly, it shall be shown that HTR-PM plants do not need accident management procedures and will not require any need for offsite emergency measures. According to the current schedule of the project the completion date of the demonstration plant will be around 2013. The reactor site has been evaluated and approved; the procurement of long-lead components has already been started. After the successful operation of the demonstration plant

  10. Fast Reactor Fuel Cycle Cost Estimates for Advanced Fuel Cycle Studies

    International Nuclear Information System (INIS)

    Harrison, Thomas

    2013-01-01

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

  11. Status of the HTR programme in France

    International Nuclear Information System (INIS)

    Ballot, B.; Gauthier, J.C.; Hittner, D.; Lebrun, J.Ph.; Lecomte, M.; Carre, F.; Delbecq, J.M.

    2007-01-01

    AREVA is convinced that HTR (High Temperature Reactor) is not in competition with large LWRs for electricity generation, and that its actual added value is its potential for addressing cogeneration and industrial process heat production. Therefore AREVA launched in 2004 the ANTARES programme for a pre-conceptual design study, with the support of EDF and together with a large research and development programme needed for the design in close collaboration with Cea. The pre-conceptual phase was finalized end of 2006. The specific feature of AREVA's concept, which distinguishes it from other ones, is its indirect cycle design powering a combined cycle power plant. Several reasons support this design choice, one of the most important being the design flexibility to adapt readily to combined heat and power applications, with a standardised nuclear heat source as independent as possible of the versatile process heat applications with which it is coupled. Standardisation should expedite licensing. In view of the volatility of the costs of fossil fuels, AREVA's choice brings to the large industrial heat applications the fuel cost predictability of nuclear fuel with the efficiency of a high temperature heat source free of greenhouse gases emissions. The reactor module produces 600 MWth which can be split into the required process heat, the remaining power driving an adapted prorated electric plant. Depending on the process heat temperature and power needs, up to 80 % of the nuclear heat is converted into useful energy

  12. Fission Product Releases from a Core into a Coolant of a Prismatic 350-MWth HTR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Min; Jo, C. K. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    A prismatic 350-MW{sub th} high temperature reactor (HTR) is a means to generate electricity and process heat for hydrogen production. The HTR will be operated for an extended fuel burnup of more than 150 GWd/MTU. Korea Atomic Energy Research Institute (KAERI) is performing a point design for the HTR which is a pre-conceptual design for the analysis and assessment of engineering feasibility of the reactor. In a prismatic HTR, metallic and gaseous fission products (FPs) are produced in the fuel, moved through fuel materials, and released into a primary coolant. The FPs released into the coolant are deposited on the various helium-wetted surfaces in the primary circuit, or they are sorbed on particulate matters in the primary coolant. The deposited or sorbed FPs are released into the environment through the leakage or venting of the primary coolant. It is necessary to rigorously estimate such radioactivity releases into the environment for securing the health and safety of the occupational personnel and the public. This study treats the FP releases from a core into a coolant of a prismatic 350-MW{sub th} HTR. These results can be utilized as input data for the estimation of FP migration from a coolant into the environment. The analysis of fission product release within a prismatic 350-MW{sub th} HTR has been done. It was assumed that the HTR was operated at constant temperature and power for 1500 EFPDs. - The final burnup is 152 GWd/tHM at packing fraction of 25 %, and the final fast fluence is about 8 X 10{sup 21} n/cm{sup 2}, E{sub n} > 0.1 MeV. - The temperatures at the compact center and at the center of a kernel located at the compact center are 884 and 893 .deg. C, respectively, when the packing fraction is 25 % and the coolant temperature is 850 .deg. C. - Xenon is the most radioactive fission product in a coolant of a prismatic HTR when there are broken TRISOs and fuel component contaminated with heavy metals. For metallic fission products, the radioactivity

  13. Gas cooled HTR

    International Nuclear Information System (INIS)

    Schweiger, F.

    1985-01-01

    In the He-cooled, graphite-moderated HTR with spherical fuel elements, the steam generator is fixed outside the pressure vessel. The heat exchangers are above the reactor level. The hot gases stream from the reactor bottom over the heat exchanger, through an annular space around the heat exchanger and through feed lines in the side reflector of the reactor back to its top part. This way, in case of shutdown there is a supplementary natural draught that helps the inner natural circulation (chimney draught effect). (orig./PW)

  14. Analysis on First Criticality Benchmark Calculation of HTR-10 Core

    International Nuclear Information System (INIS)

    Zuhair; Ferhat-Aziz; As-Natio-Lasman

    2000-01-01

    HTR-10 is a graphite-moderated and helium-gas cooled pebble bed reactor with an average helium outlet temperature of 700 o C and thermal power of 10 MW. The first criticality benchmark problem of HTR-10 in this paper includes the loading number calculation of nuclear fuel in the form of UO 2 ball with U-235 enrichment of 17% for the first criticality under the helium atmosphere and core temperature of 20 o C, and the effective multiplication factor (k eff ) calculation of full core (5 m 3 ) under the helium atmosphere and various core temperatures. The group constants of fuel mixture, moderator and reflector materials were generated with WlMS/D4 using spherical model and 4 neutron energy group. The critical core height of 150.1 cm obtained from CITATION in 2-D R-Z reactor geometry exists in the calculation range of INET China, JAERI Japan and BATAN Indonesia, and OKBM Russia. The k eff calculation result of full core at various temperatures shows that the HTR-10 has negative temperature coefficient of reactivity. (author)

  15. The Dragon project and high temperature reactor (HTR position)

    International Nuclear Information System (INIS)

    Shepherd, L.

    1981-01-01

    After introduction describing the initiation of HTR work at AERE and in West Germany and the USA, the subject is discussed in detail under the headings: the Dragon Reactor Experiment (design and objectives); fuel elements and graphite (description of cooperative research programmes; development of coated fuel particles); helium technology; other Dragon activities. (U.K.)

  16. Denatured fuel cycles

    International Nuclear Information System (INIS)

    Till, C.E.

    1979-01-01

    This paper traces the history of the denatured fuel concept and discusses the characteristics of fuel cycles based on the concept. The proliferation resistance of denatured fuel cycles, the reactor types they involve, and the limitations they place on energy generation potential are discussed. The paper concludes with some remarks on the outlook for such cycles

  17. HTR-proteus pebble bed experimental program core 4: random packing with a 1:1 moderator-to-fuel pebble ratio

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Snoj, Luka [Jozef Stefan Inst. (IJS), Ljubljana (Slovenia); Lengar, Igor [Jozef Stefan Inst. (IJS), Ljubljana (Slovenia); Koberl, Oliver [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2014-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  18. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Leland M. Montierth

    2013-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  19. Euratom research and training in generation IV systems with emphasis on V/HTR

    International Nuclear Information System (INIS)

    Goethem, G. van; Manolatos, P.; Fuetterer, M.

    2006-01-01

    In this overview paper, the following questions are addressed: (1) What are the challenges facing the European Union nuclear fission research community in the short (today), medium (2010) and long term (2040)? (2) What kind of research and technological development (RTD) does Euratom offer to respond to these challenges, in particular in the area of reactor systems and fuel cycles? In the general debate about energy supply technologies there are challenges of both a scientific and technological (S/T) as well as an economic and political (E/P) nature. Though the Community research programme acts mainly on the former, there is nevertheless important links with Community policy. These not only exist in the specific area of nuclear policy. It is shown in the particular area of nuclear fission, to what extent Euratom research, education and innovation ('Knowledge Triangle: Education, Research, and Innovation') respond to the S/T challenges: (1) sustainability, (2) economics, (3) safety, and (4) proliferation resistance. At the European Commission (EC), the research related to nuclear reactor systems and fuel cycles is principally under the responsibility of the 2 Directorates Generals (DG) DG Research (RTD, located in Brussels), which implements and manages the programme of 'indirect actions', and the DG Joint Research Centre (JRC, headquarters in Brussels and 7 scientific institutes in 5 Member States) which carries out 'direct actions' in their own laboratories. In this HTR-2006 introductory paper, the emphasis is on the indirect and direct actions of the 6 th Euratom research framework programme 2003-2006, FP-6, with special emphasis on V/HTR Generation IV research. (orig.)

  20. Potential of thorium use in the HTR reactor

    International Nuclear Information System (INIS)

    Engelmann, P.; Hansen, U.; Kolb, G.; Leushacke, D.; Teuchert, E.; Werner, H.

    1979-08-01

    In this investigation, several types of reactors and fuel circulations are dealt with as they refer to the region of the Federal Republic of Germany and are compared with each other as to their need for uranium and their costs until 2100. This includes also an investigation covering the effects of a postponed application of uranium-saving reactors, a delayed reprocessing and two variants of the nuclear energy's contribution to electricity generation. After today's light water reactor (LWR) of the pressure water reactor type (DWR) and the sodium-cooled fast breeder (SBR) which is being developed, the technically rather developed helium-cooled high temperature reactor (HTR) is dealt with as another system. The high temperature reactor is, because of its high coolant temperatures, not only suitable as a nuclear power plant, but can also be used to substitute fossile energy sources on the heat market and is being developed in Germany also for use as process heat reactor for nuclear coal gasification. Here the application of nuclear energy is only considered with regard to the region of power generation. Besides the case of the LWR and HTR-operation without reprocessing and fuel recycling for all reactor systems, the calculations also take into consideration the case of the closed fuel recycling. While LWR and SBR are based on the uranium-plutonium-fuel recycling, the thorium-uranium fuel circulation is considered for the HTR with globular fuel elements. As investigations made until today are generally restricted to the system LWR/SBR and the uranium-plutonium circulation, a main concern of the investigations presented here is to show the potential of the Thorium-utilization in high-temperature reactors and to determine how this system can also be applied during the time period concerned to set up a nuclear energy strategy which is safe and profitable as far as the uranium supply is concerned. (orig./UA) 891 UA/orig.- 892 HIS [de

  1. Nuclear fuel cycle system analysis

    International Nuclear Information System (INIS)

    Ko, W. I.; Kwon, E. H.; Kim, S. G.; Park, B. H.; Song, K. C.; Song, D. Y.; Lee, H. H.; Chang, H. L.; Jeong, C. J.

    2012-04-01

    The nuclear fuel cycle system analysis method has been designed and established for an integrated nuclear fuel cycle system assessment by analyzing various methodologies. The economics, PR(Proliferation Resistance) and environmental impact evaluation of the fuel cycle system were performed using improved DB, and finally the best fuel cycle option which is applicable in Korea was derived. In addition, this research is helped to increase the national credibility and transparency for PR with developing and fulfilling PR enhancement program. The detailed contents of the work are as follows: 1)Establish and improve the DB for nuclear fuel cycle system analysis 2)Development of the analysis model for nuclear fuel cycle 3)Preliminary study for nuclear fuel cycle analysis 4)Development of overall evaluation model of nuclear fuel cycle system 5)Overall evaluation of nuclear fuel cycle system 6)Evaluate the PR for nuclear fuel cycle system and derive the enhancement method 7)Derive and fulfill of nuclear transparency enhancement method The optimum fuel cycle option which is economical and applicable to domestic situation was derived in this research. It would be a basis for establishment of the long-term strategy for nuclear fuel cycle. This work contributes for guaranteeing the technical, economical validity of the optimal fuel cycle option. Deriving and fulfillment of the method for enhancing nuclear transparency will also contribute to renewing the ROK-U.S Atomic Energy Agreement in 2014

  2. Fuel cycle management

    International Nuclear Information System (INIS)

    Herbin, H.C.

    1977-01-01

    The fuel cycle management is more and more dependent on the management of the generation means among the power plants tied to the grid. This is due mainly because of the importance taken by the nuclear power plants within the power system. The main task of the fuel cycle management is to define the refuelling pattern of the new and irradiated fuel assemblies to load in the core as a function of: 1) the differences which exist between the actual conditions of the core and what was expected for the present cycle, 2) the operating constraints and the reactor availability, 3) the technical requirements in safety and the technological limits of the fuel, 4) the economics. Three levels of fuel cycle management can be considered: 1) a long term management: determination of enrichments and expected cycle lengths, 2) a mid term management whose aim corresponds to the evaluation of the batch to load within the core as a function of both: the next cycle length to achieve and the integrated power history of all the cycles up to the present one, 3) a short term management which deals with the updating of the loaded fuel utilisations to take into account the operation perturbations, or with the alteration of the loading pattern of the next batch to respect unexpected conditions. (orig.) [de

  3. PBMR Project - Pebble Fuel Advantages

    International Nuclear Information System (INIS)

    Slabber, Johan; Matzie, Regis; Casperson, Sten; Kriel, Willem

    2006-01-01

    An overview is presented of all the important issues that influenced the choice of pebble fuel for the High-temperature Gas-cooled Reactor (HTGR) concept developed by South Africa. Each of these issues is then discussed in detail and compared with other fuel configurations proposed for direct cycle High-temperature Reactor (HTR) applications. The comparisons are provided using objective data generated by analyses done for the design of the Pebble Bed Modular Reactor (PBMR) and data that is available in open literature for the other fuel configurations

  4. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    Boczar, P.G.

    1999-01-01

    The fuel-cycle path chosen by a particular country will depend on a range of local and global factors. The CANDU reactor provides the fuel-cycle flexibility to enable any country to optimize its fuel-cycle strategy to suit its own needs. AECL has developed the CANFLEX fuel bundle as the near-term carrier of advanced fuel cycles. A demonstration irradiation of 24 CANFLEX bundles in the Point Lepreau power station, and a full-scale critical heat flux (CHF) test in water are planned in 1998, before commercial implementation of CANFLEX fuelling. CANFLEX fuel provides a reduction in peak linear element ratings, and a significant enhancement in thermalhydraulic performance. Whereas natural uranium fuel provides many advantages, the use of slightly enriched uranium (SEU) in CANDU reactors offers even lower fuel-cycle costs and other benefits, such as uprating capability through flattening the channel power distribution across the core. Recycled uranium (RU) from reprocessing spent PWR fuel is a subset of SEU that has significant economic promise. AECL views the use of SEU/RU in the CANFLEX bundle as the first logical step from natural uranium. High neutron economy enables the use of low-fissile fuel in CANDU reactors, which opens up a spectrum of unique fuel-cycle opportunities that exploit the synergism between CANDU reactors and LWRs. At one end of this spectrum is the use of materials from conventional reprocessing: CANDU reactors can utilize the RU directly without re-enrichment, the plutonium as conventional Mixed-oxide (MOX) fuel, and the actinide waste mixed with plutonium in an inert-matrix carrier. At the other end of the spectrum is the DUPIC cycle, employing only thermal-mechanical processes to convert spent LWR fuel into CANDU fuel, with no purposeful separation of isotopes from the fuel, and possessing a high degree of proliferation resistance. Between these two extremes are other advanced recycling options that offer particular advantages in exploiting the

  5. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    Boczar, P.G

    1998-05-01

    The fuel-cycle path chosen by a particular country will depend on a range of local and global factors. The CANDU reactor provides the fuel-cycle flexibility to enable any country to optimize its fuel-cycle strategy to suit its own needs. AECL has developed the CANFLEX fuel bundle as the near-term carrier of advanced fuel cycles. A demonstration irradiation of 24 CANFLEX bundles in the Point Lepreau power station, and a full-scale critical heat flux (CHF) test in water are planned in 1998, before commercial implementation of CANFLEX fuelling. CANFLEX fuel provides a reduction in peak linear element ratings, and a significant enhancement in thermalhydraulic performance. Whereas natural uranium fuel provides many advantages, the use of slightly enriched uranium (SEU) in CANDU reactors offers even lower fuel-cycle costs and other benefits, such as uprating capability through flattening the channel power distribution across the core. Recycled uranium (RU) from reprocessing spent PWR fuel is a subset of SEU that has significant economic promise. AECL views the use of SEU/RU in the CANFLEX bundle as the first logical step from natural uranium. High neutron economy enables the use of low-fissile fuel in CANDU reactors, which opens up a spectrum of unique fuel-cycle opportunities that exploit the synergism between CANDU reactors and LWRs. At one end of this spectrum is the use of materials from conventional reprocessing: CANDU reactors can utilize the RU directly without reenrichment, the plutonium as conventional mixed-oxide (MOX) fuel, and the actinide waste mixed with plutonium in an inert-matrix carrier. At the other end of the spectrum is the DUPIC cycle, employing only thermal-mechanical processes to convert spent LWR fuel into CANDU fuel, with no purposeful separation of isotopes from the fuel, and possessing a high degree of proliferation resistance. Between these two extremes are other advanced recycling options that offer particular advantages in exploiting the

  6. The physics design of the HTR-1160

    International Nuclear Information System (INIS)

    Huebner, A.; Brandes, S.

    1975-01-01

    This paper describes the physica design of the reactor core of the helium cooled, graphite moderated high-temperature reactor HTR-1160. A discussion is given of the design criteria, the calculational methods, the burnup cycle, the power distribution and the reactivity control. (orig.) [de

  7. ITER fuel cycle

    International Nuclear Information System (INIS)

    Leger, D.; Dinner, P.; Yoshida, H.

    1991-01-01

    Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

  8. HTR process heat applications, status of technology and economical potential

    International Nuclear Information System (INIS)

    Barnet, H.

    1997-01-01

    The technical and industrial feasibility of the production of high temperature heat from nuclear fuel is presented. The technical feasibility of high temperature heat consuming processes is reviewed and assessed. The conclusion is drawn that the next technological step for pilot plant scale demonstration is the nuclear heated steam reforming process. The economical potential of HTR process heat applications is reviewed: It is directly coupled to the economical competitiveness of HTR electricity production. Recently made statements and pre-conditions on the economic competitiveness in comparison to world market coal are reported. (author). 8 figs

  9. Concept of a HTR modular plant for generation of process heat in a chemical plant

    International Nuclear Information System (INIS)

    1991-07-01

    This final report summarizes the results of a preliminary study on behalf of Buna AG and Leunawerke AG. With regard to the individual situations the study investigated the conditions for modular HTR-2 reactors to cover on-site process heat and electric power demands. HTR-2 reactor erection and operation were analyzed for their economic efficiency compared with fossil-fuel power plants. Considering the prospective product lines, the technical and economic conditions were developed in close cooperation with Buna AG and Leunawerke AG. The study focused on the technical integration of modular HTR reactors into plants with regard to safety concepts, on planning, acceptance and erection concepts which largely exclude uncalculable scheduling and financial risks, and on comparative economic analyses with regard to fossil-fuel power plants. (orig.) [de

  10. Neutronic feasibility design of a small long-life HTR

    International Nuclear Information System (INIS)

    Ding Ming; Kloosterman, Jan Leen

    2011-01-01

    Highlights: ► We propose the neutronic feasibility design of a small, long lifetime and transportable HTR. ► Comparison of cylindrical, annular and scatter cores of the small block-type HTR. ► The design of the scatter core effectively reduces the number of the fuel block and increases the lifetime and burnup of the reactor. - Abstract: Small high temperature gas-cooled reactors (HTRs) have the advantages of transportability, modular construction and flexible site selection. This paper presents the neutronic feasibility design of a 20 MWth U-Battery, which is a long-life block-type HTR. Key design parameters and possible reactor core configurations of the U-Battery were investigated by SCALE 5.1. The design parameters analyzed include fuel enrichment, the packing fraction of TRISO particles, the radii of fuel compacts and kernels, and the thicknesses of top and bottom reflectors. Possible reactor core configurations investigated include five cylindrical, two annular and four scatter reactor cores for the U-Battery. The neutronic design shows that the 20 MWth U-Battery with a 10-year lifetime is feasible using less than 20% enriched uranium, while the negative values of the temperature coefficients of reactivity partly ensure the inherent safety of the U-Battery. The higher the fuel enrichment and the packing fraction of TRISO particles are, the lower the reactivity swing during 10 years will be. There is an optimum radius of fuel kernels for each value of the fuel compact design parameter (i.e., radius) and a specific fuel lifetime. Moreover, the radius of fuel kernels has a small influence on the infinite multiplication factor of a typical fuel block in the range of 0.2–0.25 mm, when the radius of fuel compacts is 0.6225 cm and the lifetime of the fuel block is 10 years. The comparison of the cylindrical reactor cores with the non-cylindrical ones shows that neutron under-moderation is a basic neutronic characteristic of the reactor core of the U

  11. Fuel cycle cost uncertainty from nuclear fuel cycle comparison

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  12. Thorium fuel cycle management

    International Nuclear Information System (INIS)

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

    2010-01-01

    In this presentation author deals with the thorium fuel cycle management. Description of the thorium fuels and thorium fuel cycle benefits and challenges as well as thorium fuel calculations performed by the computer code HELIOS are presented.

  13. Status of development of the HTR module

    International Nuclear Information System (INIS)

    Weisbrodt, I.A.

    1989-01-01

    Growing concern about the rising global temperature of the earth due to the ''Greenhouse Effect'' is increasingly focussing worldwide interest on passively safe reactors for heat and power production. In this context the development status of the HTR-Module designed by the Siemens-Group merits strong interest. The HTR-Module has a high degree of passive safety features. Even in case of hypothetical accidents the decay heat is dissipated from the primary system to the environment by passive measures alone i.e. by heat conduction, convection and radiation. The detailed engineering for the HTR-Module continues to progress. In addition to the engineering for the layout considerable progress has been made in the detailed engineering for specific components - e.g. pressure vessel, steam generator, hot gas duct, blower etc. - and specific systems - e.g. first core, helium purification system, reactor safety system, reactor control etc. The procedure for the conceptual licence has been continued. A large number of supplementary analyses and reports have been elaborated and submitted for this procedure. Many workshop meetings have been held with the nominated experts. The hypothetical accidents have been analysed and a special report on these accidents has been submitted. The safety analyses report has been revised, taking into account the results and achievements reached during the ongoing licensing procedure. Parallel to these engineering activities outstanding in R and D work for the HTR-Module, e.g. in the field of fuel elements etc. has been continued. The HTR-Module has found worldwide interest. Respective activities are going on in Bangladesh, PR China, USSR, Indonesia etc. Relevant application studies have been carried out and/or initiated. (author). 15 refs, 16 figs

  14. HTR-10GT AMBs displacement sensor design

    International Nuclear Information System (INIS)

    Shi Zhengang; Zha Meisheng; Zhao Lei; Sun Zhuo

    2005-01-01

    The 10 MW high temperature gas-cooled test module reactor (HTR-10GT) with the core made of spherical fuel elements was designed and constructed by the Institute of Nuclear and New Energy Technology of Tsinghua University in China. In the HTR-10GT, turbo-compressor and generator rotors are connected by a flexible coupling. The rotors, restricted by actual instruments and working environment, must be supported without any contact and lubrication. Active magnetic bearing (AMB), known as its advantages over the conventional bearings., such as contact-free, no-lubricating and active damping vibration, is the best way to suspend and stabilize the position of rotors of HTR-10GT. Each rotor is suspended by two radial and one axial AMBs. The radial AMB's radial gap is 0.15 mm considering the gap of 0.4 mm between the compressor stator and blades in order to protect the compressor. The control system controls the rotor position to meet the required gaps between rotor and stator through windings current. All the position information concerning radial and axial AMB is generated by sensors for measuring the displacement of the levitated body. Some typical sensors, i.e. eddy current displacement sensor, capacitive displacement sensor, can provide position information, but, quite often, unsatisfactory anti-jamming, which is a key issue for AMB systems near generator and other electric devices in HTR-10GT. Therefore, a kind of new type sensor is designed to measure the radial and axial displacements and the vibration of the rotors. This paper focuses on the design characteristics of the HTR-10GT AMBs displacement sensors and introduction of the related experiments to demonstrate its performance. (authors)

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

    International Nuclear Information System (INIS)

    2009-04-01

    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

  16. Nuclear-fuel-cycle education: Module 1. Nuclear fuel cycle overview

    International Nuclear Information System (INIS)

    Eckhoff, N.D.

    1981-07-01

    This educational module is an overview of the nuclear-fule-cycle. The overview covers nuclear energy resources, the present and future US nuclear industry, the industry view of nuclear power, the International Nuclear Fuel Cycle Evaluation program, the Union of Concerned Scientists view of the nuclear-fuel-cycle, an analysis of this viewpoint, resource requirements for a model light water reactor, and world nuclear power considerations

  17. Large-scale fuel cycle centres

    International Nuclear Information System (INIS)

    Smiley, S.H.; Black, K.M.

    1977-01-01

    The US Nuclear Regulatory Commission (NRC) has considered the nuclear energy centre concept for fuel cycle plants in the Nuclear Energy Centre Site Survey 1975 (NECSS-75) Rep. No. NUREG-0001, an important study mandated by the US Congress in the Energy Reorganization Act of 1974 which created the NRC. For this study, the NRC defined fuel cycle centres as consisting of fuel reprocessing and mixed-oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle centre sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000-300,000MW(e). The types of fuel cycle facilities located at the fuel cycle centre permit the assessment of the role of fuel cycle centres in enhancing the safeguard of strategic special nuclear materials - plutonium and mixed oxides. Siting fuel cycle centres presents a smaller problem than siting reactors. A single reprocessing plant of the scale projected for use in the USA (1500-2000t/a) can reprocess fuel from reactors producing 50,000-65,000MW(e). Only two or three fuel cycle centres of the upper limit size considered in the NECSS-75 would be required in the USA by the year 2000. The NECSS-75 fuel cycle centre evaluation showed that large-scale fuel cycle centres present no real technical siting difficulties from a radiological effluent and safety standpoint. Some construction economies may be achievable with fuel cycle centres, which offer opportunities to improve waste-management systems. Combined centres consisting of reactors and fuel reprocessing and mixed-oxide fuel fabrication plants were also studied in the NECSS. Such centres can eliminate shipment not only of Pu but also mixed-oxide fuel. Increased fuel cycle costs result from implementation of combined centres unless the fuel reprocessing plants are commercial-sized. Development of Pu-burning reactors could reduce any economic penalties of combined centres. The need for effective fissile

  18. Future fuel cycles

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1980-01-01

    A fuel cycle must offer both financial and resource savings if it is to be considered for introduction into Ontario's nuclear system. The most promising alternative CANDU fuel cycles are examined in the context of both of these factors over a wide range of installed capacity growth rates and economic assumptions, in order to determine which fuel cycle, or cycles, should be introduced, and when. It is concluded that the optimum path for the long term begins with the prompt introduction of the low-enriched-uranium fuel cycle. For a wide range of conditions, this cycle remains the optimum throughout the very long term. Conditions of rapid nuclear growth and very high uranium price escalation rates warrant the supersedure of the low-enriched-uranium cycle by either a plutonium-topped thorium cycle or plutonium recycle, beginning between 2010 and 2025. It is also found that the uranium resource position is sound in terms of both known resources and production capability. Moreover, introduction of the low-enriched-uranium fuel cycle and 1250 MWe reactor units will assure the economic viability of nuclear power until at least 2020, even if uranium prices increase at a rate of 3.5% above inflation. The interrelationship between these two conclusions lies in the tremendous incentive for exploration which will occur if the real uranium price escalation rate is high. From a competitive viewpoint, nuclear power can withstand increases in the price of uranium. However, such increases will likely further expand the resource base, making nuclear an even more reliable energy source. (auth)

  19. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    This chapter explains the distinction between fissile and fertile materials, examines briefly the processes involved in fuel manufacture and management, describes the alternative nuclear fuel cycles and considers their advantages and disadvantages. Fuel management is usually divided into three stages; the front end stage of production and fabrication, the back end stage which deals with the fuel after it is removed from the reactor (including reprocessing and waste treatment) and the stage in between when the fuel is actually in the reactor. These stages are illustrated and explained in detail. The plutonium fuel cycle and thorium-uranium-233 fuel cycle are explained. The differences between fuels for thermal reactors and fast reactors are explained. (U.K.)

  20. Improving Fuel Cycle Design and Safety Characteristics of a Gas Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Rooijen, W.F.G. van

    2006-01-01

    The Gas Cooled Fast Reactor (GCFR)is one of the Generation IV reactor concepts. This concept specifically targets sustainability of nuclear power generation. In nuclear reactors fertile material is converted to fissile fuel. If the neutrons inducing fission are highly energetic, the opportunity exists to convert more than one fertile nucleus per fission, thereby effectively breeding new nuclear fuel. Reactors operating on this principle are called ‘Fast Breeder Reactor’. Since natural uranium contains 99.3%of the fertile isotope 238 U, breeding increases the energy harvested from the nuclear fuel. If nuclear energy is to play an important role as a source of energy in the future, fast breeder reactors are essential for breeding nuclear fuel. Fast neutrons are also more efficient to destruct heavy (Minor Actinide, MA) isotopes, such as Np, Am and Cm isotopes, which dominate the long-term radioactivity of nuclear waste. So the waste life-time can be shortened if the MA nuclei are destroyed. An important prerequisite of sustainable nuclear energy is the closed fuel cycle, where only fission products are discharged to a final repository, and all Heavy Metal (HM) are recycled. The reactor should breed just enough fissile material to allow refueling of the same reactor, adding only fertile material to the recycled material. Other key design choices are highly efficient power conversion using a direct cycle gas turbine, and better safety through the use of helium, a chemically inert coolant which cannot have phase changes in the reactor core. Because the envisaged core temperatures and operating conditions are similar to thermal-spectrum High Temperature Reactor (HTR) concepts, the research for this thesis initially focused on a design based on existing HTR fuel technology: coated particle fuel, assembled into fuel assemblies. It was found that such a fuel concept could not meet the Generation IV criteria set for GCFR: self-breeding is difficult, the temperature

  1. Burning minor actinides in a HTR energy spectrum

    International Nuclear Information System (INIS)

    Pohl, Christoph; Rütten, H. Jochem

    2012-01-01

    Highlights: ► Burn-up analysis for varying plutonium/minor actinide fuel compositions. ► The influence of varying heavy metal fuel element loads is investigated. ► Significant burn-up via radiative capture and subsequently fission is observed. ► Difference observed between fuel element burn-up and total actinide burning rate. - Abstract: The generation of nuclear energy by means of the existing nuclear reactor systems is based mainly on the fission of U-235. But this comes along with the capture of neutrons by the U-238 faction and results in a build-up of plutonium isotopes and minor actinides as neptunium, americium and curium. These actinides are dominant for the long time assessment of the radiological risk of a final disposal therefore a minimization of the long living isotopes is aspired. Burning the actinides in a high temperature helium cooled graphite moderated reactor (HTR) is one of these options. The use of plutonium isotopes to sustain the criticality of the system is intended to avoid on the one hand highly enriched uranium because of international regulations and on the other hand low enriched uranium because of the build up of new actinides from neutron capture in the U-238 fraction. Because initial minor actinide isotopes are typically not fissionable by thermal neutrons the idea is to fission instead the intermediate isotopes generated by the first neutron capture. This paper comprises calculations for plutonium/minor actinides/thorium fuel compositions and their correlated final burn-up for a generic pebble bed HTR based on the reference design of the 400 MW PBMR. In particular the cross sections and the neutron balance of the different minor actinide isotopes in the higher thermal energy spectrum of a HTR will be discussed. For a fuel mixture of plutonium and minor actinides a significant burn-up of these actinides up to 20% can be achieved but at the expense of a higher residual fraction of plutonium in the burned fuel. Combining

  2. Large-scale fuel cycle centers

    International Nuclear Information System (INIS)

    Smiley, S.H.; Black, K.M.

    1977-01-01

    The United States Nuclear Regulatory Commission (NRC) has considered the nuclear energy center concept for fuel cycle plants in the Nuclear Energy Center Site Survey - 1975 (NECSS-75) -- an important study mandated by the U.S. Congress in the Energy Reorganization Act of 1974 which created the NRC. For the study, NRC defined fuel cycle centers to consist of fuel reprocessing and mixed oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle center sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000 - 300,000 MWe. The types of fuel cycle facilities located at the fuel cycle center permit the assessment of the role of fuel cycle centers in enhancing safeguarding of strategic special nuclear materials -- plutonium and mixed oxides. Siting of fuel cycle centers presents a considerably smaller problem than the siting of reactors. A single reprocessing plant of the scale projected for use in the United States (1500-2000 MT/yr) can reprocess the fuel from reactors producing 50,000-65,000 MWe. Only two or three fuel cycle centers of the upper limit size considered in the NECSS-75 would be required in the United States by the year 2000 . The NECSS-75 fuel cycle center evaluations showed that large scale fuel cycle centers present no real technical difficulties in siting from a radiological effluent and safety standpoint. Some construction economies may be attainable with fuel cycle centers; such centers offer opportunities for improved waste management systems. Combined centers consisting of reactors and fuel reprocessing and mixed oxide fuel fabrication plants were also studied in the NECSS. Such centers can eliminate not only shipment of plutonium, but also mixed oxide fuel. Increased fuel cycle costs result from implementation of combined centers unless the fuel reprocessing plants are commercial-sized. Development of plutonium-burning reactors could reduce any

  3. HTR-E project. High-temperature components and systems

    International Nuclear Information System (INIS)

    Breuil, E.; Exner, R.

    2002-01-01

    The HTR-E European project (four years project) is proposed for the 5th Framework Programme and concerns the technical developments needed for the innovative components of a modern HTR with a direct cycle. These components have been selected with reference to the present projects (GT-MHR, PBMR): (1) the helium turbine, the recuperator heat exchanger, the electro-magnetic bearings and the helium rotating seal; (2) the tribology. Sliding innovative components in helium environment are particularly concerned. (3) the helium purification system. Recommendations on impurities contents have to be provided in accordance with the materials proposed for the innovative components. The main outcomes expected from the HTR-E project are the design recommendations and identification of further R and D needs for these components. This will be based: (1) on experience feedback from European past helium test loops and reactors; (2) on design studies, thermal-hydraulic and structural analyses; (3) and on experimental tests

  4. HTR Plans in Poland

    International Nuclear Information System (INIS)

    Sobolewski, Józef

    2017-01-01

    Target for HTR: Polish Heat Market: Today 100% heat market is dominated by fossil fuels; mostly coal in district heating and coal and gas in industry heat generation. Huge potential for nuclear reactors Currently can be addressed only in terms of LWR, i.e. T <250 ° C, useful in district heating, but not in industry. Need for new technologies •HTGR (High Temperature Gas Reactor) ~600°C, e.g. for industry steam generation. •VHTR (Very High Temperature Reactor), ... ~1000°C, e.g. for hydrogen production

  5. Economics analysis of fuel cycle cost of fusion–fission hybrid reactors based on different fuel cycle strategies

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2015-01-15

    Highlights: • Economics analysis of fuel cycle cost of FFHRs is carried out. • The mass flows of different fuel cycle strategies are established based on the equilibrium fuel cycle model. • The levelized fuel cycle costs of different fuel cycle strategies are calculated, and compared with current once-through fuel cycle. - Abstract: The economics analysis of fuel cycle cost of fusion–fission hybrid reactors has been performed to compare four fuel cycle strategies: light water cooled blanket burning natural uranium (Strategy A) or spent nuclear fuel (Strategy B), sodium cooled blanket burning transuranics (Strategy C) or minor actinides (Strategy D). The levelized fuel cycle costs (LFCC) which does not include the capital cost, operation and maintenance cost have been calculated based on the equilibrium mass flows. The current once-through (OT) cycle strategy has also been analyzed to serve as the reference fuel cycle for comparisons. It is found that Strategy A and Strategy B have lower LFCCs than OT cycle; although the LFCC of Strategy C is higher than that of OT cycle when the uranium price is at its nominal value, it would become comparable to that of OT cycle when the uranium price reaches its historical peak value level; Strategy D shows the highest LFCC, because it needs to reprocess huge mass of spent nuclear fuel; LFCC is sensitive to the discharge burnup of the nuclear fuel.

  6. Nuclear graphite wear properties and estimation of graphite dust production in HTR-10

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Xiaowei, E-mail: xwluo@tsinghua.edu.cn; Wang, Xiaoxin; Shi, Li; Yu, Xiaoyu; Yu, Suyuan

    2017-04-15

    Highlights: • Graphite dust. • The wear properties of graphite. • Pebble bed. • High Temperature Gas-cooled Reactor. • Fuel element. - Abstract: The issue of the graphite dust has been a research focus for the safety of High Temperature Gas-cooled Reactors (HTGRs), especially for the pebble bed reactors. Most of the graphite dust is produced from the wear of fuel elements during cycling of fuel elements. However, due to the complexity of the motion of the fuel elements in the pebble bed, there is no systematic method developed to predict the amount the graphite dust in a pebble bed reactor. In this paper, the study of the flow of the fuel elements in the pebble bed was carried out. Both theoretical calculation and numerical analysis by Discrete Element Method (DEM) software PFC3D were conducted to obtain the normal forces and sliding distances of the fuel elements in pebble bed. The wearing theory was then integrated with PFC3D to estimate the amount of the graphite dust in a pebble bed reactor, 10 MW High Temperature gas-cooled test Reactor (HTR-10).

  7. VENUS: cold prototype installation of the head-end of the reprocessing of HTR fuel elements. Activity report, 1 July 1976--31 December 1976

    Energy Technology Data Exchange (ETDEWEB)

    Boehnert, R.; Walter, C.

    1977-02-15

    The purpose of the VENUS Project is advance planning for the construction of a cold prototype system to incinerate HTR fuel element graphite. The Venus Project is organized into four phases between advance planning and experimental operation, corresponding to the maturity of the work. It is in the advance planning phase. Status of individual studies is given. (LK)

  8. VENUS: cold prototype installation of the head-end of the reprocessing of HTR fuel elements. Activity report, 1 July 1976--31 December 1976

    International Nuclear Information System (INIS)

    Boehnert, R.; Walter, C.

    The purpose of the VENUS Project is advance planning for the construction of a cold prototype system to incinerate HTR fuel element graphite. The Venus Project is organized into four phases between advance planning and experimental operation, corresponding to the maturity of the work. It is in the advance planning phase. Status of individual studies is given

  9. Verifiable Fuel Cycle Simulation Model (VISION): A Tool for Analyzing Nuclear Fuel Cycle Futures

    International Nuclear Information System (INIS)

    Jacobson, Jacob J.; Piet, Steven J.; Matthern, Gretchen E.; Shropshire, David E.; Jeffers, Robert F.; Yacout, A.M.; Schweitzer, Tyler

    2010-01-01

    The nuclear fuel cycle consists of a set of complex components that are intended to work together. To support the nuclear renaissance, it is necessary to understand the impacts of changes and timing of events in any part of the fuel cycle system such as how the system would respond to each technological change, a series of which moves the fuel cycle from where it is to a postulated future state. The system analysis working group of the United States research program on advanced fuel cycles (formerly called the Advanced Fuel Cycle Initiative) is developing a dynamic simulation model, VISION, to capture the relationships, timing, and changes in and among the fuel cycle components to help develop an understanding of how the overall fuel cycle works. This paper is an overview of the philosophy and development strategy behind VISION. The paper includes some descriptions of the model components and some examples of how to use VISION. For example, VISION users can now change yearly the selection of separation or reactor technologies, the performance characteristics of those technologies, and/or the routing of material among separation and reactor types - with the model still operating on a PC in <5 min.

  10. Economic comparison of fusion fuel cycles

    International Nuclear Information System (INIS)

    Brereton, S.J.; Kazimi, M.S.

    1987-01-01

    The economics of the DT, DD, and DHe fusion fuel cycles are evaluated by comparison on a consistent basis. The designs for the comparison employ HT-9 structure and helium coolant; liquid lithium is used as the tritium breeding material for the DT fuel cycle. The reactors are pulsed, superconducting tokamaks, producing 1200 MW of electric power. The DT and DD designs scan a range of values of plasma beta, assuming first stability scaling laws. The results indicate that on a purely economic basis, the DT fuel cycle is superior to both of the advanced fuel cycles. Geometric factors, materials limitations, and plasma beta were seen to have an impact on the Cost of Electricity (COE). The economics for the DD fuel cycle are more strongly affected by these parameters than is the DT fuel cycle. Fuel costs are a major factor in determining the COE for the DHe fuel cycle. Based on costs directly attributable to the fuel cycle, the DT fuel cycle appears most attractive. Technological advances, improved understanding of physics, or strides in advanced energy conversion schemes may result in altering the economic ranking of the fuel cycles indicated here. 7 refs., 6 figs., 2 tabs

  11. The plutonium fuel cycles

    International Nuclear Information System (INIS)

    Pigford, T.H.; Ang, K.P.

    1975-01-01

    The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

  12. Fuel cycles

    International Nuclear Information System (INIS)

    Hawley, N.J.

    1983-05-01

    AECL publications, from the open literature, on fuels and fuel cycles used in CANDU reactors are listed in this bibliography. The accompanying index is by subject. The bibliography will be brought up to date periodically

  13. Nuclear-fuel-cycle costs. Consolidated Fuel-Reprocessing Program

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  14. Development status of the HTGR in the world. Outline and construction status of the demonstration HTGR program (HTR-PM) of China

    International Nuclear Information System (INIS)

    Ohashi, Kazutaka; Okamoto, Futoshi; Mouri, Tomoaki; Saito, Masanao; Nishio, Hiroki; Ohashi, Junpei

    2014-01-01

    Based on successful construction and operation experiences of HTR-10 reactor with pebble bed fuel and helium coolant, HTR-PM (HTR Pebble-bed Modular) reactor program was under way with 200 MWe of twin reactors with the same core configuration as HTR-10 reactor, which, each with a single steam generator, would drive a single steam turbine. Core height was 11 meters, and main steam temperature would be at 566 C. Although HTR-PM reactor program was interrupted by effects of the Fukushima accident, first concrete basement construction was started in December 2012 with aiming at connecting the Grid in 2017. This article reviewed outline and construction status of HTR-PM reactor in China. (T. Tanaka)

  15. A Statistical Analysis on the Coating Layer Thicknesses of a TRISO of 350 MWth Block-type HTR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Min; Jo, C. K.; Cho, M. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    A tri-isotropic coated fuel particle (TRISO) is a basic fuel element of a high temperature reactor (HTR). The block-type HTR fuel is a cylindrical graphite compact in which a large number of TRISOs are embedded. There are more than 11 billion TRISOs in a 350 MW{sub th} block-type HTR core. Among the RSM quadratic models, the BBD model produces the smallest errors at both interior and exterior points. The errors in the quadratic model of the small-type CCD is the biggest, particularly at exterior points. The CCD has a disadvantage of generating a number of decimal places in its factor levels because of its axial points. It is recommended to use the BBD or the full-type CCD with an adjusted axial point which does not produce the decimal places in its factor levels. More general statistical model for a TRISO design will be secured when the number of factors and responses increases. This study treats a statistical analysis on the optimal layer thicknesses of a UCO TRISO of 350 MW{sub th} block-type HTR which cause a minimum tangential stress to act on the SiC layer. Three response surface methods (RSMs) are used as statistical methods and their resulting quadratic models are compared.

  16. A Statistical Analysis on the Coating Layer Thicknesses of a TRISO of 350 MWth Block-type HTR

    International Nuclear Information System (INIS)

    Kim, Young Min; Jo, C. K.; Cho, M. S.

    2016-01-01

    A tri-isotropic coated fuel particle (TRISO) is a basic fuel element of a high temperature reactor (HTR). The block-type HTR fuel is a cylindrical graphite compact in which a large number of TRISOs are embedded. There are more than 11 billion TRISOs in a 350 MW_t_h block-type HTR core. Among the RSM quadratic models, the BBD model produces the smallest errors at both interior and exterior points. The errors in the quadratic model of the small-type CCD is the biggest, particularly at exterior points. The CCD has a disadvantage of generating a number of decimal places in its factor levels because of its axial points. It is recommended to use the BBD or the full-type CCD with an adjusted axial point which does not produce the decimal places in its factor levels. More general statistical model for a TRISO design will be secured when the number of factors and responses increases. This study treats a statistical analysis on the optimal layer thicknesses of a UCO TRISO of 350 MW_t_h block-type HTR which cause a minimum tangential stress to act on the SiC layer. Three response surface methods (RSMs) are used as statistical methods and their resulting quadratic models are compared

  17. Fuel and fuel cycles with high burnup for WWER reactors

    International Nuclear Information System (INIS)

    Chernushev, V.; Sokolov, F.

    2002-01-01

    The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

  18. Turbo-machine deployment of HTR-10 GT

    International Nuclear Information System (INIS)

    Zhu Shutang; Wang Jie; Zhang Zhengming; Yu Suyuan

    2005-01-01

    As a testing project of gas turbine modular High Temperature Gas-cooled Reactor (HTGR), HTR-10GT has been studied and developed by Institute of Nuclear and New Energy Technology (INET) of Tsinghua University after the success of HTR-10 with steam turbine cycle. The main purposes of this project are to demonstrate the gas turbine modular HTGR, to optimize the deployment of Power Conversion Unit (PCU) and to verify the techniques of turbo-machine, operating modes and controlling measures. HTR-10GT is concentrated on the PCU design and the turbo-machine deployment. Possible turbo-machine deployments have been investigated and two of them are introduced in this paper. The preliminary design for the turbo-machine of HTR-10GT is single-shaft of vertical layout, arranged by the side of the reactor and the turbo-compressor rotary speed was selected to be 250 s -1 (15000 r/min) by considering the efficiency of turbo-compressor blade systems, the strength conditions and the mass and size characteristics of the turbo-compressor. The rotor system will be supported by electromagnetic bearings (EMBs) to curb the possible pollutions of the primary loop. Of all the components in this design, the high speed turbo-generator seems to be a world-wide technical nut. As an alternative design, a gearbox complex is used to reduce the rotary speed from the turbo-compressor 250 s -1 to 50 s -1 so that the ordinary generator can be used. (authors)

  19. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

    Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

    2003-01-01

    The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

  20. Part 5. Fuel cycle options

    International Nuclear Information System (INIS)

    Lineberry, M.J.; McFarlane, H.F.; Amundson, P.I.; Goin, R.W.; Webster, D.S.

    1980-01-01

    The results of the FBR fuel cycle study that supported US contributions to the INFCE are presented. Fuel cycle technology is reviewed from both generic and historical standpoints. Technology requirements are developed within the framework of three deployment scenarios: the reference international, the secured area, and the integral cycle. Reprocessing, fabrication, waste handling, transportation, and safeguards are discussed for each deployment scenario. Fuel cycle modifications designed to increase proliferation defenses are described and assessed for effectiveness and technology feasibility. The present status of fuel cycle technology is reviewed and key issues that require resolution are identified

  1. Benefits of barrier fuel on fuel cycle economics

    International Nuclear Information System (INIS)

    Crowther, R.L.; Kunz, C.L.

    1988-01-01

    Barrier fuel rod cladding was developed to eliminate fuel rod failures from pellet/cladding stress/corrosion interaction and to eliminate the associated need to restrict the rate at which fuel rod power can be increased. The performance of barrier cladding has been demonstrated through extensive testing and through production application to many boiling water reactors (BWRs). Power reactor data have shown that barrier fuel rod cladding has a significant beneficial effect on plant capacity factor and plant operating costs and significantly increases fuel reliability. Independent of the fuel reliability benefit, it is less obvious that barrier fuel has a beneficial effect of fuel cycle costs, since barrier cladding is more costly to fabricate. Evaluations, measurements, and development activities, however, have shown that the fuel cycle cost benefits of barrier fuel are large. This paper is a summary of development activities that have shown that application of barrier fuel significantly reduces BWR fuel cycle costs

  2. Extended fuel cycle length

    International Nuclear Information System (INIS)

    Bruyere, M.; Vallee, A.; Collette, C.

    1986-09-01

    Extended fuel cycle length and burnup are currently offered by Framatome and Fragema in order to satisfy the needs of the utilities in terms of fuel cycle cost and of overall systems cost optimization. We intend to point out the consequences of an increased fuel cycle length and burnup on reactor safety, in order to determine whether the bounding safety analyses presented in the Safety Analysis Report are applicable and to evaluate the effect on plant licensing. This paper presents the results of this examination. The first part indicates the consequences of increased fuel cycle length and burnup on the nuclear data used in the bounding accident analyses. In the second part of this paper, the required safety reanalyses are presented and the impact on the safety margins of different fuel management strategies is examined. In addition, systems modifications which can be required are indicated

  3. Practical introduction of thorium fuel cycles

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1982-01-01

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

  4. Approach to equilibrium calculations for the dragon HTR design

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1971-06-10

    The calculational methods and the model used in representing the core and the fuel management operations are described. Different layouts of the first core and approach to equilibrium schemes for the Dragon HTR design are investigated. A simple fuelling modus is found and the tchnological and economical implications are discussed in detail.

  5. Advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

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

    1990-04-01

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

  6. Preliminary design study of pebble bed reactor HTR-PM base using once-through-then-out fuel recirculation

    International Nuclear Information System (INIS)

    Topan Setiadipura; Jupiter S Pane; Zuhair

    2016-01-01

    Pebble Bed Reactor (PBR) is one of the advanced reactor type implementing strong passive safety feature. In this type of design has the potential to do a cogeneration useful for the treatment of various minerals in various islands in Indonesia. The operation of the PBR can be simplified by implementing once-through-then-out (OTTO) fuel recirculation scheme in which pebble fuel only pass the core once time. The purpose of this research is to understand quantitative influence of the changing of fuel element recirculation on the PBR core performance and to find preliminary optimization design of PBR type reactor with OTTO recirculation scheme. PEBBED software was used to find PBR equilibrium core. The calculation result gives quantitative data on the impact of implementing a different fuel recirculation, especially using OTTO scheme. Furthermore, an early optimized PBR design based on HTR-PM using OTTO scheme was obtained where the power must be downgraded into 115 MWt in order to preserve the safety feature. The simplicity of the reactor operation and the reduction of reactor component with OTTO scheme still make this early optimized design an interesting alternative design, despite its power reduction from the reference design. (author)

  7. The fuel cycle scoping system

    International Nuclear Information System (INIS)

    Dooley, G.D.; Malone, J.P.

    1986-01-01

    The Fuel Cycle Scoping System (FCSS) was created to fill the need for a scoping tool which provides the utilities with the ability to quickly evaluate alternative fuel management strategies, tails assay choices, fuel fabrication quotes, fuel financing alternatives, fuel cycle schedules, and other fuel cycle perturbations. The FCSS was specifically designed for PC's that support dBASE-III(TM), a relational data base software system by Ashton-Tate. However, knowledge of dBASE-III is not necessary in order to utilize the FCSS. The FCSS is menu driven and can be utilized as a teaching tool as well as a scoping tool

  8. Economic evaluation of fast reactor fuel cycling

    International Nuclear Information System (INIS)

    Hu Ping; Zhao Fuyu; Yan Zhou; Li Chong

    2012-01-01

    Economic calculation and analysis of two kinds of nuclear fuel cycle are conducted by check off method, based on the nuclear fuel cycling process and model for fast reactor power plant, and comparison is carried out for the economy of fast reactor fuel cycle and PWR once-through fuel cycle. Calculated based on the current price level, the economy of PWR one-through fuel cycle is better than that of the fast reactor fuel cycle. However, in the long term considering the rising of the natural uranium's price and the development of the post treatment technology for nuclear fuels, the cost of the fast reactor fuel cycle is expected to match or lower than that of the PWR once-through fuel cycle. (authors)

  9. Optimization of the fuel cycle

    International Nuclear Information System (INIS)

    Kidd, S.W.; Balu, K.; Boczar, P.G.; Krebs, W.D.

    1999-01-01

    The nuclear fuel cycle can be optimized subject to a wide range of criteria. Prime amongst these are economics, sustainability of resources, environmental aspects, and proliferation-resistance of the fuel cycle. Other specific national objectives will also be important. These criteria, and their relative importance, will vary from country to country, and with time. There is no single fuel cycle strategy that is optimal for all countries. Within the short term, the industry is attached to dominant thermal reactor technologies, which themselves have two main variants, a cycle closed by reprocessing of spent fuel and subsequent recycling and a once through one where spent fuel is stored in advance of geological disposal. However, even with current technologies, much can be done to optimize the fuel cycles to meet the relevant criteria. In the long term, resource sustainability can be assured for centuries through the use of fast breeder reactors, supporting high-conversion thermal reactors, possibly also utilizing the thorium cycle. These must, however, meet the other key criteria by being both economic and safe. (author)

  10. Chemical processing of HTR fuels applying either THOREX or PUREX flow sheets

    Energy Technology Data Exchange (ETDEWEB)

    Zimmer, E; Merz, E [Kernforschungsanlage, Juelich GmbH, Institut fuer Chemische Technologie der Nuklearen Entsorgung, Juelich (Germany)

    1985-07-01

    Two fuel cycles are considered for utilization in high temperature gas-cooled reactors (HTRs): the high-enriched thorium-uranium (HEU 93% U-235) and the low-enriched uranium (LEU 8-12% U-235) fuel concept. For both fuel compositions suitable reprocessing procedures are required which are capable to separate the actinides thorium, uranium and plutonium from fission products and from each other. In any case, the processes under consideration utilize Tri-n-butylphosphate (TBP) together with a straight-chain paraffinic diluent (C{sub 8}-C{sub 14}, to day usually dodecane) as extractant in an aqueous nitrate system; most commonly, the related processes are known by the acronyms PUREX and THOREX. The PUREX process has become the reprocessing procedure quite generally used for all fuel types containing natural, slightly or highly enriched uranium together with lower or higher contents of plutonium. The THOREX process on the other hand has been developed to separate thorium, uranium and fission products from thorium based irradiated fuel. Generally, the utilization of the thorium fuel cycle is most attractive for High Temperature Reactors. On the other hand, the strong recommendation of INFCE to abandon the use of high-enriched uranium for nuclear energy applications virtually rules out the thorium fuel cycle, since economic utilization of thorium as a fertile material requires the use of high-enriched U-235. Thus, it was decided in the Federal Republic of Germany to switch over, at least for the foreseeable future, to the low enrichment uranium-plutonium fuel cycle, well aware of its economic shortcomings. In this paper various THOREX flowsheets as well as a PUREX variant suitable for LEU fuel reprocessing are described. Both processes have in common that the main stream is always presented by the fertile material, that means thorium and U-238, respectively.

  11. Characteristic analysis of rotor dynamics and experiments of active magnetic bearing for HTR-10GT

    International Nuclear Information System (INIS)

    Yang Guojun; Xu Yang; Shi Zhengang; Gu Huidong

    2005-01-01

    A 10 MW high-temperature gas-cooled reactor (HTR-10) was constructed by the Institute of Nuclear and New Energy Technology (INET) at Tsinghua University of China. The helium turbine and generator system of 10 MW high temperature gas-cooled reactor (HTR-10GT) is the second phase for the HTR-10 project. It is to set up a direct helium cycle to replace the current steam cycle. The active magnetic bearing (AMB) instead of ordinary mechanical bearing was chosen to support the rotor in the HTR-10GT. This rotor is vertically mounted to hold the turbine machine, compressors and the power generator together. The rotor's length is 7 m, its weight is about 1500 kg and the rotating speed is 15000 r/min. The structure of the rotor is so complicated that dynamic analysis of the rotor becomes difficult. One of the challenging problems is to exceed natural frequencies with enough stability and safety during reactor start up, power change and shutdown. The dynamic analysis of the rotor is the base for the design of control system. It is important for the rotor to exceed critical speeds. Some kinds of software and methods, such as MSC.Marc, Ansys, and the Transfer Matrix Method, are compared to fully analyze rotor dynamics characteristic in this paper. The modal analysis has been done for the HTR-10GT rotor. MSC.Marc was finally selected to analyze the vibration mode and the natural frequency of the rotor. The effects of AMB stiffness on the critical speeds of the rotor were studied. The design characteristics of the AMB control system for the HTR-10GT were studied and the related experiment to exceed natural frequencies was introduced. The experimental results demonstrate the system functions and validate the control scheme, which will be used in the HTR-10GT project. (authors)

  12. Study of the Effect of Burnable Poison Particles Applying in a Pebble Bed HTR

    International Nuclear Information System (INIS)

    Wei Chunlin; Zhao Jing; Zhang Jian; Xia Bing

    2014-01-01

    In pebble bed high temperature gas cooled reactors (HTR), spherical fuel elements pass through the core several times to balance the burnup process in the fuel region, resulting in an acceptable shape and peak factor of power density in the simulation analysis. In contrast, when fuel elements pass through the core only once, the peak of power density occurs at the top of the core and its value is too high to be safe. These indicators/parameters can be improved by incorporating burnable poison in the fuel elements under certain conditions. In the current study, burnable poison particles (BPPs) in fuel elements are evaluated. In spite of the strong absorption capability of "1"0B, BPPs can decrease the depletion speed and increase the duration of "1"0B because of the self-shielding effect, resulting in improved shape and peak factor of power distribution. Several BPPs with different radius are discussed in power distribution, following the calculation for a full-scale reactor core with modified VSOP code. According the result, applying BPPs on fuel pebbles is an effective means to improve the distribution of the power density under one-through fuel load in HTR. (author)

  13. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    International Nuclear Information System (INIS)

    Roberto, Thiago D.; Alvim, Antonio C.M.

    2017-01-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  14. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D.; Alvim, Antonio C.M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Lapa, Celso M.F., E-mail: thiagodbtr@gmail.com, E-mail: lapa@ien.gov.br, E-mail: alvim@nuclear.ufrj.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  15. Fuel cycle strategies and plutonium management in Europe

    International Nuclear Information System (INIS)

    Haas, D.

    2006-01-01

    Full text of publication follows: An overview will first be drawn on the present nuclear power production in Europe, and on the fuel cycle strategies followed in various countries. Indeed, for most countries in Europe, spent fuel is presently stored at reactor site, either for further conditioning and final deep geological storage, or in the expectation on a further decision on one of these options. Other countries have opted for spent fuel reprocessing as basic fuel cycle strategy. This option presents amongst other the merit of leading to a reduction of the final high level waste volume that will need in any case a geological disposal. It also allows the recovery from the spent fuel of energetic elements, principally the plutonium (Pu). The intention at the origin was to use this Pu in a fleet of fast breeder reactors, with the aim to make the best use of the uranium ore. However, fore reasons mainly linked to the low price of uranium (but also due to technical difficulties and political opposition), the project failed to develop. Now, the countries having followed the reprocessing option have to cope with an increasing amount of separated Pu. One industrial path, namely Pu recycling as MOX is presently operational in several European countries. It allows to reduce the increase in Pu stocks, but not to reduce it. Moreover, the quality of recycled Pu (in terms of its energetic value and of its radiotoxicity) decreases at each potential step of re-irradiation. Therefore, optimised or new ways of managing the Pu stocks in the medium and long term are required. The paper will present a status of the options considered today: - optimised Pu recycling as MOX in advanced LWR reactors (100 % MOX), as well as related optimised core designs; - utilisation of non-uranium matrices instead of uranium to prevent further Pu and minor actinides creation; - once-through Pu burning (up to high rates of destruction) in HTR fuel kernels; - multi-recycling of Pu in fast reactors of

  16. BWROPT: A multi-cycle BWR fuel cycle optimization code

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

    2015-09-15

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

  17. Fuel cycle related parametric study considering long lived actinide production, decay heat and fuel cycle performances

    International Nuclear Information System (INIS)

    Raepsaet, X.; Damian, F.; Lenain, R.; Lecomte, M.

    2001-01-01

    One of the very attractive HTGR reactor characteristics is its highly versatile and flexible core that can fulfil a wide range of diverse fuel cycles. Based on a GTMHR-600 MWth reactor, analyses of several fuel cycles were carried out without taking into account common fuel particle performance limits (burnup, fast fluence, temperature). These values are, however, indicated in each case. Fuel derived from uranium, thorium and a wide variety of plutonium grades has been considered. Long-lived actinide production and total residual decay heat were evaluated for the various types of fuel. The results presented in this papers provide a comparison of the potential and limits of each fuel cycle and allow to define specific cycles offering lowest actinide production and residual heat associated with a long life cycle. (author)

  18. 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.

  19. Fuel-cycle cost comparisons with oxide and silicide fuels

    International Nuclear Information System (INIS)

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

    1982-01-01

    This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed

  20. Simulation and study on reactivity disturbs dynamic character of HTR-10 nuclear power system

    International Nuclear Information System (INIS)

    Huang Xiaojin; Feng Yuankun

    2002-01-01

    In order to not only know 10 MW High Temperature Gas Cooled Reactor (HTR-10) nuclear power system's dynamic character more deeply but also to satisfy requirements of control system's design and analysis, the dynamic model of HTR-10 nuclear power system is established on the basis of dynamic model of HTR-10 nuclear system, which supplies turbine and generate electricity system model. Using this model, system's main variables' dynamic processes are simulated when control rod takes step reactivity disturb. The concussive progresses which is caused by reactivity disturb are analyzed. The results indicate that fuel temperature changing more slowly than nuclear power makes reactivity negative feedback not to restrain power changing, and then power concussive progress comes to being

  1. Closing the fuel cycle

    International Nuclear Information System (INIS)

    Aycoberry, C.; Rougeau, J.P.

    1987-01-01

    The progressive implementation of some key nuclear fuel cycle capecities in a country corresponds to a strategy for the acquisition of an independant energy source, France, Japan, and some European countries are engaged in such strategic programs. In France, COGEMA, the nuclear fuel company, has now completed the industrial demonstration of the closed fuel cycle. Its experience covers every step of the front-end and of the back-end: transportation of spent fuels, storage, reprocessing, wastes conditioning. The La Hague reprocessing plant smooth operation, as well as the large investment program under active progress can testify of full mastering of this industry. Together with other French and European companies, COGEMA is engaged in the recycling industry, both for uranium through conversion of uranyl nitrate for its further reeichment, and for plutonium through MOX fuel fabrication. Reprocessing and recycling offer the optimum solution for a complete, economic, safe and future-oriented fuel cycle, hence contributing to the necessary development of nuclear energy. (author)

  2. Development status and operational features of the high temperature gas-cooled reactor. Final report

    International Nuclear Information System (INIS)

    Winkleblack, R.K.

    1976-04-01

    The objective of this study is to investigate the maturity of HTR-technology and to look out for possible technical problems, concerning introduction of large HTR power plants into the market. Further state and problems of introducing and closing the thorium fuel cycle is presented and judged. Finally, the state of development of advanced HTR-concepts for electricity production, the direct cycle HTR with helium turbine, and the gas-cooled fast breeder is discussed. In preparing the study, both HTR concepts with spherical and block-type fuel elements have been considered

  3. Quadratic reactivity fuel cycle model

    International Nuclear Information System (INIS)

    Lewins, J.D.

    1985-01-01

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

  4. Fuel cycle

    International Nuclear Information System (INIS)

    Bahm, W.

    1989-01-01

    The situation of the nuclear fuel cycle for LWR type reactors in France and in the Federal Republic of Germany was presented in 14 lectures with the aim to compare the state-of-the-art in both countries. In addition to the momentarily changing fuilds of fuel element development and fueling strategies, the situation of reprocessing, made interesting by some recent developmnts, was portrayed and differences in ultimate waste disposal elucidated. (orig.) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

  6. Burn-up measurement in the HTR-module-reactor

    International Nuclear Information System (INIS)

    Gerhards, E.

    1993-05-01

    The burn-up status of spherical HTR-fuel elements is determined by a γ-spectrometric analysis of Cs-137 activity. The γ-spectrum recorded by a semiconductor detector up to now is analyzed by complex mathematical and time-consuming methods. For the operation of the HTR-Module-Reactor, however, a fast evaluation of the burn-up status is necessary. It is shown that this can be ensured by a comparison between the measured spectra and simulation results. Using the computer-program HTROGEN and the program system SPECCALC especially developed for this problem the γ-spectra are evaluated as a function of the burn-up status. The method is applied to results available from the operation of the AVR-reactor. The burn-up status determined with different methods corresponds very well within the limits of accuracy. (orig.)

  7. PCTR experiments with HTR lattice in MARIUS

    Energy Technology Data Exchange (ETDEWEB)

    Gambier, G; Estiot, J C; de Lapperent, D; Laponche, B; Luffin, J; Morier, F

    1972-06-15

    PCTR experiments have been carried out in Marius III with HTR tubular fuel, enriched to around 1% in order to reach K{sub infinity} = 1 and to reduce the mass of poison. Three poisons were used - Aluminium, Copper and Vanadium. The effect of air was measured and corrections were made to the results to allow the effect of delayed neutrons and the effect of axial heterogeneities. Interpretation was made with APOLLO. (auth)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-08-01

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

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

    International Nuclear Information System (INIS)

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

    2000-08-01

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

  10. Fuel cycle services

    International Nuclear Information System (INIS)

    Gruber, Gerhard J.

    1990-01-01

    TRIGA reactor operators are increasingly concerned about the back end of their Fuel Cycle due to a new environmental policy in the USA. The question how to close the Fuel Cycle will have to be answered by all operators sooner or later. Reprocessing of the TRIGA fuel elements is not available. Only long term storage and final disposal can be considered. But for such a storage or disposal a special treatment of the fuel elements and of course a final depository is necessary. NUKEM plans to undertake efforts to assist the TRIGA operators in this area. For that reason we need to know your special needs for today and tomorrow - so that potential processors can consider whether to offer these services on the market. (orig.)

  11. Physics of fusion-fuel cycles

    International Nuclear Information System (INIS)

    McNally, J.R. Jr.

    1981-01-01

    The evaluation of nuclear fusion fuels for a magnetic fusion economy must take into account the various technological impacts of the various fusion fuel cycles as well as the relative reactivity and the required β's and temperatures necessary for economic steady-state burns. This paper will review some of the physics of the various fusion fuel cycles (D-T, catalyzed D-D, D- 3 He, D- 6 Li, and the exotic fuels: 3 He 3 He and the proton-based fuels such as P- 6 Li, P- 9 Be, and P- 11 B) including such items as: (1) tritium inventory, burnup, and recycle, (2) neutrons, (3) condensable fuels and ashes, (4) direct electrical recovery prospects, (5) fissile breeding, etc. The advantages as well as the disadvantages of the different fusion fuel cycles will be discussed. The optimum fuel cycle from an overall standpoint of viability and potential technological considerations appears to be catalyzed D-D, which could also support smaller relatively clean, lean-D, rich- 3 He satellite reactors as well as fission reactors

  12. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    Boczar, P. G.; Fehrenbach, P. J.; Meneley, D. A.

    1996-01-01

    There are many reasons for countries embarking on a CANDU R program to start with the natural uranium fuel cycle. Simplicity of fuel design, ease of fabrication, and ready availability of natural uranium all help to localize the technology and to reduce reliance on foreign technology. Nonetheless, at some point, the incentives for using natural uranium fuel may be outweighed by the advantages of alternate fuel cycles. The excellent neutron economy, on-line refuelling, and simple fuel-bundle design provide an unsurpassed degree of fuel-cycle flexibility in CANDU reactors. The easiest first step in CANDU fuel-cycle evolution may be the use of slightly enriched uranium (SEU), including recovered uranium from reprocessed LWR spent fuel. Relatively low enrichment (up to 1.2%) will result in a two- to three-fold reduction in the quantity of spent fuel per unit energy production, reductions in fuel-cycle costs, and greater flexibility in the design of new reactors. The CANFLEX (CANDU FLEXible) fuel bundle would be the optimal fuel carrier. A country that has both CANDU and PWR reactors can exploit the natural synergism between these two reactor types to minimize overall waste production, and maximize energy derived from the fuel. This synergism can be exploited through several different fuel cycles. A high burnup CANDU MOX fuel design could be used to utilize plutonium from conventional reprocessing or more advanced reprocessing options (such as co-processing). DUPIC (Direct Use of Spent PWR Fuel In CANDU) represents a recycle option that has a higher degree of proliferation resistance than dose conventional reprocessing, since it uses only dry processes for converting spent PWR fuel into CANDU fuel, without separating the plutonium. Good progress is being made in the current KAERI, AECL, and U. S. Department of State program in demonstrating the technical feasibility of DUPIC. In the longer term, CANDU reactors offer even more dramatic synergistic fuel cycles with PWR or

  13. Homogeneous Thorium Fuel Cycles in Candu Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, B.; Dyck, G.R.; Edwards, G.W.R.; Magill, M. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada)

    2009-06-15

    The CANDU{sup R} reactor has an unsurpassed degree of fuel-cycle flexibility, as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle [1]. These features facilitate the introduction and full exploitation of thorium fuel cycles in Candu reactors in an evolutionary fashion. Because thorium itself does not contain a fissile isotope, neutrons must be provided by adding a fissile material, either within or outside of the thorium-based fuel. Those same Candu features that provide fuel-cycle flexibility also make possible many thorium fuel-cycle options. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up a 'mine' of U-233 in the spent fuel, for possible recycle in thermal reactors. When U-233 is recycled from the spent fuel, thorium-based fuel cycles in Candu reactors can provide substantial improvements in the efficiency of energy production from existing fissile resources. The fissile component driving the initial fuel could be enriched uranium, plutonium, or uranium-233. Many different thorium fuel cycle options have been studied at AECL [2,3]. This paper presents the results of recent homogeneous thorium fuel cycle calculations using plutonium and enriched uranium as driver fuels, with and without U-233 recycle. High and low burnup cases have been investigated for both the once-through and U-233 recycle cases. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). 1. Boczar, P.G. 'Candu Fuel-Cycle Vision', Presented at IAEA Technical Committee Meeting on 'Fuel Cycle Options for LWRs and HWRs', 1998 April 28 - May 01, also Atomic Energy

  14. Fission product release profiles from spherical HTR fuel elements at accident temperatures

    International Nuclear Information System (INIS)

    Schenk, W.; Pitzer, D.; Nabielek, H.

    1986-10-01

    A total of 22 fuel elements with modern TRISO particles has been tested in the temperature range 1500-2500 0 C. Additionally, release profiles of iodine and other isotopes have been obtained with seven UO 2 samples at 1400-1800 0 C. For heating times up to 100 hours at the maximum temperature, the following results are pertinent to HTR accident conditions: Ag 110 m is the only fission products to be released at 1200-1600 0 C by diffusion through intact SiC, but it is of low significance in accident assessments; cesium, iodine, strontium, and noble gas releases up to 1600 0 C are solely due to various forms of contamination; at 1700-1800 0 C, corrosion induced SiC defects cause the release of Cs, Sr, I/Xe/Kr; above 2000 0 C, thermal decomposition of the silicon carbide layer sets in while pyrocarbons still remain intact. Around 1600 0 C, the accident specific contribution of cesium, strontium, iodine, and noble gases is negligible. (orig./HP) [de

  15. Answering Key Fuel Cycle Questions

    International Nuclear Information System (INIS)

    Piet, S.J.; Dixon, B.W.; Bennett, R.G.; Smith, J.D.; Hill, R.N.

    2004-01-01

    Given the range of fuel cycle goals and criteria, and the wide range of fuel cycle options, how can the set of options eventually be narrowed in a transparent and justifiable fashion? It is impractical to develop all options. We suggest an approach that starts by considering a range of goals for the Advanced Fuel Cycle Initiative (AFCI) and then posits seven questions, such as whether Cs and Sr isotopes should be separated from spent fuel and, if so, what should be done with them. For each question, we consider which of the goals may be relevant to eventually providing answers. The AFCI program has both ''outcome'' and ''process'' goals because it must address both waste already accumulating as well as completing the fuel cycle in connection with advanced nuclear power plant concepts. The outcome objectives are waste geologic repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety. The process objectives are rea diness to proceed and adaptability and robustness in the face of uncertainties

  16. PEMODELAN TERAS UNTUK ANALISIS PERHITUNGAN KONSTANTA MULTIPLIKASI REAKTOR HTR-PROTEUS

    Directory of Open Access Journals (Sweden)

    Zuhair Zuhair

    2015-04-01

    Full Text Available PTRKN sebagai salah satu unit kerja di BATAN dengan tugas pokok dan fungsi yang berkaitan erat dengan teknologi reaktor dan keselamatan nuklir, menaruh perhatian khusus pada konsep reaktor pebble bed. Dalam makalah ini pemodelan reaktor pebble bed HTR-PROTEUS dilakukan dengan program transport Monte Carlo MCNP5. Partikel bahan bakar berlapis TRISO dimodelkan secara detail dan eksak dimana distribusi acak partikel ini dalam bola bahan bakar didekati menggunakan array teratur kisi SC dengan fraksi packing 5,76% tanpa zona eksklusif. Model teras pebble bed didekati dengan memanfaatkan kisi teratur dari bola yang disusun sebagai kisi BCC berdasarkan sel berulang yang digenerasi dari sejumlah sel satuan. Hasil perhitungan MCNP5 memperlihatkan kesesuaian yang sangat baik dengan eksperimen, walaupun teras HTR-PROTEUS diprediksi lebih reaktif daripada pengukuran, khususnya di teras 4.2 dan 4.3. Pustaka ENDF/B-VI menunjukkan konsistensi dengan estimasi keff paling akurat dibandingkan pustaka ENDF/B-V, terutama ENDF/B-VI (66c. Deviasi estimasi keff yang dihitung dengan eksperimen dikaitkan sebagai konsekuensi dari komposisi reflektor grafit yang dispesifikasikan. Komparasi yang dibuat memperlihatkan bahwa MCNP5 menghasilkan keff teras HTR-PROTEUS lebih presisi daripada hasil dari MCNP4B dan MCNPBALL. Hasil ini menyimpulkan bahwa, sukses metodologi pemodelan ini menjustifikasi aplikasi MCNP5 untuk analisis reaktor pebble bed lainnya. Kata kunci: pemodelan teras HTR-PROTEUS, konstanta multiplikasi, MCNP5   PTRKN as a working unit in BATAN whose main duties and functions are related to reactor technology and nuclear safety, consern attention to pebble bed reactor concept. In this paper modeling of HTR-PROTEUS pebble bed reactor was done using Monte Carlo transport code MCNP5. The TRISO coated fuel particle is modeled in detailed and exact manner where random distributions of these particles in fuel pebble is approximated by using regular array of SC lattice

  17. Solid TRU fuels and fuel cycle technology

    International Nuclear Information System (INIS)

    Ogawa, Toru; Suzuki, Yasufumi

    1997-01-01

    Alloys and nitrides are candidate solid fuels for transmutation. However, the nitride fuels are preferred to the alloys because they have more favorable thermal properties which allows to apply a cold-fuel concept. The nitride fuel cycle technology is briefly presented

  18. National Policy on Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Soedyartomo, S.

    1996-01-01

    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

  19. The thorium fuel cycle

    International Nuclear Information System (INIS)

    Merz, E.R.

    1977-01-01

    The utilization of the thorium fuel cycle has long since been considered attractive owing to the excellent neutronic characteristics of 233 U, and the widespread and cheap thorium resources. Rapidly increasing uranium prices, public reluctance for widespread Pu recycling and expected delays for the market penetration of fast breeders have led to a reconsideration of the thorium fuel cycle merits. In addition, problems associated with reprocessing and waste handling, particularly with re-fabrication by remote handling of 233 U, are certainly not appreciably more difficult than for Pu recycling. To divert from uranium as a nuclear energy source it seems worth while intensifying future efforts for closing the Th/ 233 U fuel cycle. HTGRs are particularly promising for economic application. However, further research and development activities should not concentrate on this reactor type alone. Light- and heavy-water-moderated reactors, and even future fast breeders, may just as well take advantage of a demonstrated thorium fuel cycle. (author)

  20. LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

    International Nuclear Information System (INIS)

    Brogli, R.; Bucher, K.H.; Chawla, R.; Foskolos, K.; Luchsinger, H.; Mathews, D.; Sarlos, G.; Seiler, R.

    1990-01-01

    New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

  1. LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

    Energy Technology Data Exchange (ETDEWEB)

    Brogli, R; Bucher, K H; Chawla, R; Foskolos, K; Luchsinger, H; Mathews, D; Sarlos, G; Seiler, R [Paul Scherrer Institute, Laboratory for Reactor Physics and System Technology Wuerenlingen and Villigen, Villigen PSI (Switzerland)

    1990-07-01

    New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

  2. Survey of appropriate endothermic processes for association with the HTR

    International Nuclear Information System (INIS)

    Brown, G.; Harrison, G.E.; Gent, C.W.; Plummer, J.

    1975-01-01

    Emphasis is placed on association of the HTR system as a heat source with chemical processes requiring temperatures up to 850 to 900 0 C, corresponding to a reactor coolant temperature of 950 0 C, though processes requiring temperatures up to 1100 0 C and above are reviewed. Particular attention is given to processes for the production of hydrogen-containing gases, including coal/lignite gasification which has been the subject of a recent study. Rising fuel prices make the HTR an attractive proposition if design concepts and materials can be developed to match the requirements. Other appropriate endothermic processes considered are oil processing, including tar sands and shales, and also energy production. Since the full temperature range of the reactor system must be utilised mention is made of low grade heat uses. Even very large chemical works have relatively small energy requirement by nuclear heat standards and adoption of the HTR as a heat source is likely only if it is associated with a large chemical/metallurgical complex or with the processing of a natural resource. (author)

  3. Objectives for an HTR R and D physics programme

    Energy Technology Data Exchange (ETDEWEB)

    Johnstone, I; Scott, J A

    1973-10-15

    The paper reviews important objectives for an HTR R and D programme and the importance of particular characteristics for safety and reactor performance is discussed. Uncertainties in the physics characteristics influence reactor design through the inclusion of design margins and contingency allowances and may even eliminate certain design variants. The paper discusses quantitatively the relationship between some important uncertainties and reactor design and operating costs and derives targets for the precision with which it should be possible to compute the corresponding physics characteristics. To extrapolate to power reactor conditions, the need for a comprehensive computational scheme validated by an adequate experimental programme is emphasised. The reduction in uncertainty as the theoretical and experimental reactor physics development proceeds in order to meet the desired target uncertainty is illustrated with respect to the UK's programme in support of the low enriched HTR concept. The current situation for this concept is discussed and compared briefly with that for the Th cycle HTR for which somewhat less zero energy experimental data are available. (auth)

  4. Plutonium in an enduring fuel cycle

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1998-05-01

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

  5. Fast breeder fuel cycle

    International Nuclear Information System (INIS)

    1978-09-01

    Basic elements of the ex-reactor part of the fuel cycle (reprocessing, fabrication, waste handling and transportation) are described. Possible technical and proliferation measures are evaluated, including current methods of accountability, surveillance and protection. The reference oxide based cycle and advanced cycles based on carbide and metallic fuels are considered utilizing conventional processes; advanced nonaqueous reprocessing is also considered. This contribution provides a comprehensive data base for evaluation of proliferation risks

  6. IAEA activities on nuclear fuel cycle 1997

    Energy Technology Data Exchange (ETDEWEB)

    Oi, N [International Atomic Energy Agency, Vienna (Austria). Nuclear Fuel Cycle and Materials Section

    1997-12-01

    The presentation discussing the IAEA activities on nuclear fuel cycle reviews the following issues: organizational charts of IAEA, division of nuclear power and the fuel cycle, nuclear fuel cycle and materials section; 1997 budget estimates; budget trends; the nuclear fuel cycle programme.

  7. IAEA activities on nuclear fuel cycle 1997

    International Nuclear Information System (INIS)

    Oi, N.

    1997-01-01

    The presentation discussing the IAEA activities on nuclear fuel cycle reviews the following issues: organizational charts of IAEA, division of nuclear power and the fuel cycle, nuclear fuel cycle and materials section; 1997 budget estimates; budget trends; the nuclear fuel cycle programme

  8. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-01-01

    Status of different nuclear fuel cycle phases in 1992 is discussed including the following issues: uranium exploration, resources, supply and demand, production, market prices, conversion, enrichment; reactor fuel technology; spent fuel management, as well as trends of these phases development up to the year 2010. 10 refs, 11 figs, 15 tabs

  9. The uranium-plutonium breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    Salmon, A.; Allardice, R.H.

    1979-01-01

    All power-producing systems have an associated fuel cycle covering the history of the fuel from its source to its eventual sink. Most, if not all, of the processes of extraction, preparation, generation, reprocessing, waste treatment and transportation are involved. With thermal nuclear reactors more than one fuel cycle is possible, however it is probable that the uranium-plutonium fuel cycle will become predominant; in this cycle the fuel is mined, usually enriched, fabricated, used and then reprocessed. The useful components of the fuel, the uranium and the plutonium, are then available for further use, the waste products are treated and disposed of safely. This particular thermal reactor fuel cycle is essential if the fast breeder reactor (FBR) using plutonium as its major fuel is to be used in a power-producing system, because it provides the necessary initial plutonium to get the system started. In this paper the authors only consider the FBR using plutonium as its major fuel, at present it is the type envisaged in all, current national plans for FBR power systems. The corresponding fuel cycle, the uranium-plutonium breeder reactor fuel cycle, is basically the same as the thermal reactor fuel cycle - the fuel is used and then reprocessed to separate the useful components from the waste products, the useful uranium and plutonium are used again and the waste disposed of safely. However the details of the cycle are significantly different from those of the thermal reactor cycle. (Auth.)

  10. Fuel cycle cost study with HEU and LEU fuels

    International Nuclear Information System (INIS)

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

    1984-01-01

    Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  12. The thorium fuel cycle

    International Nuclear Information System (INIS)

    Merz, E.R.

    1977-01-01

    The utilization of the thorium fuel cycle has long since been considered attractive due to the excellent neutronic characteristics of 233 U, and the widespread and cheap thorium resources. Although the uranium ore as well as the separative work requirements are usually lower for any thorium-based fuel cycle in comparison to present uranium-plutonium fuel cycles of thermal water reactors, interest by nuclear industry has hitherto been marginal. Fast increasing uranium prices, public reluctance against widespread Pu-recycling and expected retardations for the market penetration of fast breeders have led to a reconsideration of the thorium fuel cycle merits. In addition, it could be learned in the meantime that problems associated with reprocessing and waste handling, but particularly with a remote refabrication of 233 U are certainly not appreciably more difficult than for Pu-recycling. This may not only be due to psychological constraints but be based upon technological as well as economical facts, which have been mostly neglected up till now. In order to diversify from uranium as a nuclear energy source it seems to be worthwhile to greatly intensify efforts in the future for closing the Th/ 233 U fuel cycle. HTGR's are particularly promising for economic application. However, further R and D activites should not be solely focussed on this reactor type alone. Light and heavy-water moderated reactors, as well as even fast breeders later on, may just as well take advantage of a demonstrated thorium fuel cycle. A summary is presented of the state-of-the-art of Th/ 233 U-recycling technology and the efforts still necessary to demonstrate this technology all the way through to its industrial application

  13. World nuclear fuel cycle requirements 1991

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-10

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs.

  14. World nuclear fuel cycle requirements 1991

    International Nuclear Information System (INIS)

    1991-01-01

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, ''burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs

  15. Evaluation of the HTR-10 Reactor as a Benchmark for Physics Code QA

    International Nuclear Information System (INIS)

    William K. Terry; Soon Sam Kim; Leland M. Montierth; Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-01-01

    The HTR-10 is a small (10 MWt) pebble-bed research reactor intended to develop pebble-bed reactor (PBR) technology in China. It will be used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and co-generation of heat, and provide experience in PBR design, operation, and construction. As the only currently operating PBR in the world, the HTR-10 can provide data of great interest to everyone involved in PBR technology. In particular, if it yields data of sufficient quality, it can be used as a benchmark for assessing the accuracy of computer codes proposed for use in PBR analysis. This paper summarizes the evaluation for the International Reactor Physics Experiment Evaluation Project (IRPhEP) of data obtained in measurements of the HTR-10's initial criticality experiment for use as benchmarks for reactor physics codes

  16. IFR fuel cycle demonstration in the EBR-II Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; Rigg, R.H.; Benedict, R.W.; Carnes, M.D.; Herceg, J.E.; Holtz, R.E.

    1991-01-01

    The next major milestone of the IFR (Integral Fast Reactor) program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase which includes completion of facility modifications, and installation and cold checkout of process equipment. This paper reviews the design and construction of the facility, the design and fabrication of the process equipment, and the schedule and initial plan for its operation. (author)

  17. IFR fuel cycle demonstration in the EBR-II Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; Rigg, R.H.; Benedict, R.W.; Carnes, M.D.; Herceg, J.E.; Holtz, R.E.

    1991-01-01

    The next major milestone of the IFR program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase which includes completion of facility modifications, and installation and cold checkout of process equipment. This paper reviews the design and construction of the facility, the design and fabrication of the process equipment, and the schedule and initial plan for its operation. 5 refs., 4 figs

  18. DUPIC fuel cycle economics assessment (1)

    International Nuclear Information System (INIS)

    Choi, H. B.; Roh, G. H.; Kim, D. H.

    1999-04-01

    This is a state-of-art report that describes the current status of the DUPIC fuel cycle economics analysis conducted by the DUPIC fuel compatibility assessment group of the DUPIC fuel development project. For the DUPIC fuel cycle economics analysis, the DUPIC fuel compatibility assessment group has organized the 1st technical meeting composed of 8 domestic specialists from government, academy, industry, etc. and a foreign specialist of hot-cell design from TRI on July 16, 1998. This report contains the presentation material of the 1st technical meeting, published date used for the economics analysis and opinions of participants, which could be utilized for further DUPIC fuel cycle and back-end fuel cycle economics analyses. (author). 11 refs., 7 charts

  19. IFR fuel cycle

    International Nuclear Information System (INIS)

    Battles, J.E.; Miller, W.E.; Lineberry, M.J.; Phipps, R.D.

    1992-01-01

    The next major milestone of the IFR program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase, which includes completion of facility modifications and installation and cold checkout of process equipment. This paper reviews the development of the electrorefining pyroprocess, the design and construction of the facility for the hot demonstration, the design and fabrication of the equipment, and the schedule and initial plan for its operation

  20. Fuel cycle management in Finland

    International Nuclear Information System (INIS)

    Vaeyrynen, H.; Mikkola, I.

    1987-01-01

    Both Finnish utilities producing nuclear power - Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (Industrial Power Co. Ltd, TVO) - have created efficient fuel cycle management systems. The systems however differ in almost all respects. The reason is that the principal supplier for IVO is the Soviet Union and for TVO is Sweden. A common feature of both systems at the front end of the cycle is the building of stockpiles in order to provide for interruptions in fuel deliveries. Quality assurance supervision at the fuel factory for IVO is regulated by the Soviet Chamber of Commerce and Industry and a final control is made in Finland. The in-core fuel management is done by IVO using codes developed in Finland. The whole IVO fuel cycle is basically a leasing arrangement. The spent fuel is returned to the USSR after five years cooling. TVO carries out the in-core fuel management using a computer code system supplied by Asea-Atom. TVO is responsable for the back end of the cycle and makes preparations for the final disposal of the spent fuel in Finland. 6 refs., 2 figs

  1. The DUPIC alternative for backend fuel cycle

    International Nuclear Information System (INIS)

    Lee, J.S.; Yang, M.S.; Park, H.S.; Boczar, P.; Sullivan, J.; Gadsby, R.D.

    1997-01-01

    The DUPIC fuel cycle was conceived as an alternative to the conventional fuel cycle backed options, with a view to multiple benefits expectable from burning spent PWR fuel again in CANDU reactors. It is based on the basic idea that the bulk of spent PWR fuel can be directly refabricated into a reusable fuel for CANDU of which high efficiency in neutron utilization would exhaustively burn the fissile remnants in the spent PWR fuel to a level below that of natural uranium. Such ''burn again'' strategy of the DUPIC fuel cycle implies that the spent PWR fuel will become CANDU fuel of higher burnup with relevant benefits such as spent PWR fuel disposition, saving of natural uranium fuel, etc. A salient feature of the DUPIC fuel cycle is neither the fissile content nor the bulk radioactivity is separated from the DUPIC mass flow which must be contained and shielded all along the cycle. This feature can be considered as a factor of proliferation resistance by deterrence against access to sensitive materials. It means also the requirement for remote systems technologies for DUPIC fuel operation. The conflicting aspects between better safeguardability and harder engineering problems of the radioactive fuel operation may be the important reason why the decades' old concept, since INFCE, of ''hot'' fuel cycle has not been pursued with much progress. In this context, the DUPIC fuel cycle could be a live example for development of proliferation resistant fuel cycle. As the DUPIC fuel cycle looks for synergism of fuel linkage from PWR to CANDU (or in broader sense LWR to HWR), Korea occupies a best position for DUPIC exercise with her unique strategy of reactor mix of both reactor types. But the DUPIC benefits can be extended to global bonus, expectable from successful development of the technology. (author)

  2. Fuel cycles using adulterated plutonium

    International Nuclear Information System (INIS)

    Brooksbank, R.E.; Bigelow, J.E.; Campbell, D.O.; Kitts, F.G.; Lindauer, R.B.

    1978-01-01

    Adjustments in the U-Pu fuel cycle necessitated by decisions made to improve the nonproliferation objectives of the US are examined. The uranium-based fuel cycle, using bred plutonium to provide the fissile enrichment, is the fuel system with the highest degree of commercial development at the present time. However, because purified plutonium can be used in weapons, this fuel cycle is potentially vulnerable to diversion of that plutonium. It does appear that there are technologically sound ways in which the plutonium might be adulterated by admixture with 238 U and/or radioisotopes, and maintained in that state throughout the fuel cycle, so that the likelihood of a successful diversion is small. Adulteration of the plutonium in this manner would have relatively little effect on the operations of existing or planned reactors. Studies now in progress should show within a year or two whether the less expensive coprocessing scheme would provide adequate protection (coupled perhaps with elaborate conventional safeguards procedures) or if the more expensive spiked fuel cycle is needed as in the proposed civex pocess. If the latter is the case, it will be further necessary to determine the optimum spiking level, which could vary as much as a factor of a billion. A very basic question hangs on these determinations: What is to be the nature of the recycle fuel fabrication facilities. If the hot, fully remote fuel fabrication is required, then a great deal of further development work will be required to make the full cycle fully commercial

  3. Dynamic Simulations of Advanced Fuel Cycles

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  4. Details of modelling HTR core physics: the use of pseudo nuclide traces

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Oppe, J.; Haas, J.B.M. de; Da Cruz, D.F.

    2003-01-01

    At present most combined neutronic and thermal hydraulic analyses of reactors, and the HTR is no exception, are being performed by codes employing few-group (typically 2-group) neutronics on the basis of parametrized few-group macroscopic (and microscopic) cross sections for homogenized areas, depending on quantities like irradiation (fuel only), 135 Xe concentration, temperature, etc. The irradiation parameter (time-integrated power per unit initial heavy metal mass) is sufficient for keeping track of the evolution of areas containing fuel. However, the use of the same parameter in areas without fuel, e.g. containing burnable poison, requires some special provisions. This can be met by the introduction of pseudo nuclides, with very specific cross sections and reaction chains, in the procedure to generate the parametrized few-group cross sections. It is shown that the time-evolution of a non-fuelled burnable poison area, as calculated by the 2-group (HTR) reactor code PANTHERMIX employing pseudo nuclides, compares well to the time-evolution obtained from an explicit burnup calculation by the WIMS8A/SNAP code. Examples are also shown using the pseudo nuclide method to keep track of the fast fluence (time-integrated flux above 0.1 MeV) in a continuous reload pebble-bed HTR reactor calculation by PANTHERMIX. Although the present implementation of the pseudo nuclide method exhibits some peculiarities connected to the specific codes in use (WIMS8A and PANTHERMIX) it is considered to be sufficiently general to be applicable in other code suites, requiring only limited modifications. (authors)

  5. Details of modelling HTR core physics: the use of pseudo nuclide traces

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.; Oppe, J.; Haas, J.B.M. de; Da Cruz, D.F. [Nuclear Research and consultancy Group (NRG), Petten (Netherlands)

    2003-07-01

    At present most combined neutronic and thermal hydraulic analyses of reactors, and the HTR is no exception, are being performed by codes employing few-group (typically 2-group) neutronics on the basis of parametrized few-group macroscopic (and microscopic) cross sections for homogenized areas, depending on quantities like irradiation (fuel only), {sup 135}Xe concentration, temperature, etc. The irradiation parameter (time-integrated power per unit initial heavy metal mass) is sufficient for keeping track of the evolution of areas containing fuel. However, the use of the same parameter in areas without fuel, e.g. containing burnable poison, requires some special provisions. This can be met by the introduction of pseudo nuclides, with very specific cross sections and reaction chains, in the procedure to generate the parametrized few-group cross sections. It is shown that the time-evolution of a non-fuelled burnable poison area, as calculated by the 2-group (HTR) reactor code PANTHERMIX employing pseudo nuclides, compares well to the time-evolution obtained from an explicit burnup calculation by the WIMS8A/SNAP code. Examples are also shown using the pseudo nuclide method to keep track of the fast fluence (time-integrated flux above 0.1 MeV) in a continuous reload pebble-bed HTR reactor calculation by PANTHERMIX. Although the present implementation of the pseudo nuclide method exhibits some peculiarities connected to the specific codes in use (WIMS8A and PANTHERMIX) it is considered to be sufficiently general to be applicable in other code suites, requiring only limited modifications. (authors)

  6. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    Boczar, P.G.; Fehrenbach, P.J.; Meneley, D.A.

    1996-04-01

    The easiest first step in CANDU fuel-cycle evolution may be the use of slightly enriched uranium (SEU), including recovered uranium from reprocessed LWR spent fuel. Relatively low enrichment (up to 1.2%) will result in a twoto three-fold reduction in the quantity of spent fuel per unit energy production, reductions in fuel-cycle costs, and greater flexibility in the design of new reactors. The CANFLEX (CANDU FLEXible) fuel bundle would be the optimal fuel carrier. A country that has both CANDU and PWR reactors can exploit the natural synergism between these two reactor types to minimize overall waste production, and maximize energy derived from the fuel. This synergism can be exploited through several different fuel cycles. A high burnup CANDU MOX fuel design could be used to utilize plutonium from conventional reprocessing or more advanced reprocessing options (such as co-processing). DUPIC (Direct Use of Spent PWR Fuel In CANDU) represents a recycle option that has a higher degree of proliferation resistance than does conventional reprocessing, since it uses only dry processes for converting spent PWR fuel into CANDU fuel, without separating the plutonium. Good progress is being made in the current KAERI, AECL, and U.S. Department of State program in demonstrating the technical feasibility of DUPIC. In the longer term, CANDU reactors offer even more dramatic synergistic fuel cycles with PWR or FBR reactors. If the objective of a national fuel-cycle program is the minimization of actinide waste or destruction of long-lived fission products, then studies have shown the superiority of CANDU reactors in meeting this objective. Long-term energy security can be assured either through the thorium cycle or through a CANDU 1 FBR system, in which the FBR would be operated as a 'fuel factory,' providing the fissile material to power a number of lower-cost, high efficiency CANDU reactors. In summary, the CANDU reactor's simple fuel design, high neutron economy, and on

  7. Recent IAEA activities on CANDU-PHWR fuels and fuel cycles

    International Nuclear Information System (INIS)

    Inozemtsev, V.; Ganguly, C.

    2005-01-01

    Pressurized Heavy Water Reactors (PHWR), widely known as CANDU, are in operation in Argentina, Canada, China, India, Pakistan, Republic of Korea and Romania and account for about 6% of the world's nuclear electricity production. The CANDU reactor and its fuel have several unique features, like horizontal calandria and coolant tubes, on-power fuel loading, thin-walled collapsible clad coated with graphite on the inner surface, very high density (>96%TD) natural uranium oxide fuel and amenability to slightly enriched uranium oxide, mixed uranium plutonium oxide (MOX), mixed thorium plutonium oxide, mixed thorium uranium (U-233) oxide and inert matrix fuels. Several Technical Working Groups (TWG) of IAEA periodically discuss and review CANDU reactors, its fuel and fuel cycle options. These include TWGs on water-cooled nuclear power reactor Fuel Performance and Technology (TWGFPT), on Nuclear Fuel Cycle Options and spent fuel management (TWGNFCO) and on Heavy Water Reactors (TWGHWR). In addition, IAEA-INPRO project also covers Advanced CANDU Reactors (ACR) and DUPIC fuel cycles. The present paper summarises the Agency's activities in CANDU fuel and fuel cycle, highlighting the progress during the last two years. In the past we saw HWR and LWR technologies and fuel cycles separate, but nowadays their interaction is obviously growing, and their mutual influence may have a synergetic character if we look at the world nuclear fuel cycle as at an integrated system where the both are important elements in line with fast neutron, gas cooled and other advanced reactors. As an international organization the IAEA considers this challenge and makes concrete steps to tackle it for the benefit of all Member States. (author)

  8. Economic aspects of Dukovany NPP fuel cycle

    International Nuclear Information System (INIS)

    Vesely, P.; Borovicka, M.

    2001-01-01

    The paper discusses some aspects of high burnup program implementation at Dukovany NPP and its influence on the fuel cycle costs. Dukovany internal fuel cycle is originally designed as a three years cycle of the Out-In-In fuel reloading patterns. These reloads are not only uneconomical but they additionally increased the radiation load of the reactor pressure vessel due to high neutron leakage typical for Out-In-In loading pattern. To avoid the high neutron leakage from the core a transition to 4-year fuel cycle is started in 1987. The neutron leakage from the core is sequentially decreased by insertion of older fuel assemblies at the core periphery. Other developments in fuel cycle are: 1) increasing of enrichment in control assemblies (3.6% of U-235); 2) improvement in fuel assembly design (reduce the assembly shroud thickness from 2.1 to 1.6 mm); 3) introduction of Zr spacer grid instead of stainless steel; 4) introduction of new type of assembly with profiled enrichment with average value of 3.82%. Due to increased reactivity of the new assemblies the transition to the partial 5-year fuel cycle is required. Typical fuel loading pattern for 3, 3.5, 4 and 5-year cycles are shown in the presented paper. An evaluation of fuel cost is also discussed by using comparative analysis of different fuel cycle options. The analysis shows that introduction of the high burnup program has decrease relative fuel cycle costs

  9. Critical evaluation of the experiments and mathematical models for the determination of fission product release from the spherical fuel elements in cases of core heating accidents in modular HTR's

    International Nuclear Information System (INIS)

    Bailly, H.W.

    1987-01-01

    In this work, the thermal behaviour of modular reactors in cases of core heating accidents and the physical phenomena relevant for a release of radioactive materials from HTR fuel elements are explained as far as is necessary for understanding the work. The present mathematical models by which the release of radioactive materials from HTR fuel elements due to diffusion or breaking particles in cases of core heating accidents are also described, examined and evaluated with regard to their applicability to module reactors. The experiments used to verify the mathematical models are also evaluated. The mathematical models are in nearly all cases computer programs, which describe the complicated process of releasing radioactive materials quantitative mathematically. One should point out that these models are constantly being developed further, in line with the increasing amount of knowledge. To conclude the work, proposals are made for improving the certainty of information from experiments and mathematical models to determine the release behaviour of modular reactors. (orig./GL) [de

  10. Fuel cycle oriented approach

    International Nuclear Information System (INIS)

    Petit, A.

    1987-01-01

    The term fuel cycle oriented approach is currently used to designate two quite different things: the attempt to consider all or part of a national fuel cycle as one material balance area (MBA) or to consider individual MBAs existing in a state while designing a unique safeguards approach for each and applying the principle of nondiscrimination to fuel cycles as a whole, rather than to individual facilities. The merits of such an approach are acceptability by the industry and comparison with the contemplated establishment of long-term criteria. The following points concern the acceptability by the industry: (1) The main interest of the industry is to keep an open international market and therefore, to have effective and efficient safeguards. (2) The main concerns of the industry regarding international safeguards are economic burden, intrusiveness, and discrimination. Answers to these legitimate concerns, which retain the benefits of a fuel cycle oriented approach, are needed. More specifically, the problem of reimbursing the operator the costs that he has incurred for the safeguards must be considered

  11. Analysis of possible fuel cycles

    International Nuclear Information System (INIS)

    Boehm, H.; Kessler, G.; Engelmann, P.; Maerkl, H.; Stoll, W.

    1978-01-01

    A brief survey is presented of the most important fuel cycles. A rough analysis of fuel cycles is attempted under the aspects of proliferation, status of technical feasibility, resource conservation and waste management and the most important criteria for such an analysis are discussed. Among the multitude of potential combinations of fuel cycles and types of reactors only a few have reached a level of technical feasibility which would make them eligible for commercial implementation within the next decade. However, if, for instance, the higher proliferation resistance of a specific fuel cycle is to be utilized to diminish the worldwide proliferation hazard, that cycle would first of all have to be introduced on an industrial scale as quickly as possible. The analysis shows that the reduction of the bazard of worldwide proliferation will continue to be the objective primarily of international agreements and measures taken in the political realm. (orig.) [de

  12. Feasibility study on tandem fuel cycle

    International Nuclear Information System (INIS)

    Han, P.S.; Suh, I.S.; Rim, C.S.; Kim, B.K.; Suh, K.S.; Ro, S.K.; Juhn, P.I.; Kim, S.Y.

    1983-01-01

    The objective of this feasibility study is to review and assess the current state of technology concerning the tandem fuel cycle. Based on the results from this study, a long-term development plan suitable for Korea has been proposed for this cycle, i.e., the PWR → CANDU tandem fuel cycle which used plutonium and uranium, recovered from spent PWR fuel by co-processing, as fuel material for CANDU reactors. (Author)

  13. French development program on fuel cycle

    International Nuclear Information System (INIS)

    Viala, M.; Bourgeois, M.

    1991-01-01

    The need to close the fuel cycle of fast reactors makes the development of the cycle installations (fuel fabrication, irradiated assembly conditioning before reprocessing, reprocessing and waste management) especially independent with the development of the reactor. French experience with the integrated cycle over a period of about 25 years, the tonnage of fuels fabricated (more than 100 t of mixed oxides) for the Rapsodie, Phoenix and SuperPhoenix reactors, and the tonnage of reprocessed fuel (nearly 30 t of plutonium fuel) demonstrate the control of the cycle operations. The capacities of the cycle installations in existence and under construction are largely adequate for presents needs, even including a new European EFR reactor. They include the Cadarache fuel fabrication complex, the La Hague UP2-800 reprocessing plant, and the Marcoule pilot facility. Short- and medium-term R and D programs are connected with fuel developments, with the primary objective of very high burnups. For the longer term and for a specific plant to reprocess fast reactor fuels, the programs could concern new fabrication and reprocessing systems and the study of the consequences of the reduction in fuel out-of-core time

  14. Proliferation resistance fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-01

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

  15. Thorium fuel cycle - Potential benefits and challenges

    International Nuclear Information System (INIS)

    2005-05-01

    There has been significant interest among Member States in developing advanced and innovative technologies for safe, proliferation resistant and economically efficient nuclear fuel cycles, while minimizing waste and environmental impacts. This publication provides an insight into the reasons for renewed interest in the thorium fuel cycle, different implementation scenarios and options for the thorium cycle and an update of the information base on thorium fuels and fuel cycles. The present TECDOC focuses on the upcoming thorium based reactors, current information base, front and back end issues, including manufacturing and reprocessing of thorium fuels and waste management, proliferation-resistance and economic issues. The concluding chapter summarizes future prospects and recommendations pertaining to thorium fuels and fuel cycles

  16. Sustainability Features of Nuclear Fuel Cycle Options

    Directory of Open Access Journals (Sweden)

    Stefano Passerini

    2012-09-01

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

  17. Synergistic fuel cycles of the future

    International Nuclear Information System (INIS)

    Meneley, D.A.; Dastur, A.R.

    1997-01-01

    Good neutron economy is the basis of the fuel cycle flexibility in the CANDU reactor. This paper describes the fuel cycle options available to the CANDU owner with special emphasis on resource conservation and waste management. CANDU fuel cycles with low initial fissile content operate with relatively high conversion ratio. The natural uranium cycle provides over 55 % of energy from the plutonium that is created during fuel life. Resource utilization is over 7 MWd/kg NU. This can be improved by slight enrichment (between 0.9 and 1.2 wt % U235) of the fuel. Resource utilization increases to 11 MWd/kg NU with the Slightly Enriched Uranium cycle. Thorium based cycles in CANDU operate at near-breeder efficiency. Obey provide attractive options when used with natural uranium or separated (reactor grade and weapons grade) plutonium as driver fuels. In the latter case, the energy from the U233 plus the initial plutonium content amounts to 3.4 GW(th).d/kg Pu-fissile. The same utilization is expected from the use of FBR plutonium in a CANDU thorium cycle. Extension of natural resource is achieved by the use of spent fuels in CANDU. The LWR/CANDU Tandem cycle leads to an additional 77 % of energy through the use of reprocessed LWR fuel (which has a fissile content of 1.6 wt %) in CANDU. Dry reprocessing of LWR fuel with the OREOX process (a more safeguardable alternative to the PUREX process) provides an additional 50 % energy. Uranium recovered (RU) from separation of plutonium contained in spent LWR fuel provides an additional 15 MWd/kg RU. CANDU's low fissile requirement provides the possibility, through the use of non-fertile targets, of extracting energy from the minor actinides contained in spent fuel. In addition to the resource utilization advantage described above, there is a corresponding reduction in waste arisings with such cycles. This is especially significant when separated plutonium is available as a fissile resource. (author)

  18. Nuclear fuel cycles : description, demand and supply estimates

    International Nuclear Information System (INIS)

    Gadallah, A.A.; Abou Zahra, A.A.; Hammad, F.H.

    1985-01-01

    This report deals with various nuclear fuel cycles description as well as the world demand and supply estimates of materials and services. Estimates of world nuclear fuel cycle requirements: nuclear fuel, heavy water and other fuel cycle services as well as the availability and production capabilities of these requirements, are discussed for several reactor fuel cycle strategies, different operating and under construction fuel cycle facilities in some industrialized and developed countries are surveyed. Various uncertainties and bottlenecks which are recently facing the development of some fuel cycle components are also discussed, as well as various proposals concerning fuel cycle back-end concepts. finally, the nuclear fuel cycles activities in some developing countries are reviewed with emphasis on the egyptian plans to introduce nuclear power in the country. 11 fig., 16 tab

  19. The Nuclear Fuel Cycle Information System

    International Nuclear Information System (INIS)

    1987-02-01

    The Nuclear Fuel Cycle Information System (NFCIS) is an international directory of civilian nuclear fuel cycle facilities. Its purpose is to identify existing and planned nuclear fuel cycle facilities throughout the world and to indicate their main parameters. It includes information on facilities for uranium ore processing, refining, conversion and enrichment, for fuel fabrication, away-from-reactor storage of spent fuel and reprocessing, and for the production of zirconium metal and Zircaloy tubing. NFCIS currently covers 271 facilities in 32 countries and includes 171 references

  20. The development on-line monitoring system of active magnetic bearings for HTR-10GT

    International Nuclear Information System (INIS)

    Shi Zhengang; Shi Lei; Zha Meisheng; Yu Suyuan

    2005-01-01

    High Temperature Gas-cooled Reactor (HTR) is recognized as an advanced type of reactor incorporating many design enhancements such as inherent safety features, fuel cycle flexibility, highly fuel utilization, highly efficient electricity generation and process heat application. The research and development of HTR started at the middle of the 1970's, and came to be a part of the Chinese High Technology Program in 1986. A plan to build a 10 MW High Temperature Gas-cooled Reactor (HTR-10) was approved by the State Science and Technology Commission in 1990, and in 1995 the construction was initiated at the Institute of Nuclear Energy Technology (INET), Tsinghua University. The full power 10 MW operation for 72 hours have reached in 2003, and have been checked and accepted by the State Science and Technology Commission. In order to advance the HTR-10 performance, the project of the Helium Gas Turbine Generator for the HTR-10 was authorized by the State Science and Technology Commission, and stared in 2003. In this project, active magnetic bearings (AMBs) are chosen to support the generator rotor and the turbocompressor rotor in the power conversion unit because of their numerous advantages over the conventional bearings. In order to detect how the AMB system works in operation and make diagnosis whether the system behaves normally or not, the monitoring system based on the virtual instruments is designed to monitor the working conditions of the PCU, and to ensure its normal operation. This monitoring system consists of the industry personal computer (PC), the data acquisition system, the measurement transmitters and the LabVIEW system platform. It is located at the PCU control room, and communicates with the master control room by Controller Area Net (CAN). The development is divided into the following three steps: First, a data acquisition platform to collect and acquire all the necessary and useful data from the operation of the AMB system is developed. Second, the

  1. Nuclear-fuel-cycle optimization: methods and modelling techniques

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1982-01-01

    This book present methods applicable to analyzing fuel-cycle logistics and optimization as well as in evaluating the economics of different reactor strategies. After an introduction to the phases of a fuel cycle, uranium cost trends are assessed in a global perspective. Subsequent chapters deal with the fuel-cycle problems faced by a power utility. The fuel-cycle models cover the entire cycle from the supply of uranium to the disposition of spent fuel. The chapter headings are: Nuclear Fuel Cycle, Uranium Supply and Demand, Basic Model of the LWR (light water reactor) Fuel Cycle, Resolution of Uncertainties, Assessment of Proliferation Risks, Multigoal Optimization, Generalized Fuel-Cycle Models, Reactor Strategy Calculations, and Interface with Energy Strategies. 47 references, 34 figures, 25 tables

  2. 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.

  3. 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.

  4. 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.

  5. Fuel cycles for the 80's

    International Nuclear Information System (INIS)

    1980-01-01

    Papers presented at the American Nuclear Society's topical meeting on the fuel cycle are summarized. Present progress and goals in the areas of fuel fabrication, fuel reprocessing, spent fuel storage, accountability, and safeguards are reported. Present governmental policies which affect the fuel cycle are also discussed. Individual presentations are processed for inclusion in the Energy Data Base

  6. Fuel cycle centres

    International Nuclear Information System (INIS)

    Hagen, M.

    1977-01-01

    The concept of co-locating and integrating fuel cycle facilities at one site is discussed. This concept offers considerable advantages, especially in minimizing the amount of radioactive material to be transported on public roads. Safeguards and physical protection as relating to such an integrated system of facilities are analysed in detail, also industrial and commercial questions. An overall risk-benefit evaluation turns out to be in favour of fuel cycle centres. These centres seem to be specifically attractive with regard to the back end of the fuel cycle, including on-site disposal of radioactive wastes. The respective German approach is presented as an example. Special emphasis is given to the site selection procedures in this case. Time scale and cost for the implementation of this concept are important factors to be looked at. Since participation of governmental institutions in these centres seems to be indispensable their respective roles as compared to industry must be clearly defined. The idea of adjusting fuel cycle centres to regional rather than national use might be an attractive option, depending on the specific parameters in the region, though results of existing multinational ventures are inconclusive in this respect. Major difficulties might be expected e.g. because of different national safety regulations and standards as well as commercial conditions among partner countries. Public acceptance in the host country seems to be another stumbling block for the realization of this type of multinational facilities

  7. HTR's role in process heat applications

    International Nuclear Information System (INIS)

    Kuhr, Reiner

    2008-01-01

    Advanced high-temperature nuclear reactors create a number of new opportunities for nuclear process heat applications. These opportunities are based on the high-temperature heat available, smaller reactor sizes, and enhanced safety features that allow siting close to process plants. Major sources of value include the displacement of premium fuels and the elimination of CO 2 emissions from combustion of conventional fuels and their use to produce hydrogen. High value applications include steam production and cogeneration, steam methane reforming, and water splitting. Market entry by advanced high-temperature reactor technology is challenged by the evolution of nuclear licensing requirements in countries targeted for early applications, by the development of a customer base not familiar with nuclear technology and related issues, by convergence of oil industry and nuclear industry risk management, by development of public and government policy support, by resolution of nuclear waste and proliferation concerns, and by the development of new business entities and business models to support commercialization. New HTR designs may see a larger opportunity in process heat niche applications than in power given competition from larger advanced light water reactors. Technology development is required in many areas to enable these new applications, including the commercialization of new heat exchangers capable of operating at high temperatures and pressures, convective process reactors and suitable catalysts, water splitting system and component designs, and other process-side requirements. Key forces that will shape these markets include future fuel availability and pricing, implementation and monetization of CO 2 emission limits, and the formation of international energy and environmental policy that will support initiatives to provide the nuclear licensing frameworks and risk distribution needed to support private investment. This paper was developed based on a plenary

  8. Nuclear fuel cycle, nuclear fuel makes the rounds: choosing a closed fuel cycle, nuclear fuel cycle processes, front-end of the fuel cycle: from crude ore to enriched uranium, back-end of the fuel cycle: the second life of nuclear fuel, and tomorrow: multiple recycling while generating increasingly less waste

    International Nuclear Information System (INIS)

    Philippon, Patrick

    2016-01-01

    France has opted for a policy of processing and recycling spent fuel. This option has already been deployed commercially since the 1990's, but will reach its full potential with the fourth generation. The CEA developed the processes in use today, and is pursuing research to improve, extend, and adapt these technologies to tomorrow's challenges. France has opted for a 'closed cycle' to recycle the reusable materials in spent fuel (uranium and plutonium) and optimise ultimate waste management. France has opted for a 'closed' nuclear fuel cycle. Spent fuel is processed to recover the reusable materials: uranium and plutonium. The remaining components (fission products and minor actinides) are the ultimate waste. This info-graphic shows the main steps in the fuel cycle currently implemented commercially in France. From the mine to the reactor, a vast industrial system ensures the conversion of uranium contained in the ore to obtain uranium oxide (UOX) fuel pellets. Selective extraction, purification, enrichment - key scientific and technical challenges for the teams in the Nuclear Energy Division (DEN). The back-end stages of the fuel cycle for recycling the reusable materials in spent fuel and conditioning the final waste-forms have reached maturity. CEA teams are pursuing their research in support of industry to optimise these processes. Multi-recycle plutonium, make even better use of uranium resources and, over the longer term, explore the possibility of transmuting the most highly radioactive waste: these are the challenges facing future nuclear systems. (authors)

  9. Enhancing the Sustainability of Nuclear Power: the Pebble Bed HTR in Deep Burn Mode

    International Nuclear Information System (INIS)

    Da Cruz, D.F.; De Haas, J.B.M.; Van Heek, A.I.

    2004-01-01

    The scenario of a utility in an industrialized country starting new nuclear construction with a single PBMR reactor has been considered. To make the new construction project acceptable by government and society, a maximum effort to obtain sustainability (i.e. minimization of resource use and waste production) will have to be shown. Therefore the usual open cycle for HTR has been abandoned, and the spent fuel will be reprocessed once. The long-lived transuranic (TRU) elements Pu, Np, Am and Cm are all re-fabricated into so-called transmutation fuel elements, and loaded back into the same reactor, in our case a 110 MWe PBMR with low-enriched uranium cycle. In this study, the reactor physical prospects have been investigated: to what extent the amount of TRU could be reduced. In this way, 75% of the initial amount of TRU waste is being destructed, while the time span in which the waste is more radio-toxic than uranium ore is being reduced to one-third. Also, the amount of fresh driver fuel needed is decreases by 25%. A preliminary cost analysis has been performed as well. It shows that there is also a cost advantage of operating the reactor in Deep Burn mode in industrialized countries, where the waste storage fees charged per volume are relatively high. (authors)

  10. Calculation of the Fission Product Release for the HTR-10 based on its Operation History

    International Nuclear Information System (INIS)

    Xhonneux, A.; Druska, C.; Struth, S.; Allelein, H.-J.

    2014-01-01

    Since the first criticality of the HTR-10 test reactor in 2000, a rather complex operation history was performed. As the HTR-10 is the only pebble bed reactor in operation today delivering experimental data for HTR simulation codes, an attempt was made to simulate the whole reactor operation up to the presence. Special emphasis was put on the fission product release behaviour as it is an important safety aspect of such a reactor. The operation history has to be simulated with respect to the neutronics, fluid mechanics and depletion to get a detailed knowledge about the time-dependent nuclide inventory. In this paper we report about such a simulation with VSOP 99/11 and our new fission product release code STACY. While STACY (Source Term Analysis Code System) so far was able to calculate the fission product release rates in case of an equilibrium core and during transients, it now can also be applied to running-in-phases. This coupling demonstrates a first step towards an HCP Prototype. Based on the published power histogram of the HTR-10 and additional information about the fuel loading and shuffling, a coupled neutronics, fluid dynamics and depletion calculation was performed. Special emphasis was put on the complex fuel-shuffling scheme within both VSOP and STACY. The simulations have shown that the HTR-10 up to now generated about 2580 MWd while reshuffling the core about 2.3 times. Within this paper, STACY results for the equilibrium core will be compared with FRESCO-II results being published by INET. Compared to these release rates, which are based on a few user defined life histories, in this new approach the fission product release rates of Ag-110m, Cs-137, Sr-90 and I-131 have been simulated for about 4000 tracer pebbles with STACY. For the calculation of the HTR-10 operation history time-dependent release rates are being presented as well. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  12. Development of FR fuel cycle in japan (1) development scope of fuel cycle technology

    International Nuclear Information System (INIS)

    Nakamura, H.; Funasaka, H.; Namekawa, T.

    2008-01-01

    A fast reactor (FR) cycle has a potential to realize a sustainable energy supply system that is harmonized with environment by fully recycling both uranium (U) and transuranium (TRU) elements. In Japan, a Feasibility Study on Commercialized FR Cycle Systems (FS) was launched in July 1999, and through two different study phases, a final report was presented in 2006. As a result of FS, a combined system of sodium-cooled FR with mixed-oxide (MOX) fuel, advanced aqueous reprocessing and simplified pelletizing fuel fabrication was considered to be most promising for commercialization. The advanced aqueous reprocessing system, which is called the New Extraction system for TRU recovery (NEXT), consists of a U crystallization process for the bulk of U recovery, a simplified solvent extraction process for residual U, plutonium (Pu) and neptunium (Np) without Pu partitioning and purification, and a process for recovering americium (Am) and curium (Cm) from the raffinate. The ratio of Pu/U concentration in the mother solution after crystallization is adequate for MOX fuel fabrication, and thus complicated powder mixing processes for adjusting Pu content in MOX fuel can be eliminated in the subsequent simplified fuel fabrication system. In this system, lubricant-mixing process can also be eliminated by adopting the advanced technology in which lubricant is coated on the inner surface of a die before fuel powder supply. Such a simplification could help us overcoming the difficulty to treat MA bearing fuel powders in a hot cell. Ministry of Education, Culture, Sports, Science and Technology (MEXT) reviewed these results of FS in 2006 and identified the most promising FR cycle concept proposed in the FS phase II study as a mainline choice for commercialization. According to such a governmental assessment, R and D activities of FR cycle systems were decided to be concentrated mainly to the innovative technology development for the mainline concept. The stage of R and D project was

  13. Advanced nuclear fuel cycles activities in IAEA

    International Nuclear Information System (INIS)

    Nawada, H.P.; Ganguly, C.

    2007-01-01

    Full text of publication follows. Of late several developments in reprocessing areas along with advances in fuel design and robotics have led to immense interest in partitioning and transmutation (P and T). The R and D efforts in the P and T area are being paid increased attention as potential answers to ever-growing issues threatening sustainability, environmental protection and non-proliferation. Any fuel cycle studies that integrate partitioning and transmutation are also known as ''advanced fuel cycles'' (AFC), that could incinerate plutonium and minor actinide (MA) elements (namely Am, Np, Cm, etc.) which are the main contributors to long-term radiotoxicity. The R and D efforts in developing these innovative fuel cycles as well as reactors are being co-ordinated by international initiatives such as Innovative Nuclear Power Reactors and Fuel Cycles (INPRO), the Generation IV International Forum (GIF) and the Global Nuclear Energy Partnership (GENP). For these advanced nuclear fuel cycle schemes to take shape, the development of liquid-metal-cooled reactor fuel cycles would be the most essential step for implementation of P and T. Some member states are also evaluating other concepts involving the use of thorium fuel cycle or inert-matrix fuel or coated particle fuel. Advanced fuel cycle involving novel partitioning methods such as pyrochemical separation methods to recover the transuranic elements are being developed by some member states which would form a critical stage of P and T. However, methods that can achieve a very high reduction (>99.5%) of MA and long-lived fission products in the waste streams after partitioning must be achieved to realize the goal of an improved protection of the environment. In addition, the development of MA-based fuel is also an essential and crucial step for transmutation of these transuranic elements. The presentation intends to describe progress of the IAEA activities encompassing the following subject-areas: minimization of

  14. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Patarin, L.

    2002-01-01

    This book treats of the different aspects of the industrial operations linked with the nuclear fuel, before and after its use in nuclear reactors. The basis science of this nuclear fuel cycle is chemistry. Thus a recall of the elementary notions of chemistry is given in order to understand the phenomena involved in the ore processing, in the isotope enrichment, in the fabrication of fuel pellets and rods (front-end of the cycle), in the extraction of recyclable materials (residual uranium and plutonium), and in the processing and conditioning of wastes (back-end of the fuel cycle). Nuclear reactors produce about 80% of the French electric power and the Cogema group makes 40% of its turnover at the export. Thus this book contains also some economic and geopolitical data in order to clearly position the stakes. The last part, devoted to the management of wastes, presents the solutions already operational and also the research studies in progress. (J.S.)

  15. Fuel cycle and waste management: A perspective from British nuclear fuels plc

    International Nuclear Information System (INIS)

    Holmes, R.G.G.; Fairhall, G.A.; Robbins, R.A.

    1996-01-01

    The phrase fuel cycle and waste management implies two separate and distinct activities. British Nuclear Fuels plc (BNFL) has adopted a holistic approach to the fuel cycle that integrates the traditional fuel cycle activities of conversion to uranium hexafluoride, fuel fabrication, power generation, and reprocessing with waste arisings, its subsequent treatment, and disposal

  16. A fuel cycle cost study with HEU and LEU fuels

    International Nuclear Information System (INIS)

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

    1985-01-01

    Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

  17. A fuel cycle cost study with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Matos, J E; Freese, K E [Argonne National Laboratory, Argonne, IL (United States)

    1985-07-01

    Fuel cycle costs are compared for a range of {sup 235}U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

  18. Characteristics of fuel cycle waste

    International Nuclear Information System (INIS)

    Aquilina, C.A.; Everette, S.E.

    1982-01-01

    The Low-Level Waste Management System started in 1979 to describe and model the existing commercial low-level waste management system. The system description produced is based on the identification of the different elements making up both the fuel and non-fuel cycle and their relationships to each other. A systems model based on the system description can accurately reflect the flow of low-level waste from generator to disposal site and is only limited by the reliability of the information it uses. For both the fuel cycle and non-fuel cycle large quantities of information is required in order to allow the system to operate at its full potential. Work is ongoing to collect this information. Significant progress is being made in the fuel cycle area primarily because the majority of fuel cycle low-level radioactive waste is produced by commercial power reactor plant operations. The Low-Level Waste Management system is only beginning to derive the benefits to be obtained from an accurate low-level waste management information system. As data is verified and analyzed, results on a national as well as individual organization level will be gained. Comparisons to previous studies will be made. Accurate projections of waste volumes and activities to be produced, projected impacts of waste streams of treatment or management changes are only examples of information to be produced. 1 figure, 1 table

  19. Development of a dry-mechanical graphite separation process and elimination of the separated carbon for the reprocessing of spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Kronschnabel, H.

    1982-01-01

    Due to the C-14 distribution the separation of the particle-free outer region of the spherical HTR fuel element with subsequent solidification of the separated carbon makes it possible to reduce by half the remaining C-14 inventory in the inner particle region to be further treated. Separation of the particle-free outer region by a newly developed sphere-peeling milling machine, conditioning the graphite into compacts and in-situ cementation into a salt-mine are the basic elements of this head-end process variation. An annual cavern volume of approx. 2000 m 3 will be needed to ultimately store the graphite of the particle-free outer region, which corresponds to a reprocessing capacity of 50 GWsub(e) installed HTR power. The brush-disintegration of the remaining inner particle region and the resulting peel-brush-preparation are capable of separating 95% of the graphite without any heavy metal losses. With the mentioned reprocessing capacity an annual cavern volume of approx. 16.500 m 3 is required. (orig.) [de

  20. WWER-440 fuel cycles possibilities using improved fuel assemblies design

    International Nuclear Information System (INIS)

    Mikolas, P.; Svarny, J.

    2008-01-01

    Practically five years cycle has been achieved in the last years at NPP Dukovany. There are two principal means how it could be achieved. First, it is necessary to use fuel assemblies with higher fuel enrichment and second, to use fuel loading with very low leakage. Both these conditions are fulfilled at NPP Dukovany at this time. It is known, that the fuel cycle economy can be improved by increasing the fuel residence time in the core up to six years. There are at least two ways how this goal could be achieved. The simplest way is to increase enrichment in fuel. There exists a limit, which is 5.0 w % of 235 U. Taking into account some uncertainty, the calculation maximum is 4.95 w % of 235 U. The second way is to change fuel assembly design. There are several possibilities, which seem to be suitable from the neutron - physical point of view. The first one is higher mass content of uranium in a fuel assembly. The next possibility is to enlarge pin pitch. The last possibility is to 'omit' FA shroud. This is practically unrealistic; anyway, some other structural parts must be introduced. The basic neutron physical characteristics of these cycles for up-rated power are presented showing that the possibilities of fuel assemblies with this improved design in enlargement of fuel cycles are very promising. In the end, on the basis of neutron physical characteristics and necessary economical input parameters, a preliminary evaluation of economic contribution of proposals of advanced fuel assemblies on fuel cycle economy is presented (Authors)

  1. Back end of an enduring fuel cycle

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1998-03-01

    An enduring nuclear fuel cycle is an essential part of sustainable consumption, the process whereby world's riches are consumed in a responsible manner so that future generations can continue to enjoy at least some of them. In many countries, the goal of sustainable development has focused attention on the benefits of nuclear technologies. However, sustenance of the nuclear fuel cycle is dependent on sensible management of all the resources of the fuel cycle, including energy, spent fuels, and all of its side streams. The nuclear fuel cycle for energy production has suffered many traumas since the mid seventies. The common basis of technologies producing nuclear explosives and consumable nuclear energy has been a preoccupation for some, predicament for others, and a perception problem for many. It is essential to reestablish a reliable back end of the nuclear fuel cycle that can sustain the resource requirements of an enduring full cycle. This paper identifies some pragmatic steps necessary to reverse the trend and to maintain a necessary fuel cycle option for the future

  2. Fuel cycle based safeguards

    International Nuclear Information System (INIS)

    De Montmollin, J.M.; Higinbotham, W.A.; Gupta, D.

    1985-07-01

    In NPT safeguards the same model approach and absolute-quantity inspection goals are applied at present to all similar facilities, irrespective of the State's fuel cycle. There is a continuing interest and activity on the part of the IAEA in new NPT safeguards approaches that more directly address a State's nuclear activities as a whole. This fuel cycle based safeguards system is expected to a) provide a statement of findings for the entire State rather than only for individual facilities; b) allocate inspection efforts so as to reflect more realistically the different categories of nuclear materials in the different parts of the fuel cycle and c) provide more timely and better coordinated information on the inputs, outputs and inventories of nuclear materials in a State. (orig./RF) [de

  3. Fuel cycle parameters for strategy studies

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1979-05-01

    This report summarizes seven fuel cycle parameters (efficiency, specific power, burnup, equilibrium net fissile feed, equilibrium net fissile surplus, first charge fissile content, and whether or not fuel reprocessing is required) to be used in long-term strategy analyses of fuel cycles based on natural UO 2 , low enriched uranium, mixed oxides, plutonium topped thorium, uranium topped thorium, and the fast breeder oxide cycle. (LL)

  4. Alternative fuel cycles and non-proliferation aspects

    International Nuclear Information System (INIS)

    Kessler, G.

    1980-10-01

    The most important physical characteristics of the U/Pu and the Th/U fuel cycles and the technical data of the most significant converter reactors operating with Th/U fuel are outlined in the report. Near breeders as well as breeders with a thermal neutron spectrum are briefly discussed, and the potential of breeders with fast neutron spectra in the Th/U fuel is outlined. The essential criteria for the comparison of the alternative fuel cycles with the reference Pu/U cycle are the consumption of natural uranium, the numbers of U-233 producing and U-233 consuming converter reactors and the amounts of fission material transported and handled within the fuel cycle (reprocessing, refabrication). Although the alternative U/Th fuel cycles are feasible with some advantages and some disadvantages as compared to the reference U/Pu cycle, not much experience has so far been gathered with pilot plants of the fuel cycle. The respective status in reprocessing, refabrication and waste disposal is briefly discussed. Finally, a comparison of the risk potential inherent in secular storage is presented and questions of resistance to proliferation and of safeguards of the U/Th cycle are discussed

  5. Status of the HTR 500 design program

    International Nuclear Information System (INIS)

    Baust, E.; Arndt, E.

    1987-01-01

    Since 1982 BBC/HRB have offered the HTR 500 as the follow-on project of the THTR 300, the first large pebble bed reactor. The technical concept of the HTR-500 largely corresponds to the THTR 300 which has been in operation for almost 2 years now. In developing the design concept of the HTR 500 the ideas and demands of the reactor users in the FRG interested in the HTR have been taken into consideration to a large extent. In 1982 these potential users formed a working group 'Arbeitsgemeinschaft Hochtemperaturreaktor' (AHR), representing 16 power indusry companies and in early 1983, awarded a contract to HRB to perform a conceptual design study on the HTR 500. Within this conceptual design study BBC/HRB developed the safety concept of the HTR 500, prepared a detailed description of the overall power plant, and performed a cost calculation. These activities were completed in 1984. Based on the positive results of this conceptual design study, BBC/HRB are expecting to be granted a design contract by the users company Hochtemperaturreaktor GmbH (HRG) to establish the final complete design plans and documents for the HTR 500. (author)

  6. Globalization of the nuclear fuel cycle impact of developments on fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Van Den Durpel, L.; Bertel, E. [OCDE-NEA, Nuclear Development Div., 92 - Issy-les-Moulineaux (France)

    1999-07-01

    Nuclear energy will have to cope more and more with a rapid changing environment due to economic competitive pressure and the de-regulatory progress. In current economic environment, utilities will have to focus strongly on the reduction of their total generation costs, covering the fuel cycle costs, which are only partly under their control. Developments in the fuel cycle will be in the short-term rather evolutionary addressing the current needs of utilities. However, within the context of sustainable development and more and more inclusion of externalities in energy generation costs, more performing developments in the fuel cycle could become important and feasible. A life-cycle design approach of the fuel cycle will be requested in order to cover all factors in order to decrease significantly the nuclear energy generation cost to compete with other alternative fuels in the long-term. This paper will report on some of the trends one could distinguish in the fuel cycle with emphasis on cost reduction. OECD/NEA is currently conducting a study on the fuel cycle aiming to assess current and future nuclear fuel cycles according the potential for further improvement of the full added-value chain of these cycles from a mainly technological and economical perspective including environmental and social considerations. (authors)

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

    International Nuclear Information System (INIS)

    Durpel, L. van den; Bertel, E.

    2000-01-01

    Nuclear energy will have to cope more and more with a rapid changing environment due to economic competitive pressure and the deregulatory progress. In current economic environment, utilities will have to focus strongly on the reduction of their total generation costs, covering the fuel cycle costs, which are only partly under their control. Developments in the fuel cycle will be in the short-term rather evolutionary addressing the current needs of utilities. However, within the context of sustainable development and more and more inclusion of externalities in energy generation costs, more performing developments in the fuel cycle could become important and feasible. A life-cycle design approach of the fuel cycle will be requested in order to cover all factors in order to decrease significantly the nuclear energy generation cost to complete with other alternative fuels in the long-term. This paper will report on some of the trends one could distinguish in the fuel cycle with emphasis on cost reduction. OECD/NEA is currently conducting a study on the fuel cycle aiming to assess current and future nuclear fuel cycles according to the potential for further improvement of the full added-value chain of these cycles from a mainly technological and economic perspective including environmental and social considerations. (orig.) [de

  8. ITER fuel cycle systems layout

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-10-01

    The ITER fuel cycle building (FCB) will contain the following systems: fuel purification - permeator based; fuel purification - molecular sieves; impurity treatment; waste water storage and treatment; isotope separation; waste water tritium extraction; tritium extraction from solid breeder; tritium extraction from test modules; tritium storage, shipping and receiving; tritium laboratory; atmosphere detritiation systems; fuel cycle control centre; tritiated equipment maintenance space; control maintenance space; health physics laboratory; access, access control and facilities. The layout of the FCB and the requirements for these systems are described. (10 figs.)

  9. The evolving nuclear fuel cycle

    International Nuclear Information System (INIS)

    Gale, J.D.; Hanson, G.E.; Coleman, T.A.

    1993-01-01

    Various economics and political pressures have shaped the evolution of nuclear fuel cycles over the past 10 to 15 yr. Future trends will no doubt be similarly driven. This paper discusses the influences that long cycles, high discharge burnups, fuel reliability, and costs will have on the future nuclear cycle. Maintaining the economic viability of nuclear generation is a key issue facing many utilities. Nuclear fuel has been a tremendous bargain for utilities, helping to offset major increases in operation and maintenance (O ampersand M) expenses. An important factor in reducing O ampersand M costs is increasing capacity factor by eliminating outages

  10. Introducing advanced nuclear fuel cycles in Canada

    International Nuclear Information System (INIS)

    Duret, M.F.

    1978-05-01

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

  11. The second Euratom sponsored 9000C HTR fuel irradiation experiment in the HFR Petten Project E 96.02: Pt.2. Post-irradiation examination

    International Nuclear Information System (INIS)

    Roettger, R.; Bueger, J. de; Schoots, T.

    1977-01-01

    A large variety of HTR fuel specimens, loose coated particles, coupons and compacts provided by Belgonucleaire, the Dragon Project and the KFA Juelich have been irradiated in the HFR at Petten at about 900 0 C up to a maximum fast neutron fluence of about 7x10 21 cm -2 (EDN) as a Euratom sponsored experiment. The maximum burn-ups were between 11 and 18.5% FIMA. The results of the post-irradiation examinations, comprising visual inspection, dimensional measurements, microradiography, metallography, and burn-up determinations are presented in this part 2 of the final report. The examinations have shown that the endurance limit of most of the tested fuel varieties is beyond the reached irradiation values

  12. Advanced nuclear fuel cycles and radioactive waste management

    International Nuclear Information System (INIS)

    2006-01-01

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

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

    KAUST Repository

    Jaasim, Mohammed

    2017-03-13

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

  14. The economics of thorium fuel cycles

    International Nuclear Information System (INIS)

    James, R.A.

    1978-01-01

    The individual cost components and the total fuel cycle costs for natural uranium and thorium fuel cycles are discussed. The thorium cycles are initiated by using either enriched uranium or plutonium. Subsequent thorium cycles utilize recycled uranium-233 and, where necessary, either uranium-235 or plutonium as topping. A calculation is performed to establish the economic conditions under which thorium cycles are economically attractive. (auth)

  15. Innovation in the fuel cycle industry

    International Nuclear Information System (INIS)

    Lamorlette, Guy

    1998-01-01

    The fuel cycle industry will have to adapt to the production of new fuel and in the same time will have to improve its performance. Innovation will be a key factor of success. Innovation must be driven by the needs of the fuel cycle industry to achieve. The fuel cycle requirement of tomorrow, Innovative processes for mining high grade uranium, Innovative enrichment process, Sorting the pellets at Melox plant, Innovation in action, and Innovative waste management in la Hague are presented. A number of innovative solutions are already implemented and are in action on industrial facilities. As problems are becoming more and more tough to address, international cooperation will be required. The fuel cycle industry, as a part of the nuclear power industry, is committed to seek improvements through performance upgrade and innovation. (Cho. G. S.). 10 refs., 4 figs

  16. Survey of HTR related research at IRI, Delft, Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Wallerbos, E.J.M.; Van der Hagen, T.H.J.J.; Van Dam, H. [Interfaculty Reactor Institute IRI, Delft University of Technology, Delft (Netherlands); Tuerkcan, E. [ECN Nuclear Research, Petten (Netherlands)

    1998-09-01

    High temperature helium-cooled reactors have a large potential for inherent safety. Therefore, several projects on HTR research are being carried out or were carried out at the Interfaculty Reactor Institute (IRI) of the Delft University of Technology in Delft, Netherlands. As part of a larger research programme measurements of core reactivity, reactivity worth of safety rods and of small samples being oscillated in the reactor core were carried out at the PROTEUS facility of the Paul Scherrer Institute at Villigen, Switzerland. Together with other partners in the Netherlands a small inherently safe co-generation plant with a pebble-bed HTR core was designed and analysed. It was verified that such a reactor can operate continuously for 10 years by adding continuously fuel pebbles until the maximum available core height is reached. As a new, innovative, inherently safe reactor type the design of a fluidized-bed reactor with coated fuel particles on a helium gas stream is discussed and results are shown for the analysis of inherent criticality safety under varying coolant flow rates. IRI is also taking part in the new IAEA Co-ordinated Research Programme, which involves participation in the start-up experiments of the Japanese HTTR and carrying out calculations for the core physics benchmark test. 11 refs.

  17. HTR-10 management information system

    International Nuclear Information System (INIS)

    Liu Ruoxiao; Wu Zhongwang; Xi Shuren

    2000-01-01

    The HTR-10 Management information system (REMIS) strengthens the managerial level and usage of the information of HTR-10, thereby enhances the ability and efficiency of the design and management work. REMIS is designed based on the Client/Server framework. Database management system is SQL Server 6.5 for NT, While the client side is developed by Borland C ++ Builder, and it is based on Windows 95/98. The network protocol is TCP/IP. REMIS collects date of the HTR-10 at four parameters: Reactor properties, Design parameters, Equipment properties Reactor system flow charts. Final discussing extended prospect of REMIS

  18. The safety of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2005-01-01

    The procurement and preparation of fuel for nuclear power reactors, followed by its recovery, processing and management subsequent to reactor discharge, are frequently referred to as the ''front end'' and ''back end'' of the nuclear fuel cycle. The facilities associated with these activities have an extensive and well-documented safety record accumulated over the past 50 years by technical experts and safety authorities. This information has enabled an in-depth analysis of the complete fuel cycle. Preceded by two previous editions in 1981 and 1993, this new edition of the Safety of the Nuclear Fuel Cycle represents the most up-to-date analysis of the safety aspects of the nuclear fuel cycle. It will be of considerable interest to nuclear safety experts, but also to those wishing to acquire extensive information about the fuel cycle more generally. (author)

  19. The safety of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2005-10-01

    The procurement and preparation of fuel for nuclear power reactors, followed by its recovery, processing and management subsequent to reactor discharge, are frequently referred to as the 'front end' and 'back end' of the nuclear fuel cycle. The facilities associated with these activities have an extensive and well-documented safety record accumulated over the past 50 years by technical experts and safety authorities. This information has enabled an in-depth analysis of the complete fuel cycle. Preceded by two previous editions in 1981 and 1993, this new edition of The Safety of the Nuclear Fuel Cycle represents the most up-to-date analysis of the safety aspects of the nuclear fuel cycle. It will be of considerable interest to nuclear safety experts, but also to those wishing to acquire extensive information about the fuel cycle more generally. (author)

  20. Coupling fuel cycles with repositories: how repository institutional choices may impact fuel cycle design

    International Nuclear Information System (INIS)

    Forsberg, C.; Miller, W.F.

    2013-01-01

    The historical repository siting strategy in the United States has been a top-down approach driven by federal government decision making but it has been a failure. This policy has led to dispatching fuel cycle facilities in different states. The U.S. government is now considering an alternative repository siting strategy based on voluntary agreements with state governments. If that occurs, state governments become key decision makers. They have different priorities. Those priorities may change the characteristics of the repository and the fuel cycle. State government priorities, when considering hosting a repository, are safety, financial incentives and jobs. It follows that states will demand that a repository be the center of the back end of the fuel cycle as a condition of hosting it. For example, states will push for collocation of transportation services, safeguards training, and navy/private SNF (Spent Nuclear Fuel) inspection at the repository site. Such activities would more than double local employment relative to what was planned for the Yucca Mountain-type repository. States may demand (1) the right to take future title of the SNF so if recycle became economic the reprocessing plant would be built at the repository site and (2) the right of a certain fraction of the repository capacity for foreign SNF. That would open the future option of leasing of fuel to foreign utilities with disposal of the SNF in the repository but with the state-government condition that the front-end fuel-cycle enrichment and fuel fabrication facilities be located in that state

  1. Fuel-management simulations for once-through thorium fuel cycle in CANDU reactors

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Boczar, P.G.; Ellis, R.J.; Ardeshiri, F.

    1999-01-01

    High neutron economy, on-power refuelling and a simple fuel bundle design result in unsurpassed fuel cycle flexibility for CANDU reactors. These features facilitate the introduction and exploitation of thorium fuel cycles in existing CANDU reactors in an evolutionary fashion. Detailed full-core fuel-management simulations concluded that a once-through thorium fuel cycle can be successfully implemented in an existing CANDU reactor without requiring major modifications. (author)

  2. High-conversion HTRs and their fuel cycle

    International Nuclear Information System (INIS)

    Gutmann, H.; Hansen, U.; Larsen, H.; Price, M.S.T.

    1976-01-01

    The high-temperature reactors using graphite as structural core material and helium as coolant represent thermal reactor designs with a very high degree of neutron economy which, when using the thorium fuel cycle, offer, at least theoretically, the possibility of thermal breeding. Though this was already known from previous studies, the economic climate at that time was such that the achievement of high conversion ratios conflicted with the incentive for low fuel cycle costs. Consequently, thorium cycle conversion ratios of around 0.6 were found optimum, and the core and fuel element layout followed from the economic ground rules. The recent change in attitude, brought about partly by the slow process of realization of the limits to the earth's accessible high-grade uranium ore resources and more dramatically by the oil crisis, makes it necessary to concentrate attention again on the high conversion fuel cycles. This report discusses the principles of the core design and the fuel cycle layout for High Conversion HTRs (HCHTRs). Though most of the principles apply equally to HTRs of the pebble-bed and the prismatic fuel element design types, the paper concentrates on the latter. Design and fuel cycle strategies for the full utilization of the high conversion potential are compared with others that aim at easier reprocessing and the ''environmental'' fuel cycle. The paper concludes by discussing operating and fuel cycle characteristics and economics of HCHTRs, and how the latter impinge on the allowable price for uranium ore and the available uranium resources. (author)

  3. Fuel cycle cost analysis on molten-salt reactors

    International Nuclear Information System (INIS)

    Shimazu, Yoichiro

    1976-01-01

    An evaluation is made of the fuel cycle costs for molten-salt reactors (MSR's), developed at Oak Ridge National Laboratory. Eight combinations of conditions affecting fuel cycle costs are compared, covering 233 U-Th, 235 U-Th and 239 Pu-Th fuels, with and without on-site continuous fuel reprocessing. The resulting fuel cycle costs range from 0.61 to 1.18 mill/kWh. A discussion is also given on the practicability of these fuel cycles. The calculations indicate that somewhat lower fuel cycle costs can be expected from reactor operation in converter mode on 235 U make-up with fuel reprocessed in batches every 10 years to avoid fission product precipitation, than from operation as 233 U-Th breeder with continuous reprocessing. (auth.)

  4. Advanced fuel cycles and burnup increase of WWER-440 fuel

    International Nuclear Information System (INIS)

    Proselkov, V.; Saprykin, V.; Scheglov, A.

    2003-01-01

    Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

  5. 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. 

  6. Analysis of environmental friendliness of DUPIC fuel cycle

    International Nuclear Information System (INIS)

    Ko, Won Il; Kim, Ho Dong

    2001-07-01

    Some properties of irradiated DUPIC fuels are compared with those of other fuel cycles. It was indicated that the toxicity of the DUPIC option based on 1 GWe-yr is much smaller than those of other fuel cycle options, and is just about half the order of magnitude of other fuel cycles. From the activity analysis of 99 Tc and 237 Np, which are important to the long-term transport of fission products stored in geologic media, the DUPIC option, was being contained only about half of those other options. It was found from the actinide content estimation that the MOX option has the lowest plutonium arising based on 1 GWe-year and followed by the DUPIC option. However, fissile Pu content generated in the DUPIC fuel was the lowest among the fuel cycle options. From the analysis of radiation barrier in proliferation resistance aspect, the fresh DUPIC fuel can play a radiation barrier part, better than CANDU spent fuels as well as fresh MOX fuel. It is indicated that the DUPIC fuel cycle has the excellent resistance to proliferation, compared with an existing reprocessing option and CANDU once-through option. In conclusions, DUPIC fuel cycle would have good properties on environmental effect and proliferation resistance, compared to other fuel cycle cases

  7. Overview of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Knief, R.A.

    1978-01-01

    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

  8. IFR fuel cycle--pyroprocess development

    International Nuclear Information System (INIS)

    Laidler, J.J.; Miller, W.E.; Johnson, T.R.; Ackerman, J.P.; Battles, J.E.

    1992-01-01

    The Integral Fast Reactor (IFR) fuel cycle is based on the use of a metallic fuel alloy, with nominal composition U-2OPu-lOZr. In its present state of development, this fuel system offers excellent high-burnup capabilities. Test fuel has been carried to burnups in excess of 20 atom % in EBR-II irradiations, and to peak burnups over 15 atom % in FFTF. The metallic fuel possesses physical characteristics, in particular very high thermal conductivity, that facilitate a high degree of passive inherent safety in the IFR design. The fuel has been shown to provide very large margins to failure in overpower transient events. Rapid overpower transient tests carried out in the TREAT reactor have shown the capability to withstand up to 400% overpower conditions before failing. An operational transient test conducted in EBR-II at a power ramp rate of 0.1% per second reached its termination point of 130% of normal power without any fuel failures. The IFR metallic fuel also exhibits superior compatibility with the liquid sodium coolant. Equally as important as the performance advantages offered by the use of metallic fuel is the fact that this fuel system permits the use of an innovative reprocessing method, known as ''pyroprocessing,'' featuring fused-salt electrorefining of the spent fuel. Development of the IFR pyroprocess has been underway at the Argonne National Laboratory for over five years, and great progress has been made toward establishing a commercially-viable process. Pyroprocessing offers a simple, compact means for closure of the fuel cycle, with anticipated significant savings in fuel cycle costs

  9. High conversion HTRs and their fuel cycle

    International Nuclear Information System (INIS)

    Gutmann, H.; Hansen, U.; Larsen, H.; Price, M.S.T.

    1975-01-01

    This report discusses the principles of the core design and the fuel cycle layout for High Conversion HTRs (HCHTRs). Though most of the principles apply equally to HTRs of the pebble-bed and the prismatic fuel element design types, the paper concentrates on the latter. Design and fuel cycle strategies for the full utilisation of the high conversion potential are compared with others that aim at easier reprocessing and the 'environmental' fuel cycle. The paper concludes by discussing operating and fuel cycle characteristics and economics of HCHTRs, and how the latter impinge on the allowable price for uranium ore and the available uranium resources. (orig./UA) [de

  10. Romanian nuclear fuel cycle development

    International Nuclear Information System (INIS)

    Rapeanu, S.N.; Comsa, Olivia

    1998-01-01

    Romanian decision to introduce nuclear power was based on the evaluation of electricity demand and supply as well as a domestic resources assessment. The option was the introduction of CANDU-PHWR through a license agreement with AECL Canada. The major factors in this choice have been the need of diversifying the energy resources, the improvement the national industry and the independence of foreign suppliers. Romanian Nuclear Power Program envisaged a large national participation in Cernavoda NPP completion, in the development of nuclear fuel cycle facilities and horizontal industry, in R and D and human resources. As consequence, important support was being given to development of industries involved in Nuclear Fuel Cycle and manufacturing of equipment and nuclear materials based on technology transfer, implementation of advanced design execution standards, QA procedures and current nuclear safety requirements at international level. Unit 1 of the first Romanian nuclear power plant, Cernavoda NPP with a final profile 5x700 Mw e, is now in operation and its production represents 10% of all national electricity production. There were also developed all stages of FRONT END of Nuclear Fuel Cycle as well as programs for spent fuel and waste management. Industrial facilities for uranian production, U 3 O 8 concentrate, UO 2 powder and CANDU fuel bundles, as well as heavy water plant, supply the required fuel and heavy water for Cernavoda NPP. The paper presents the Romanian activities in Nuclear Fuel Cycle and waste management fields. (authors)

  11. HTR Development in China

    International Nuclear Information System (INIS)

    Wang Dazhong

    2014-01-01

    The roles of HTRs in China: 1. Due to the inherent safety features, high efficiency of electricity generation, site flexibility, the modular HTR can act as a supplement to LWR for small and medium size power generation. 2. Co-generation to supply steam up to 600℃, for petroleum refinery, oil sand and oil shale processing, sea water desalination and district heating, etc. 3. Hydrogen production at 900~1000 ℃ by V/HTR. Conclusions and prospects: • China’s energy system will experience transition and reform in the future; • Nuclear energy will play an irreplaceable role in China’s energy development; • Due to the excellent features of inherent safety, the HTR is a promising technology for electricity generation and process heat utilization; • Further international cooperation and exchanges need to be enhanced

  12. Economic Analysis of Several Nuclear Fuel Cycles

    International Nuclear Information System (INIS)

    Ko, Won Il; Gao, Fanxing; Kim, Sung Ki

    2012-01-01

    Economics is one of the essential criteria to be considered for the future deployment of the nuclear power. With regard to the competitive power market, the cost of electricity from nuclear power plants is somewhat highly competitive with those from the other electricity generations, averaging lower in cost than fossil fuels, wind, or solar. However, a closer look at the nuclear power production brings an insight that the cost varies within a wide range, highly depending on a nuclear fuel cycle option. The option of nuclear fuel cycle is a key determinant in the economics, and therefrom, a comprehensive comparison among the proposed fuel cycle options necessitates an economic analysis for thirteen promising options based on the material flow analysis obtained by an equilibrium model as specified in the first article (Modeling and System Analysis of Different Fuel Cycle Options for Nuclear Power Sustainability (I): Uranium Consumption and Waste Generation). The objective of the article is to provide a systematic cost comparison among these nuclear fuel cycles. The generation cost (GC) generally consists of a capital cost, an operation and maintenance cost (O and M cost), a fuel cycle cost (FCC), and a decontaminating and decommissioning (D and D) cost. FCC includes a frontend cost and a back-end cost, as well as costs associated with fuel recycling in the cases of semi-closed and closed cycle options. As a part of GC, the economic analysis on FCC mainly focuses on the cost differences among fuel cycle options considered and therefore efficiently avoids the large uncertainties of the Generation-IV reactor capital costs and the advanced reprocessing costs. However, the GC provides a more comprehensive result covering all the associated costs, and therefrom, both GC and FCC have been analyzed, respectively. As a widely applied tool, the levelized cost (mills/KWh) proves to be a fundamental calculation principle in the energy and power industry, which is particularly

  13. Comparison of HTGR fuel design, manufacture and quality control methods between Japan and China

    International Nuclear Information System (INIS)

    Fu Xioming; Takahashi, Masashi; Ueta, Shouhei; Sawa, Kazuhiro

    2002-05-01

    The first-loading fuel for the HTTR was started to fabricate at Nuclear Fuel Industries (NFI) in 1995 and the HTTR reached criticality in 1998. Meanwhile, 10 MW high temperature reactor (HTR-10) was constructed in Institute of Nuclear Energy Technology (INET) of Tsinghua University, and the first-loading fuel was fabricated concurrently. The HTR-10 reached criticality in December 2000. Though fuel type is different, i.e., pin-in-block type for the HTTR and pebble bed type for the HTR-10, the fabrication method of TRISO coated fuel particles is similar to each other. This report describes comparison of fuel design, fabrication process and quality inspection between them. (author)

  14. The challenge of introducing HTR plants on to the international power plant market

    International Nuclear Information System (INIS)

    Bogen, J.; Stoelzl, D.

    1987-01-01

    The international power plant market today is characterized by high increase in energy consumption for developing countries with limitations of investment capital and low increase in energy consumption for industrialized countries with limitations of additional power plant capacities. As a consequence there is a low demand for large new power stations. This leads to a tendency for small and medium sized power plant units - meeting high environmental standards - for which the total investment volume is low and full load operation of a plant can be realized earlier due to the small block capacity. - For nuclear power plants the High-Temperature-Reactor (HTR)-line with spherical fuel elements and a core structure of graphite is specially suited for this small and medium sized nuclear reactor (SMSNR) capacity. The excellent safety characteristics, the high availability, the low radiation doses for the operation personnel and the environment of the HTR line has been demonstrated by 20 years of operation of the AVR-15 MWe experimental power plant in Juelich F.R.G. and since 1985 by operation of the THTR-300 MWe prototype plant at Hamm-Uentrop F.R.G. Up-dated concepts of the HTR-line are under design for electricity generation (HTR-500), for co-generation of power and heat (HTR-100) and for district heating purposes only (GHR-10). By implementing two HTR projects the Brown Boveri Group is in the position to realize the collected experiences from design, licensing, erection, commissioning and operation for the follow-on projects. This leads to practical and sound technical solutions convenient for existing manufacturing processes, well known materials, standardized components and usual manufacturing tolerances. Specific plant characteristics can be used for advantages in the competition. (author)

  15. The nuclear fuel cycle, an overview

    International Nuclear Information System (INIS)

    Ballery, J.L.; Cazalet, J.; Hagemann, R.

    1995-01-01

    Because uranium is widely distributed on the face of the Earth, nuclear energy has a very large potential as an energy source in view of future depletion of fossil fuel reserves. Also future energy requirements will be very sizeable as populations of developing countries are often growing and make the energy question one of the major challenges for the coming decades. Today, nuclear contributes some 340 GWe to the energy requirements of the world. Present and future nuclear programs require an adequate fuel cycle industry, from mining, refining, conversion, enrichment, fuel fabrication, fuel reprocessing and the storage of the resulting wastes. The commercial fuel cycle activities amount to an annual business in the 7-8 billions of US Dollars in the hands of a large number of industrial operators. This paper gives details about companies and countries involved in each step of the fuel cycle and about the national strategies and options chosen regarding the back end of the fuel cycle (waste storage and reprocessing). These options are illustrated by considering the policy adopted in three countries (France, United Kingdom, Japan) versed in reprocessing. (J.S.). 13 figs., 2 tabs

  16. Physics characteristics of CANDU cores with advanced fuel cycles

    International Nuclear Information System (INIS)

    Garvey, P.M.

    1985-01-01

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

  17. Compound process fuel cycle concept

    International Nuclear Information System (INIS)

    Ikegami, Tetsuo

    2005-01-01

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

  18. Fuel cycle centers revisited: Consolidation of fuel cycle activities in a few countries

    International Nuclear Information System (INIS)

    Kratzer, M.B.

    1996-01-01

    Despite varied expressions, the general impression remains that the international fuel cycle center concept, whatever its merits, is visionary. It also is quite possibly unattainable in light of strong national pressures toward independence and self-sufficiency in all things nuclear. Is the fuel cycle center an idea that has come and gone? Is it an idea whose time has not yet come? Or is it, as this paper suggests, an idea that has already arrived on the scene, attracting little attention or even acknowledgement of its presence? The difficult in answering this questions arises, in part, from the fact that despite its long and obvious appeal, there has been very little systematic analysis of the concept itself. Such obvious questions as how many and where fuel cycle centers should be located; what characteristics should the hot country or countries possess; and what are the institutional forms or features that endow the concept with enhanced proliferation protection have rarely been seriously and systematically addressed. The title of this paper focuses on limiting the geographic spread of fuel cycle facilities, and some may suggest that doing so does not necessarily call for any type of international or multinational arrangements applicable to those that exist. It is a premise of this paper, however, that a restriction on the number of countries possessing sensitive fuel cycle facilities necessarily involves some degree of multinationalization. This is not only because in every instance a nonproliferation pledge and international or multinational safeguards, or both, will be applied to the facility, but also because a restriction on the number of countries possessing these facilities implies that those in existence will serve a multinational market. This feature in itself is an important form of international auspices. Thus, the two concepts--limitation and multinationalization--if not necessarily one and the same, are at least de facto corollaries

  19. Rapsodie: A closed fuel cycle

    International Nuclear Information System (INIS)

    Levallet, E.H.; Costa, L.; Mougniot, J.C.; Robin, J.

    1977-01-01

    The Fortissimo Version of the core of the RAPSODIE fast reactor produces 40 MWTh. Since its start up in May 1970 in the CEN-CADARACHE its availability has stayed around 85%. Some of the mixed oxyde fuel pins UO 2 - 30% PuO 2 have already reached 150.000 MWd/t. The reprocessing is done in the pilot plant located in the La Hague Center and the plutonium obtained has already been re-used in the reactor. The Rapsodie-Fortissimo cycle is therefore now a closed cycle. This cycle is quite representative of fast reactor cycle characteristics and thus provides a remarkable research and development tool for the study of fabrication, in-reactor performances, transport, storage and reprocessing. These studies concern in particular the evolution of fission products and heavy isotopes content in fuel which controls both reprocessing schemes and intensity of emitted radiations. A program for the analysis of irradiated fuel has been developed either using samples collected all along the cycle, or following the actual reprocessing subassemblies. A set of basic data and calculation models has been established with two objectives: to give a better interpretation of the experimental program on one hand, and to extrapolate these results to the fuel cycle of fast reactors in general on the other hand. The first results have been quite encouraging up to now [fr

  20. Back-end of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Choi, J.S.

    2002-01-01

    Current strategies of the back-end nuclear fuel cycles are: (1) direct-disposal of spent fuel (Open Cycle), and (2) reprocessing of the spent fuel and recycling of the recovered nuclear materials (Closed Cycle). The selection of these strategies is country-specific, and factors affecting selection of strategy are identified and discussed in this paper. (author)

  1. International nuclear fuel cycle fact book. Revision 6

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  2. A Small Modular Reactor Design for Multiple Energy Applications: HTR50S

    Energy Technology Data Exchange (ETDEWEB)

    Yan, X.; Tachibana, Y.; Ohashi, H.; Sato, H.; Tazawa, Y.; Kunitomi, K. [Japan Atomic Energy Agency, Ibaraki (Japan)

    2013-06-15

    HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's 950 .deg. C, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to 750 .deg. C for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to 900 .deg. C for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

  3. Introductory remarks about the international fuel cycle

    International Nuclear Information System (INIS)

    Wolfe, B.

    1989-01-01

    The reason why nuclear power has promise is because of the promise of its fuel cycle. The fuel cycle is in fairly good shape and has demonstrated the characteristics of good economics, good general characterization, and good maintenance of the various parts of the fuel cycle. The thermal recycling of fuel is an area in which the economics have changed to the point that, at least in many parts of the world, it's no longer economical

  4. A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

    Science.gov (United States)

    Syarip; Po, L. C. C.

    2018-05-01

    In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

  5. IAEA Activities in the Area of Fast Reactors and Related Fuels and Fuel Cycles

    International Nuclear Information System (INIS)

    Monti, S.; Basak, U.; Dyck, G.; Inozemtsev, V.; Toti, A.; Zeman, A.

    2013-01-01

    Summary: • The IAEA role to support fast reactors and associated fuel cycle development programmes; • Main IAEA activities on fast reactors and related fuel and fuel cycle technology; • Main IAEA deliverables on fast reactors and related fuel and fuel cycle technology

  6. Regulation of fuel cycle facilities in the UK

    International Nuclear Information System (INIS)

    Ascroft-Hutton, H.W.

    2001-01-01

    The UK has facilities for the production of uranium hexafluoride, its enrichment, conversion into fuel and for the subsequent reprocessing of irradiated fuel and closure of the fuel cycle. All of these facilities must be licensed under UK legislation. HM Nuclear Installations Inspectorate has delegated powers to issue the licence and to attach any conditions it considers necessary in the interests of safety. The fuel cycle facilities in the UK have been licensed since 1971. This paper describes briefly the UK nuclear regulatory framework and the fuel cycle facilities involved. It considers the regulatory practices adopted together with similarities and differences between regulation of fuel cycle facilities and power reactors. The safety issues associated with the fuel cycle are discussed and NII's regulatory strategy for these facilities is set out. (author)

  7. International nuclear fuel cycle fact book. Revision 6

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-01-01

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

  8. Social awareness on nuclear fuel cycle

    International Nuclear Information System (INIS)

    Tanigaki, Toshihiko

    2006-01-01

    In the present we surveyed public opinion regarding the nuclear fuel cycle to find out about the social awareness about nuclear fuel cycle and nuclear facilities. The study revealed that people's image of nuclear power is more familiar than the image of the nuclear fuel cycle. People tend to display more recognition and concern towards nuclear power and reprocessing plants than towards other facilities. Comparatively speaking, they tend to perceive radioactive waste disposal facilities and nuclear power plants as being highly more dangerous than reprocessing plants. It is found also that with the exception of nuclear power plants don't know very much whether nuclear fuel cycle facilities are in operation in Japan or not. The results suggests that 1) the relatively mild image of the nuclear fuel cycle is the result of the interactive effect of the highly dangerous image of nuclear power plants and the less dangerous image of reprocessing plants; and 2) that the image of a given plant (nuclear power plant, reprocessing plant, radioactive waste disposal facility) is influenced by the fact of whether the name of the plant suggests the presence of danger or not. (author)

  9. Reference thorium fuel cycle

    International Nuclear Information System (INIS)

    Driggers, F.E.

    1978-08-01

    In the reference fuel cycle for the TFCT program, fissile U will be denatured by mixing with 238 U; the plants will be located in secure areas, with Pu being recycled within these secure areas; Th will be recycled with recovered U and Pu; the head end will handle a variety of core and blanket fuel assembly designs for LWRs and HWRs; the fuel may be a homogeneous mixture either of U and Th oxide pellets or sol-gel microspheres; the cladding will be Zircaloy; and MgO may be added to the fuel to improve Th dissolution. Th is being considered as the fertile component of fuel in order to increase proliferation resistance. Spent U recovered from Th-based fuels must be re-enriched before recycle to prevent very rapid buildup of 238 U. Stainless steel will be considered as a backup to Zircaloy cladding in case Zr is incompatible with commercial aqueous dissolution. Storage of recovered irradiated Th will be considered as a backup to its use in the recycle of recovered Pu and U. Estimates are made of the time for introducing the Th fuel cycle into the LWR power industry. Since U fuel exposures in LWRs are likely to increase from 30,000 to 50,000 MWD/MT, the Th reprocessing plant should also be designed for Th fuel with 50,000 MWD/MT exposure

  10. Aspects of the fast reactors fuel cycle

    International Nuclear Information System (INIS)

    Zouain, D.M.

    1982-06-01

    The fuel cycle for fast reactors, is analysed, regarding the technical aspects of the developing of the reprocessing stages and the fuel fabrication. The environmental impact of LMFBRs and the waste management of this cycle are studied. The economic aspects of the fuel cycle, are studied too. Some coments about the Brazilian fast reactors programs are done. (E.G.) [pt

  11. Survey of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Zech, H.J.; Pickert, F.K.

    1975-01-01

    A brief outline of the technical aspects of the fuel cycle, starting from the mining of uranium up to fuel element fabrication, is followed by a more detailed description of the management of the outer fuel cycle. This includes the system of contracts and their reciprocal technical and chronological interdepence, as well as financial aspects, market conditions and trends. (RB) [de

  12. Ecological effects of fuel cycle activities

    Energy Technology Data Exchange (ETDEWEB)

    Barnthouse, L; Cada, G; Kroodsma, R; Shriner, D; Tolbert, V; Turner, R

    1994-07-01

    The purpose of this paper is to summarize the approach used to characterize ecological impacts of the coal fuel cycle. The same approach is used for many of the impacts in other fuel cycles as well. The principal analytical approach being used in the study is an accounting framework - that is, a series of matrices that map each phase of the fuel cycle to a suite of possible. emissions, each emission to a suite of impact categories, and each impact category to an external cost. This paper summarizes the ecological impacts of all phases of the coal fuel cycle, defines the ecological impact categories used in the study's 'accounting framework', and discusses alternative approaches to quantification. Externalities associated with CO{sub 2}-induced global climate change are beyond the scope of this paper and are not discussed.

  13. Ecological effects of fuel cycle activities

    International Nuclear Information System (INIS)

    Barnthouse, L.; Cada, G.; Kroodsma, R.; Shriner, D.; Tolbert, V.; Turner, R.

    1994-01-01

    The purpose of this paper is to summarize the approach used to characterize ecological impacts of the coal fuel cycle. The same approach is used for many of the impacts in other fuel cycles as well. The principal analytical approach being used in the study is an accounting framework - that is, a series of matrices that map each phase of the fuel cycle to a suite of possible. emissions, each emission to a suite of impact categories, and each impact category to an external cost. This paper summarizes the ecological impacts of all phases of the coal fuel cycle, defines the ecological impact categories used in the study's 'accounting framework', and discusses alternative approaches to quantification. Externalities associated with CO 2 -induced global climate change are beyond the scope of this paper and are not discussed

  14. Nuclear Fusion Fuel Cycle Research Perspectives

    International Nuclear Information System (INIS)

    Chung, Hongsuk; Koo, Daeseo; Park, Jongcheol; Kim, Yeanjin; Yun, Sei-Hun

    2015-01-01

    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

  15. Serving the fuel cycle: preparing tomorrow's packagings

    International Nuclear Information System (INIS)

    Roland, V.

    2001-01-01

    The main fleet of transport packagings serving today the fuel cycle was born more than 20 years ago. Or was it they? The present paper will show that serving the fuel cycle by preparing tomorrow's logistics is actually an on-going process, rather than a rupture. We shall review the great packagings of the fuel cycle: In the front end, the major actors are the UF 4 , UF 6 , enriched UF 6 , UO 2 powders, fresh fuel packagings. In the back end of the fuel cycle, we find the dry transport casks of the TN-12, TN-17, TN-13, family and also the Excellox wet flasks. In the waste management, a whole fleet of containers, culminating in the TN Gemini, are available or being created. (author)

  16. Nuclear fuel cycle modelling using MESSAGE

    International Nuclear Information System (INIS)

    Guiying Zhang; Dongsheng Niu; Guoliang Xu; Hui Zhang; Jue Li; Lei Cao; Zeqin Guo; Zhichao Wang; Yutong Qiu; Yanming Shi; Gaoliang Li

    2017-01-01

    In order to demonstrate the possibilities of application of MESSAGE tool for the modelling of a Nuclear Energy System at the national level, one of the possible open nuclear fuel cycle options based on thermal reactors has been modelled using MESSAGE. The steps of the front-end and back-end of nuclear fuel cycle and nuclear reactor operation are described. The optimal structure for Nuclear Power Development and optimal schedule for introducing various reactor technologies and fuel cycle options; infrastructure facilities, nuclear material flows and waste, investments and other costs are demonstrated. (author)

  17. Nuclear fuel cycle information workshop

    International Nuclear Information System (INIS)

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    Suripto, A.; Sastratenaya, A.S.; Sutarno, D.

    2000-01-01

    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)

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

    International Nuclear Information System (INIS)

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

    1977-03-01

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

  20. The nuclear fuel cycle versus the carbon cycle

    International Nuclear Information System (INIS)

    Ewing, R.C.

    2005-01-01

    Nuclear power provides approximately 17% of the world's electricity, which is equivalent to a reduction in carbon emissions of ∼0.5 gigatonnes (Gt) of C/yr. This is a modest reduction as compared with global emissions of carbon, ∼7 Gt C/yr. Most analyses suggest that in order to have a significant and timely impact on carbon emissions, carbon-free sources, such as nuclear power, would have to expand total production of energy by factors of three to ten by 2050. A three-fold increase in nuclear power capacity would result in a projected reduction in carbon emissions of 1 to 2 Gt C/yr, depending on the type of carbon-based energy source that is displaced. This three-fold increase utilizing present nuclear technologies would result in 25,000 metric tonnes (t) of spent nuclear fuel (SNF) per year, containing over 200 t of plutonium. This is compared to a present global inventory of approximately 280,000 t of SNF and >1,700 t of Pu. A nuclear weapon can be fashioned from as little as 5 kg of 239 Pu. However, there is considerable technological flexibility in the nuclear fuel cycle. There are three types of nuclear fuel cycles that might be utilized for the increased production of energy: open, closed, or a symbiotic combination of different types of reactor (such as, thermal and fast neutron reactors). The neutron energy spectrum has a significant effect on the fission product yield, and the consumption of long-lived actinides, by fission, is best achieved by fast neutrons. Within each cycle, the volume and composition of the high-level nuclear waste and fissile material depend on the type of nuclear fuel, the amount of burn-up, the extent of radionuclide separation during reprocessing, and the types of materials used to immobilize different radionuclides. As an example, a 232 Th-based fuel cycle can be used to breed fissile 233 U with minimum production of Pu. In this paper, I will contrast the production of excess carbon in the form of CO 2 from fossil fuels with

  1. CANDU fuel cycles - present and future

    International Nuclear Information System (INIS)

    Mooradian, A.J.

    1976-05-01

    The present commercially proven Canadian nuclear power system is based on a once-through natural uranium fuel cycle characterized by high uranium utilization and a high conversion efficiency. The cycle closes with secure retrievable storage of spent fuel. This cycle is based on a CANDU reactor concept which is now well understood. Both active and passive fuel storage options have been investigated and will be described in this paper. Future development of the CANDU system is focussed on conservation of uranium by plutonium and thorium recycle. The full exploitation of these options requires continued emphasis on neutron conservation, efficiency of extraction and fuel refabrication processes. The results of recent studies are discussed in this paper. (author)

  2. CFTSIM-ITER dynamic fuel cycle model

    International Nuclear Information System (INIS)

    Busigin, A.; Gierszewski, P.

    1998-01-01

    Dynamic system models have been developed for specific tritium systems with considerable detail and for integrated fuel cycles with lesser detail (e.g. D. Holland, B. Merrill, Analysis of tritium migration and deposition in fusion reactor systems, Proceedings of the Ninth Symposium Eng. Problems of Fusion Research (1981); M.A. Abdou, E. Vold, C. Gung, M. Youssef, K. Shin, DT fuel self-sufficiency in fusion reactors, Fusion Technol. (1986); G. Spannagel, P. Gierszewski, Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket, Fusion Eng. Des. (1991); W. Kuan, M.A. Abdou, R.S. Willms, Dynamic simulation of a proposed ITER tritium processing system, Fusion Technol. (1995)). In order to provide a tool to understand and optimize the behavior of the ITER fuel cycle, a dynamic fuel cycle model called CFTSIM is under development. The CFTSIM code incorporates more detailed ITER models, specifically for the important isotope separation system, and also has an easier-to-use graphical interface. This paper provides an overview of CFTSIM Version 1.0. The models included are those with significant and varying tritium inventories over a test campaign: fueling, plasma and first wall, pumping, fuel cleanup, isotope separation and storage. An illustration of the results is shown. (orig.)

  3. Alternative fuel cycles

    International Nuclear Information System (INIS)

    Penn, W.J.

    1979-05-01

    Uranium resource utilization and economic considerations provide incentives to study alternative fuel cycles as future options to the PHWR natural uranium cycle. Preliminary studies to define the most favourable alternatives and their possible introduction dates are discussed. The important and uncertain components which influence option selection are reviewed, including nuclear capacity growth, uranium availability and demand, economic potential, and required technological developments. Finally, a summary of Ontario Hydro's program to further assess cycle selection and define development needs is given. (auth)

  4. Fuel cycle problems in fusion reactors

    International Nuclear Information System (INIS)

    Hickman, R.G.

    1976-01-01

    Fuel cycle problems of fusion reactors evolve around the breeding, recovery, containment, and recycling of tritium. These processes are described, and their implications and alternatives are discussed. Technically, fuel cycle problems are solvable; economically, their feasibility is not yet known

  5. Status and development of the thorium fuel cycle

    International Nuclear Information System (INIS)

    Yi Weijing; Wei Renjie

    2003-01-01

    A perspective view of the thorium fuel cycle is provided in this paper. The advantages and disadvantages of the thorium fuel cycle are given and the development of thorium fuel cycle in several types of reactors is introduced. The main difficulties in developing the thorium fuel cycle lie in the reprocessing and disposal of the waste and its economy, and the ways tried by foreign countries to solve the problems are presented in the paper

  6. Procedures and results of the probabilistic safety study of the HTR-1160 plant

    International Nuclear Information System (INIS)

    Kroeger, W.; Bongartz, R.

    1981-01-01

    A research team of the Institute for Nuclear Safety Research of the Juelich Nuclear Research Center (KFA) and staff members of the Gesellschaft fuer Reaktorsicherheit, sponsored by the Federal Ministry of the Interior, carried out a safety and risk analysis of high temperature reactors. The studies, which included the transfer to German conditions and the modification in some points of methodology of the American AIPA Study, were performed on the German concept of an 1160 MWe HTR with block-type fuel elements. They referred to accidents and possible impacts on the environment, residual risks and measures to reduce them. The study covered a total of approx. 15 groups of initiating events, including potential external impacts. The dominating initiating events are transients in a pressurized reactor. Differences relative to the light water reactor concept arise with respect to releases as a result of accidents and, above all, release times; they are due to different physical characteristics. HTR'S are characterized by thermal inertia and resistance to temperatures. If the results of the study are extended to the HTR line with a pebble bed core currently in the planning phase, the power densities alone, which are considerably lower in some designs, are indicative of an even more effective fission product retention than is already found in the HTR-1160 plant analyzed here. (orig.) [de

  7. Fuel cycle technologies - The next 50 years

    International Nuclear Information System (INIS)

    Chamberlain, L.N.; Ion, S.E.; Patterson, J.

    1997-01-01

    World energy demands are set to increase through the next Millennium. As fossil fuel reserves fall and environmental concerns increase there is likely to be a growing dependence on nuclear and renewable sources for electricity generation. This paper considers some of the desirable attributes of the nuclear fuel cycle in the year 2050 and emphasises the importance of considering the whole of the fuel cycle in an integrated way - the concept of the 'holistic' fuel cycle. We then consider how some sectors of the fuel cycle will develop, through a number of multi- national contributions covering: enrichment, fuel, aqueous reprocessing, non-aqueous reprocessing, P and T, MOX, direct disposal, waste. Finally, we summarize some of the key technical and institutional challenges that lie ahead if nuclear power is going to play its part in ensuring that planet Earth is a safe and hospitable place to live. (author)

  8. The R&D of HTGR high temperature helium sampling loop: From HTR-10 to HTR-PM

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Chao, E-mail: fangchao@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China); Bao, Xuyin; Yang, Chen; Yang, Yanran; Cao, Jianzhu [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China)

    2016-09-15

    A High Temperature Helium Sampling Loop (HTHSL) for studying the transportation (deposition) behavior and total amount of solid fission products in high-temperature helium coming from the steam generator (SG) in the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) and High Temperature Reactor-Pebble bed Modules (HTR-PM) are researched and designed, respectively. Through the optimal design and simulation based on thermohydraulics analysis, the three-sleeve structure of deposition sampling device (DSD) could realize full-length temperature control evenly so that it could be used to study fission products in the primary circuit of HTR-10. On the other hand, an improved DSD is also designed for HTR-PM based on corresponding simulations, which could be used to sample the important nuclei in the high temperature helium from SG. These schemes offer two different methods to obtain the original source term in the high temperature helium, which will provide deeper understanding for the analysis of source terms of HTGR.

  9. Modeling the Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Jacobson, Jacob J.; Dunzik-Gougar, Mary Lou; Juchau, Christopher A.

    2010-01-01

    A review of existing nuclear fuel cycle systems analysis codes was performed to determine if any existing codes meet technical and functional requirements defined for a U.S. national program supporting the global and domestic assessment, development and deployment of nuclear energy systems. The program would be implemented using an interconnected architecture of different codes ranging from the fuel cycle analysis code, which is the subject of the review, to fundamental physical and mechanistic codes. Four main functions are defined for the code: (1) the ability to characterize and deploy individual fuel cycle facilities and reactors in a simulation, while discretely tracking material movements, (2) the capability to perform an uncertainty analysis for each element of the fuel cycle and an aggregate uncertainty analysis, (3) the inclusion of an optimization engine able to optimize simultaneously across multiple objective functions, and (4) open and accessible code software and documentation to aid in collaboration between multiple entities and facilitate software updates. Existing codes, categorized as annualized or discrete fuel tracking codes, were assessed according to the four functions and associated requirements. These codes were developed by various government, education and industrial entities to fulfill particular needs. In some cases, decisions were made during code development to limit the level of detail included in a code to ease its use or to focus on certain aspects of a fuel cycle to address specific questions. The review revealed that while no two of the codes are identical, they all perform many of the same basic functions. No code was able to perform defined function 2 or several requirements of functions 1 and 3. Based on this review, it was concluded that the functions and requirements will be met only with development of a new code, referred to as GENIUS.

  10. Back end of the fuel cycle

    International Nuclear Information System (INIS)

    Wolfe, B.; Lambert, R.W.

    1975-01-01

    At present, that portion of the nuclear fuel cycle involving reprocessing, waste management, and mixed-oxide fuel fabrication is in an unsettled state. Government regulatory requirements with respect to all aspects of the back end of the fuel cycle are still being formulated, and there is little positive experience on the operation of commercial reprocessing or mixed-oxide fabrication plants. In view of this unsettled situation, it will be difficult to meet the reprocessing and mixed-oxide fabrication needs of the next decade in the pattern previously anticipated. The costs in the back end of the fuel cycle are much higher than had been anticipated several years ago, a situation similar to that of almost all large endeavors in this country. On the other hand, the added costs are small relative to total power costs and do not affect the economic advantage of nuclear power as compared to other power sources. A rough economic analysis indicates that the question for the back end of the fuel cycle has changed from one of optimizing profitability to one of determining the most economic disposition of spent fuel. Long-term spent fuel storage is a practical and economically acceptable way to provide time for determining a sound course of action for the back end of the fuel cycle. Indeed, if one could count on a breeder economy before the end of the century, one possible course of action is to store light-water fuel until the plutonium can be used in breeders. However, for philosophical as well as practical reasons, it is important that the uncertainties in the course of action should be resolved as quickly as possible. Long-term storage should not be an excuse to delay resolution of the basic questions. (U.S.)

  11. Nuclear fuel cycle scenarios at CGNPC

    International Nuclear Information System (INIS)

    Xiao, Min; Zhou, Zhou; Nie, Li Hong; Mao, Guo Ping; Hao, Si Xiong; Shen, Kang

    2008-01-01

    Established in 1994, China Guangdong Nuclear Power Holding Co. (CGNPC) now owns two power stations GNPS and LNPS Phase I, with approximate 4000 MWe of installed capacity. With plant upgrades, advanced fuel management has been introduced into the two plants to improve the plant economical behavior with the high burnup fuel implemented. For the purpose of sustainable development, some preliminary studies on nuclear fuel cycle, especially on the back-end, have been carried out at CGNPC. According to the nuclear power development plan of China, the timing for operation and the capacity of the reprocessing facility are studied based on the amount of the spent fuel forecast in the future. Furthermore, scenarios of the fuel cycles in the future in China with the next generation of nuclear power were considered. Based on the international experiences on the spent fuel management, several options of spent fuel reprocessing strategies are investigated in detail, for example, MOX fuel recycling in light water reactor, especially in the current reactors of CGNPC, spent fuel intermediated storage, etc. All the investigations help us to draw an overall scheme of the nuclear fuel cycle, and to find a suitable road-map to achieve the sustainable development of nuclear power. (authors)

  12. Regional nuclear fuel cycle centers study project

    International Nuclear Information System (INIS)

    Bennett, L.; Catlin, R.G.; Meckoni, V.

    1977-01-01

    The concept of regional fuel cycle centers (RFCC) has attracted wide interest. The concept was endorsed by many countries in discussions at the General Conference of the International Atomic Energy Agency and at the General Assembly of the United Nations. Accordingly, in 1975, the IAEA initiated a detailed study of the RFCC concept. The Agency study has concentrated on what is referred to as the ''back-end'' of the fuel cycle because that is the portion which is currently problematic. The study covers transport, storage, processing and recycle activities starting from the time the spent fuel leaves the reactor storage pools and through all steps until the recycled fuel is in finished fuel elements and shipped to the reactor. A detailed evaluation of the specific features of large regional fuel cycle centers established on a multinational basis vis-a-vis smaller dispersed fuel cycle facilities set up on a national basis has been carried out. The methodology for assessment of alternative strategies for fuel storage, reprocessing, and recycling of plutonium has been developed, characteristic data on material flows and cost factors have been generated, and an analytic system has been developed to carry out such evaluations including appropriate sensitivity analysis. Studies in related areas on institutional and legal, organizational, environmental, materials control and other essential aspects have also been made. The material developed during the course of this Study would enable any group of interested Member States to examine and work out alternative strategies pertinent to their present and projected nuclear fuel cycle needs, as well as evolve institutional, legal and other appropriate frameworks or agreements for the establishment of fuel cycle centers on a multinational cooperative basis

  13. Progress and problems in modelling HTR core dynamics

    International Nuclear Information System (INIS)

    Scherer, W.; Gerwin, H.

    1991-01-01

    In recent years greater effort has been made to establish theoretical models for HTR core dynamics. At KFA Juelich the TINTE (TIme dependent Neutronics and TEmperatures) code system has been developed, which is able to model the primary circuit of an HTR plant using modern numerical techniques and taking into account the mutual interference of the relevant physical variables. The HTR core is treated in 2-D R-Z geometry for both nucleonics and thermo-fluid-dynamics. 2-energy-group diffusion theory is used in the nuclear part including 6 groups of delayed neutron precursors and 14 groups of decay heat producers. Local and non-local heat sources are incorporated, thus simulating gamma ray transport. The thermo-fluid-dynamics module accounts for heterogeneity effects due to the pebble bed structure. Pipes and other components of the primary loop are modelled in 1-D geometry. Forced convection may be treated as well as natural convection in case of blower breakdown accidents. Validation of TINTE has started using the results of a comprehensive experimental program that has been performed at the Arbeitsgemeinschaft Versuchsreaktor GmbH (AVR) high temperature pebble bed reactor at Juelich. In the frame of this program power transients were initiated by varying the helium blower rotational speed or by moving the control rods. In most cases a good accordance between experiment and calculation was found. Problems in modelling the special AVR reactor geometry in two dimensions are described and suggestions for overcoming the uncertainties of experimentally determined control rod reactivities are given. The influence of different polynomial expansions of xenon cross sections to long term transients is discussed together with effects of burnup during that time. Up to now the TINTE code has proven its general applicability to operational core transients of HTR. The effects of water ingress on reactivity, fuel element corrosion and cooling gas properties are now being

  14. 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.

  15. Why HTR/VHTR? A European point of view

    International Nuclear Information System (INIS)

    Basini, V.; Bogusch, E.; Breuil, E.; Buckthorpe, D.; Chauvet, V.; Ftitterer, M.; Van Heek, A.; Hittner, D.; Von Lensa, W.; Pirson, J.; Verrier, D.

    2008-01-01

    The (European) High Temperature Reactor Technology Network (HTR-TN) was created in 2000 by the main industrial and Research actors of nuclear energy in Europe for elaborating a strategy for developing advanced HTR technology towards industrial application and for taking initiatives for implementing this strategy, most particularly through the Euratom funded R and D programmes. HTR-TN members are convinced that the main market push for industrial deployment of a new generation of HTR will not come from utility needs for electricity generation, but from industrial process heat needs: even if HTR can be considered for satisfying particular niches of the electricity market, there will not be any incentive for utilities already experienced in the exploitation of large LWR to take the risk of a significant technology change, when no evident competitive edge would result from it. On the contrary, HTR is the sole nuclear system that can address heat needs of a large number of industrial processes that require a higher temperature than the temperature provided by all other types of industrial reactors. The possibility for HTR to address the industrial process heat market is a strong asset, as it opens to HTR a large market which is presently looking for solutions to reduce drastically CO 2 emissions, but at the same time it is a huge challenge: industrial exploitation of nuclear energy has been for the time being focused on electricity generation for which user requirements are relatively uniform. The versatility of process heat needs in terms of power, temperature, reliability, etc. will require a much larger flexibility of the nuclear heat source, which is not usual for nuclear industry, looking for competitiveness through standardisation. Therefore HTR-TN considers that the top priority innovation for HTR present development should not be missed: it is to demonstrate at an industrial scale the technical, industrial and economical feasibility of the coupling of a HTR with

  16. Fuel cycle cost comparisons with oxide and silicide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

    1983-09-01

    This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. The status of the development and demonstration of the oxide and silicide fuels are presented in several papers in these proceedings. Routine utilization of these fuels with the uranium densities considered here requires that they are successfully demonstrated and licensed. Thermal-hydraulic safety margins, shutdown margins, mixed cores, and transient analyses are not addressed here, but analyses of these safety issues are in progress for a limited number of the most promising design options. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data is presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed. All safety criteria for the reactor with these fuel element designs need to be satisfied as well. With LEU oxide fuel, 31 g U/cm{sup 3} 1 and 0.76 mm--thick fuel meat, elements with 18-22 plates 320-391 g {sup 235}U) result in the same or lower total costs than with the HEU element 23 plates, 280 g {sup 235}U). Higher LEU loadings (more plates per element) are needed for larger excess reactivity requirements. However, there is little cost advantage to using more than 20 of these plates per element. Increasing the fuel meat thickness from 0.76 mm to 1.0 mm with 3.1 g U/cm{sup 3} in the design with 20 plates per element could result in significant cost reductions if the

  17. The IFR modern nuclear fuel cycle

    International Nuclear Information System (INIS)

    Hannum, W.H.

    1991-01-01

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

  18. Implications of alternative fuel cycles

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    The United States is re-examining alternative fuel cycles and nuclear power strategies, and doubtful attempts are being made to justify the economics of the 'throw-away' fuel cycle. At an international forum on 'An acceptable nuclear energy future for the world' at Fort Lauderdale, Karl Cohen of General Electric and a leading authority on this topic put the implications into perspective. Extracts from his address are presented

  19. The HTR safety concept demonstrated by selected examples

    International Nuclear Information System (INIS)

    Sommer, H.; Stoelzl, D.

    1981-01-01

    The licensing experience gained in the Federal Republic of Germany is based on the licensing procedures for the THTR-300 and the HTR-1160. In the course of the licensing procedures for these reactors a safety concept for an HTR has been developed. This experience constitutes the basis for the design of future HTR's. (author)

  20. The future of HTR development and market chances

    International Nuclear Information System (INIS)

    Baust, E.; Weisbrodt, I.

    1989-01-01

    In more than thirty years of development, the pebble bed high-temperature reactor has been brought to the threshold of commercial maturity. On the basis of the experience accumulated with the 15 MW AVR reactor and the THTR-300, unit sizes tailored to demand (HTR-500, modular HTR, GHR-10) will be developed for the electricity and heat markets of the future. The high-temperature reactor is a meaningful supplement to the proven line of light-water reactors and is particularly suitable for being exported to developing countries and industrial threshold countries because of its special technical and inherent safeguards properties. There is broad worldwide interest in the HTR, as is evidenced by several existing agreements on cooperation. It is for this reason that market chances are believed to exist for the HTR after the expected revival of the nuclear power market. ABB and Siemens therefore have decided to develop and market the HTR jointly in the future as a matter of long term strategy by working through a joint subsidiary, HTR-GmbH. (orig.) [de

  1. Research on application of burnable poison in pebble bed HTR

    International Nuclear Information System (INIS)

    Wei Chunlin; Zhang Jian; Shan Wenzhi; Jing Xingqing

    2013-01-01

    Burnable poison in fuel ball was used in pebble bed high-temperature gas-cooled reactor (HTR) to optimize the shape and the peak factor of power distribution in certain conditions. Two options are available and evaluated, that is the homogeneous burnable poison in graphite matrix and burnable poison particles (BPPs) in fuel balls. Due to the absorption cross section of "1"0B, the depletion speed for homogeneous burnable poison is very fast, and difficult to control, on the other side, the depletion speed of BPPs can be optimized respecting to its size, and better shape and peak value of power distribution can be achieved. (authors)

  2. Gadolinia experience and design for PWR fuel cycles

    International Nuclear Information System (INIS)

    Stephenson, L. C.

    2000-01-01

    The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

  3. Fuel cycles of WWER-1000 based on assemblies with increased fuel mass

    International Nuclear Information System (INIS)

    Kosourov, E.; Pavlovichev, A.; Shcherenko, A.

    2011-01-01

    Modern WWER-1000 fuel cycles are based on FAs with the fuel column height of 3680 mm, diameters of the fuel pellet and its central hole of 7.6 and 1.2 mm respectively. The highest possible fuel enrichment has reached its license limit that is 4.95 %. Research in the field of modernization, safety justification and licensing of equipment for fuel manufacture, storage and transportation are required for further fuel enrichment increase (above 5 %). So in the nearest future an improvement of technical and economic characteristics of fuel cycles is possible if assembly fuel mass is increased. The available technology of the cladding thinning makes it possible. If the fuel rod outer diameter is constant and the clad inner diameter is increased to 7.93 mm, the diameter of the fuel pellet can be increased to 7.8 mm. So the suppression of the pellet central hole allows increasing assembly fuel weight by about 8 %. In this paper we analyze how technical and economic characteristics of WWER-1000 fuel cycle change when an advanced FA is applied instead of standard one. Comparison is made between FAs with equal time interval between refueling. This method of comparison makes it possible to eliminate the parameters that constitute the operation component of electricity generation cost, taking into account only the following technical and economic characteristics: 1)cycle length; 2) average burnup of spent FAs; 3) specific natural uranium consumption; 4)specific quantity of separative work units; 5) specific enriched uranium consumption; 6) specific assembly consumption. Collected data allow estimating the efficiency of assembly fuel weight increase and verifying fuel cycle characteristics that may be obtained in the advanced FAs. (authors)

  4. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    1998-05-01

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

  5. Fast breeder fuel cycle, worldwide and French prospects

    International Nuclear Information System (INIS)

    Rapin, M.

    1982-01-01

    A review is given of fast breeder fuel cycle development from both the technological and the economical points of view. LMFBR fuel fabrication, reactor operation, spent fuel storage and transportation, reprocessing and fuel cycle economics are topics considered. (U.K.)

  6. A preliminary neutronic evaluation of the high temperature gas-cooled test reactor HTR-10 using the scale 6.0 code

    International Nuclear Information System (INIS)

    Sousa, Romulo V.; Fortini, Angela; Pereira, Claubia; Carvalho, Fernando R. de; Oliveira, Arno H.

    2013-01-01

    The High Temperature Gas-cooled Test Reactor HTR-10 is a 10 MW modular pebble bed type reactor, which core is filled with 27,000 spherical fuel elements, e.g. TRISO coated particles. This reactor was built by the Institute of Nuclear Energy Technology (INET), Tsinghua University, China, and its first criticality was attained on December 1, 2000. The main objectives of the HTR-10 are to verify and demonstrate the technical and safety features of the modular HTGR (High Temperature Gas-cooled Reactor) and to establish an experimental base for developing nuclear process heat applications. In this work, using the Standardized Computer Analysis for Licensing Evaluation (SCALE) 6.0, a nuclear code developed by Oak Ridge National Laboratory (ORNL), the HTR-10 first critical core is modeled by the DEN/UFMG. The K eff was obtained and compared with the reference value obtained by the Idaho National Laboratory. The result presents good agreement with experimental value. The goal is to validate the DEN/UFMG model to be applied in transmutation studies changing the fuel. (author)

  7. Fuel fabrication and reprocessing for nuclear fuel cycle with inherent safety demands

    Energy Technology Data Exchange (ETDEWEB)

    Shadrin, Andrey Yurevich; Dvoeglazov, Konstantin Nikolaevich; Ivanov, Valentine Borisovich; Volk, Vladimir Ivanovich; Skupov, Mikhail Vladimirovich; Glushenkov, Alexey Evgenevich [Joint Stock Company ' ' The High Technological Research Institute of Inorganic Materials' ' , Moscow (Russian Federation); Troyanov, Vladimir Mihaylovich; Zherebtsov, Alexander Anatolievich [Innovation and Technology Center of Project ' ' PRORYV' ' , State Atomic Energy Corporation ' ' Rosatom' ' , Moscow (Russian Federation)

    2015-06-01

    The strategies adopted in Russia for a closed nuclear fuel cycle with fast reactors (FR), selection of fuel type and recycling technologies of spent nuclear fuel (SNF) are discussed. It is shown that one of the possible technological solutions for the closing of a fuel cycle could be the combination of pyroelectrochemical and hydrometallurgical methods of recycling of SNF. This combined scheme allows: recycling of SNF from FR with high burn-up and short cooling time; decreasing the volume of stored SNF and the amount of plutonium in a closed fuel cycle in FR; recycling of any type of SNF from FR; obtaining the high pure end uranium-plutonium-neptunium end-product for fuel refabrication using pellet technology.

  8. The IFR modern nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Hannum, W.H.

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Hosteny, R.P.

    1978-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hosteny, R.P.

    1978-10-01

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

  11. Out-of-core fuel cycle optimization for nonequilibrium cycles

    International Nuclear Information System (INIS)

    Comes, S.A.; Turinsky, P.J.

    1988-01-01

    A methodology has been developed for determining the family of near-optimum fuel management schemes that minimize the levelized fuel cycle costs of a light water reactor over a multicycle planning horizon. Feed batch enrichments and sizes, burned batches to reinsert, and burnable poison loadings are determined for each cycle in the planning horizon. Flexibility in the methodology includes the capability to assess the economic benefits of various partially burned bath reload strategies as well as the effects of using split feed enrichments and enrichment palettes. Constraint limitations are imposed on feed enrichments, discharge burnups, moderator temperature coefficient, and cycle energy requirements

  12. Proposed fuel cycle for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Burris, L.; Walters, L.C.

    1985-01-01

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

  13. RU fuel development program for an advanced fuel cycle in Korea

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Williams, Kent Alan; Shropshire, David E.

    2009-01-01

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

  15. Nuclear power and its fuel cycle

    International Nuclear Information System (INIS)

    Wymer, R.G.

    1986-01-01

    A series of viewgraphs describes the nuclear fuel cycle and nuclear power, covering reactor types, sources of uranium, enrichment of uranium, fuel fabrication, transportation, fuel reprocessing, and radioactive wastes

  16. The safety of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-01-01

    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

  17. Economic analysis of thorium-uranium fuel cycle introduced into PWRs

    International Nuclear Information System (INIS)

    Fan Li; Sun Qian

    2014-01-01

    Using PWR of Daya Bay Unit l as the reference reactor, a validated computer code was used to calculate the fuel cycle costs for uranium fuel cycle and thorium-uranium fuel cycle over the following 20 0perational years respectively. The calculation results show that the thorium-uranium fuel cycle is economically competitive with the uranium fuel cycle when reprocessing mode is adopted. For thorium-uranium fuel cycle, if the price of natural uranium is higher than 120 $ /pound U_3O_8, the fuel cycle cost of the direct disposal mode is greater than that of the reprocessing mode. Therefore, when the uranium price may maintain a high level long-termly, adopting reprocessing mode will benefit the economic advantage for the thorium-uranium fuel cycle introduced into PWRs. (authors)

  18. For a Global HTR Marketing Initiative

    International Nuclear Information System (INIS)

    Bredimas, Alexandre; Venneri, Francesco; Richards, Matthew

    2014-01-01

    HTRs are at a crossroads in their history. The technology is proven and the current technical developments relatively mastered but the marketing track record is disappointing. This paper comes to the conclusion that an international, collaborative marketing and communication plan must be implemented in order to address the marketing bottleneck of HTRs. The paper reflects about the HTR product specificities, its unique selling points and its positioning against other nuclear designs and gas cogeneration. It summarises the global market status and demonstrates that the global market for HTRs is there, for electricity generation, industrial cogeneration and polygeneration. The paper finally argues that HTR vendors have a shared interest to unite in order to succeed in activating the market demand for HTR, and suggests an action plan for an international collaboration among HTR vendors to market and communicate globally on HTRs and reach together a critical mass of business leads worldwide, a mutually beneficial outcome. (author)

  19. Integrated fuel-cycle models for fast breeder reactors

    International Nuclear Information System (INIS)

    Ott, K.O.; Maudlin, P.J.

    1981-01-01

    Breeder-reactor fuel-cycle analysis can be divided into four different areas or categories. The first category concerns questions about the spatial variation of the fuel composition for single loading intervals. Questions of the variations in the fuel composition over several cycles represent a second category. Third, there is a need for a determination of the breeding capability of the reactor. The fourth category concerns the investigation of breeding and long-term fuel logistics. Two fuel-cycle models used to answer questions in the third and fourth area are presented. The space- and time-dependent actinide balance, coupled with criticality and fuel-management constraints, is the basis for both the Discontinuous Integrated Fuel-Cycle Model and the Continuous Integrated Fuel-Cycle Model. The results of the continuous model are compared with results obtained from detailed two-dimensional space and multigroup depletion calculations. The continuous model yields nearly the same results as the detailed calculation, and this is with a comparatively insignificant fraction of the computational effort needed for the detailed calculation. Thus, the integrated model presented is an accurate tool for answering questions concerning reactor breeding capability and long-term fuel logistics. (author)

  20. Globalisation of the nuclear fuel cycle

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  1. Commercialization of nuclear fuel cycle business

    International Nuclear Information System (INIS)

    Yakabe, Hideo

    1998-01-01

    Japan depends on foreign countries almost for establishing nuclear fuel cycle. Accordingly, uranium enrichment, spent fuel reprocessing and the safe treatment and disposal of radioactive waste in Japan is important for securing energy. By these means, the stable supply of enriched uranium, the rise of utilization efficiency of uranium and making nuclear power into home-produced energy can be realized. Also this contributes to the protection of earth resources and the preservation of environment. Japan Nuclear Fuel Co., Ltd. operates four business commercially in Rokkasho, Aomori Prefecture, aiming at the completion of nuclear fuel cycle by the technologies developed by Power Reactor and Nuclear Fuel Development Corporation and the introduction of technologies from foreign countries. The conditions of location of nuclear fuel cycle facilities and the course of the location in Rokkasho are described. In the site of about 740 hectares area, uranium enrichment, burying of low level radioactive waste, fuel reprocessing and high level waste control have been carried out, and three businesses except reprocessing already began the operation. The state of operation of these businesses is reported. Hereafter, efforts will be exerted to the securing of safety through trouble-free operation and cost reduction. (K.I.)

  2. Economic analyses of LWR fuel cycles

    International Nuclear Information System (INIS)

    Field, F.R.

    1977-05-01

    An economic comparison was made of three options for handling irradiated light-water reactor (LWR) fuel. These options are reprocessing of spent reactor fuel and subsequent recycle of both uranium and plutonium, reprocessing and recycle of uranium only, and direct terminal storage of spent fuel not reprocessed. The comparison was based on a peak-installed nuclear capacity of 507 GWe by CY 2000 and retirement of reactors after 30 years of service. Results of the study indicate that: Through the year 2000, recycle of uranium and plutonium in LWRs saves about $12 billion (FY 1977 dollars) compared with the throwaway cycle, but this amounts to only about 1.3% of the total cost of generating electricity by nuclear power. If deferred costs are included for fuel that has been discharged from reactors but not reprocessed, the economic advantage increases to $17.7 billion. Recycle of uranium only (storage of plutonium) is approximately $7 billion more expensive than the throwaway fuel cycle and is, therefore, not considered an economically viable option. The throwaway fuel cycle ultimately requires >40% more uranium resources (U 3 O 8 ) than does reprocessing spent fuel where both uranium and plutonium are recycled

  3. Prospects for Australian involvement in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Chandra, S.; Hallenstein, C.

    1988-05-01

    A review of recent overseas developments in the nuclear industry by The Northern Territory Department of Mines and Energy suggests that there are market prospects in all stages of the fuel cycle. Australia could secure those markets through aggressive marketing and competitive prices. This report gives a profile of the nuclear fuel cycle and nuclear fuel cycle technologies, and describes the prospects of Australian involvement in the nuclear fuel cycle. It concludes that the nuclear fuel cycle industry has the potential to earn around $10 billion per year in export income. It recommend that the Federal Government: (1) re-examines its position on the Slayter recommendation (1984) that Australia should develop new uranium mines and further stages of the nuclear fuel cycle, and (2) gives it's in-principle agreement to the Northern Territory to seek expressions of interest from the nuclear industry for the establishment of an integrated nuclear fuel cycle industry in the Northern Territory

  4. The impact of the multilateral approach to the nuclear fuel cycle in Malaysia's nuclear fuel cycle policy

    International Nuclear Information System (INIS)

    Baharuddin, B.; Ferdinand, P.

    2014-01-01

    Since the Pakistan-India nuclear weapon race, the North Korean nuclear test and the September 11 attack revealed Abdul Qadeer Khan's clandestine nuclear black market and the fear that Iran's nuclear program may be used for nuclear weapon development, scrutiny of activities related to nuclear technologies, especially technology transfer has become more stringent. The nuclear supplier group has initiated a multilateral nuclear fuel cycle regime with the purpose of guaranteeing nuclear fuel supply and at the same time preventing the spread of nuclear proliferation. Malaysia wants to develop a programme for the peaceful use of nuclear energy and it needs to accommodate itself to this policy. When considering developing a nuclear fuel cycle policy, the key elements that Malaysia needs to consider are the extent of the fuel cycle technologies that it intends to acquire and the costs (financial and political) of acquiring them. Therefore, this paper will examine how the multilateral approach to the nuclear fuel cycle may influence Malaysia's nuclear fuel cycle policy, without jeopardising the country's rights and sovereignty as stipulated under the NPT. (authors)

  5. Cost aspects of the research reactor fuel cycle

    International Nuclear Information System (INIS)

    2010-01-01

    Research reactors have made valuable contributions to the development of nuclear power, basic science, materials development, radioisotope production for medicine and industry, and education and training. In doing so, they have provided an invaluable service to humanity. Research reactors are expected to make important contributions in the coming decades to further development of the peaceful uses of nuclear technology, in particular for advanced nuclear fission reactors and fuel cycles, fusion, high energy physics, basic research, materials science, nuclear medicine, and biological sciences. However, in the context of decreased public sector support, research reactors are increasingly faced with financial constraints. It is therefore of great importance that their operations are based on a sound understanding of the costs of the complete research reactor fuel cycle, and that they are managed according to sound financial and economic principles. This publication is targeted at individuals and organizations involved with research reactor operations, with the aim of providing both information and an analytical framework for assessing and determining the cost structure of fuel cycle related activities. Efficient management of fuel cycle expenditures is an important component in developing strategies for sustainable future operation of a research reactor. The elements of the fuel cycle are presented with a description of how they can affect the cost efficient operation of a research reactor. A systematic review of fuel cycle choices is particularly important when a new reactor is being planned or when an existing reactor is facing major changes in its fuel cycle structure, for example because of conversion of the core from high enriched uranium (HEU) to low enriched uranium (LEU) fuel, or the changes in spent fuel management provision. Review and optimization of fuel cycle issues is also recommended for existing research reactors, even in cases where research reactor

  6. Equilibrium transuranic management scheme for diverse fuel cycle analysis

    International Nuclear Information System (INIS)

    Haas, Jason; Lee, John C.

    2008-01-01

    A key issue cited in the U.S. Department of Energy's report to Congress (2003) on the research path for the Advanced Fuel Cycle Initiative (AFCI) is an accurate estimation of life cycle costs for the construction, operation, decontamination and decommissioning of all fuel cycle facilities. In this report we discuss the methodology and validation of a fuel cycle model based on equilibrium operation. We apply our model to a diverse set of advanced reactors and fuel types in order to determine the most effective transmuting system while simultaneously minimizing fuel cycle costs. Our analysis shows that a nearly instant equilibrium modeling of fuel cycle scenarios can accurately approximate the detailed complex dynamic models developed by national laboratories. Our analysis also shows that the cost of transmuting Spent Nuclear Fuel (SNF) from a UO 2 fueled Pressurized Water Reactor (PWR) is minimized by utilizing the thorium cycle in sodium cooled fast reactors and is near the cost for long term repository storage of SNF at Yucca Mountain. (authors)

  7. Fuel cycle studies

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    Programs are being conducted in the following areas: advanced solvent extraction techniques, accident consequences, fuel cycles for nonproliferation, pyrochemical and dry processes, waste encapsulation, radionuclide transport in geologic media, hull treatment, and analytical support for LWBR

  8. The HTR-PM Plant Full Scope Training Simulator

    International Nuclear Information System (INIS)

    Wang Junsan; Wang Yuding; Zhou Shuyong; Cai Ruizhong; Cao Jianting

    2014-01-01

    This paper describes the technical aspects of the Full Scope Training Simulator developed for HTR-PM Plant in Shidao Bay, Shandong Province, China. An overview of the HTR-PM plant and simulator structure is presented. The models developed for the simulator are discussed in detail. Some important verification tests have been conducted on the HTR-PM Plant Training Simulator. (author)

  9. Moving towards sustainable thorium fuel cycles

    International Nuclear Information System (INIS)

    Hyland, B.; Hamilton, H.

    2011-01-01

    The CANDU reactor has an unsurpassed degree of fuel-cycle flexibility as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle design. These features facilitate the introduction and full exploitation of thorium fuel cycles in CANDU reactors in an evolutionary fashion. Thoria (ThO 2 ) based fuel offers both fuel performance and safety advantages over urania (UO 2 ) based fuel, due its higher thermal conductivity which results in lower fuel-operating temperatures at similar linear element powers. Thoria fuel has demonstrated lower fission gas release than UO 2 under similar operating powers during test irradiations. In addition, thoria has a higher melting point than urania and is far less reactive in hypothetical accident scenarios owing to the fact that it has only one oxidation state. This paper examines one possible strategy for the introduction of thorium fuel cycles into CANDU reactors. In the short term, the initial fissile material would be provided in a heterogeneous bundle of low-enriched uranium and thorium. The medium term scenario uses homogeneous Pu/Th bundles in the CANDU reactor, further increasing the energy derived from the thorium. In the long term, the full energy potential from thorium would be realized through the recycle of the U-233 in the used fuel. With U-233 recycle in CANDU reactors, plutonium would then only be required to top up the fissile content to achieve the desired burnup. (author)

  10. International Nuclear Fuel Cycle Fact Book. Revision 5

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  11. International Nuclear Fuel Cycle Fact Book. Revision 5

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-01-01

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

  12. International nuclear fuel cycle fact book. Revision 4

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-03-01

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

  13. International nuclear fuel cycle fact book. Revision 4

    International Nuclear Information System (INIS)

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

    1984-03-01

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

  14. Nuclear Fuel Cycle Options Catalog: FY16 Improvements and Additions

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-31

    The United States Department of Energy, Office of Nuclear Energy, Fuel Cycle Technology Program sponsors nuclear fuel cycle research and development. As part of its Fuel Cycle Options campaign, the DOE has established the Nuclear Fuel Cycle Options Catalog. The catalog is intended for use by the Fuel Cycle Technologies Program in planning its research and development activities and disseminating information regarding nuclear energy to interested parties. The purpose of this report is to document the improvements and additions that have been made to the Nuclear Fuel Cycle Options Catalog in the 2016 fiscal year.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-01

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

  16. CANDU-6 fuel optimization for advanced cycles

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  17. Advanced fuel cycles for WWER-1000 reactors

    International Nuclear Information System (INIS)

    Semchenkov, Y. M.; Pavlovichev, A. M.; Pavlov, V. I.; Spirkin, E. I.; Styrin, Y. A.; Kosourov, E. K.

    2007-01-01

    Main stages of Russian uranium fuel development regarding improvement of safety and economics of fuel load operation are presented. Intervals of possible changes in fuel cycle duration have been demonstrated for the use of current and perspective fuel. Examples of equilibrium fuel load patterns have been demonstrated and main core neutronics parameters have been presented. Problems on the use of axial blankets with reduced enrichment in WWER-1000 fuel assemblies are considered. Some results are presented regarding core neutronic characteristics of WWER-1000 at the use of regenerated uranium and uranium-plutonium fuel. Examples of equilibrium fuel cycles for the core partially loaded with MOX fuel from weapon-grade plutonium are also considered (Authors)

  18. Nonproliferation characteristics of advanced fuel cycle concepts

    International Nuclear Information System (INIS)

    Persiani, P.J.

    1998-01-01

    The purpose of this study is to comment on the proliferation characteristic profiles of some of the proposed fuel cycle alternatives to help ensure that nonproliferation concerns are introduced into the early stages of a fuel cycle concept development program, and to perhaps aid in the more effective implementation of the international nonproliferation regime initiatives and safeguards methods and systems. Alternative cycle concepts proposed by several countries involve the recycle of spent fuel without the separation of plutonium from uranium and fission products

  19. Spent fuel management and closed nuclear fuel cycle

    International Nuclear Information System (INIS)

    Kudryavtsev, E.G.

    2012-01-01

    Strategic objectives set by Rosatom Corporation in the field of spent fuel management are given. By 2030, Russia is to create technological infrastructure for innovative nuclear energy development, including complete closure of the nuclear fuel cycle. A target model of the spent NPP nuclear fuel management system until 2030 is analyzed. The schedule for key stages of putting in place the infrastructure for spent NPP fuel management is given. The financial aspect of the problem is also discussed [ru

  20. Dynamic Analysis of the Thorium Fuel Cycle in CANDU Reactors

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Park, Chang Je

    2006-02-01

    The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO 2 UO 2 and ThO 2 UO 2 -DUPIC fuels. The recycling is performed through the dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, the thorium fuel CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products of the multiple recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. From the analysis results, it was found that the closed or partially closed thorium fuel cycle can be constructed through the dry process technology. Also, it is known that both the homogeneous and heterogeneous thorium fuel cycles can reduce the SF accumulation and save the natural uranium resource compared with the once-through cycle. From the material balance view point, the heterogeneous thorium fuel cycle seems to be more feasible. It is recommended, however, the economic analysis should be performed in future

  1. Dynamic Analysis of the Thorium Fuel Cycle in CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Park, Chang Je

    2006-02-15

    The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO{sub 2}UO{sub 2} and ThO{sub 2}UO{sub 2}-DUPIC fuels. The recycling is performed through the dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, the thorium fuel CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products of the multiple recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. From the analysis results, it was found that the closed or partially closed thorium fuel cycle can be constructed through the dry process technology. Also, it is known that both the homogeneous and heterogeneous thorium fuel cycles can reduce the SF accumulation and save the natural uranium resource compared with the once-through cycle. From the material balance view point, the heterogeneous thorium fuel cycle seems to be more feasible. It is recommended, however, the economic analysis should be performed in future.

  2. Some alternatives to the mixed oxide fuel cycle

    International Nuclear Information System (INIS)

    Deonigi, D.E.; Eschbach, E.A.; Goldsmith, S.; Pankaskie, P.J.; Rohrmann, C.A.; Widrig, R.D.

    1977-02-01

    While on initial examination each of the six fuel cycle concepts (tandem cycle, extended burnup, fuel rejuvenation, coprocessing, partial reprocessing, and thorium) described in the report may have some potential for improving safeguards, none of the six appears to have any other major or compelling advantages over the mixed oxide (MOX) fuel cycle. Compared to the MOX cycle, all but coprocessing appear to have major disadvantages, including severe cost penalties. Three of the concepts-tandem, extended burnup, and rejuvenation--share the basic problems of the throwaway cycle (GESMO Alternative 6): without reprocessing, high-level waste volumes and costs are substantially increased, and overall uranium utilization decreases for three reasons. First, the parasitic fission products left in the fuel absorb neutrons in later irradiation steps reducing the overall neutronic efficiencies of these cycles. Second, discarded fuel still has sufficient fissile values to warrant recycle. Third, perhaps most important, the plutonium needed for breeder start-up will not be available; without the breeder, uranium utilization would drop by about a factor of sixty. Two of the concepts--coprocessing and partial reprocessing--involve variations of the basic MOX fuel cycle's chemical reprocessing step to make plutonium diversion potentially more difficult. These concepts could be used with the MOX fuel cycle or in conjunction with the tandem, extended burnup and rejuvenation concepts to eliminate some of the problems with those cycles. But in so doing, the basic impetus for those cycles--elimination of reprocessing for safeguards purposes--no longer exists. Of all the concepts considered, only coprocessing--and particularly the ''master blend'' version--appears to have sufficient promise to warrant a more detailed study. The master blend concept could possibly make plutonium diversion more difficult with minimal impact on the reprocessing and MOX fuel fabrication operations

  3. Modifications to HFEF/S for IFR fuel cycle demonstration

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; Forrester, R.J.; Carnes, M.D.; Rigg, R.H.

    1988-01-01

    Modifications have begun to the Hot Fuel Examination Facility-South (HFEF/S) in order to demonstrate the technology of the integral fast reactor (IFR) fuel cycle. This paper describes the status of the modifications to the facility and briefly reviews the status of the development of the process equipment. The HFEF/S was the demonstration facility for the early Experimental Breeder Reactor II (EBR-II) melt refining/injection-casting fuel cycle. Then called the Fuel Cycle Facility, ∼400 EBR-II fuel assemblies were recycled in the two hot cells of the facility during the 1964-69 period. Since then it has been utilized as a fuels examination facility. The objective of the IFR fuel cycle program is to upgrade HFEF/S to current standards, install new process equipment, and demonstrate the commercial feasibility of the IFR pyroprocess fuel cycle

  4. Structural and Functional Analysis of Human HtrA3 Protease and Its Subdomains.

    Directory of Open Access Journals (Sweden)

    Przemyslaw Glaza

    Full Text Available Human HtrA3 protease, which induces mitochondria-mediated apoptosis, can be a tumor suppressor and a potential therapeutic target in the treatment of cancer. However, there is little information about its structure and biochemical properties. HtrA3 is composed of an N-terminal domain not required for proteolytic activity, a central serine protease domain and a C-terminal PDZ domain. HtrA3S, its short natural isoform, lacks the PDZ domain which is substituted by a stretch of 7 C-terminal amino acid residues, unique for this isoform. This paper presents the crystal structure of the HtrA3 protease domain together with the PDZ domain (ΔN-HtrA3, showing that the protein forms a trimer whose protease domains are similar to those of human HtrA1 and HtrA2. The ΔN-HtrA3 PDZ domains are placed in a position intermediate between that in the flat saucer-like HtrA1 SAXS structure and the compact pyramidal HtrA2 X-ray structure. The PDZ domain interacts closely with the LB loop of the protease domain in a way not found in other human HtrAs. ΔN-HtrA3 with the PDZ removed (ΔN-HtrA3-ΔPDZ and an N-terminally truncated HtrA3S (ΔN-HtrA3S were fully active at a wide range of temperatures and their substrate affinity was not impaired. This indicates that the PDZ domain is dispensable for HtrA3 activity. As determined by size exclusion chromatography, ΔN-HtrA3 formed stable trimers while both ΔN-HtrA3-ΔPDZ and ΔN-HtrA3S were monomeric. This suggests that the presence of the PDZ domain, unlike in HtrA1 and HtrA2, influences HtrA3 trimer formation. The unique C-terminal sequence of ΔN-HtrA3S appeared to have little effect on activity and oligomerization. Additionally, we examined the cleavage specificity of ΔN-HtrA3. Results reported in this paper provide new insights into the structure and function of ΔN-HtrA3, which seems to have a unique combination of features among human HtrA proteases.

  5. Structural and Functional Analysis of Human HtrA3 Protease and Its Subdomains.

    Science.gov (United States)

    Glaza, Przemyslaw; Osipiuk, Jerzy; Wenta, Tomasz; Zurawa-Janicka, Dorota; Jarzab, Miroslaw; Lesner, Adam; Banecki, Bogdan; Skorko-Glonek, Joanna; Joachimiak, Andrzej; Lipinska, Barbara

    2015-01-01

    Human HtrA3 protease, which induces mitochondria-mediated apoptosis, can be a tumor suppressor and a potential therapeutic target in the treatment of cancer. However, there is little information about its structure and biochemical properties. HtrA3 is composed of an N-terminal domain not required for proteolytic activity, a central serine protease domain and a C-terminal PDZ domain. HtrA3S, its short natural isoform, lacks the PDZ domain which is substituted by a stretch of 7 C-terminal amino acid residues, unique for this isoform. This paper presents the crystal structure of the HtrA3 protease domain together with the PDZ domain (ΔN-HtrA3), showing that the protein forms a trimer whose protease domains are similar to those of human HtrA1 and HtrA2. The ΔN-HtrA3 PDZ domains are placed in a position intermediate between that in the flat saucer-like HtrA1 SAXS structure and the compact pyramidal HtrA2 X-ray structure. The PDZ domain interacts closely with the LB loop of the protease domain in a way not found in other human HtrAs. ΔN-HtrA3 with the PDZ removed (ΔN-HtrA3-ΔPDZ) and an N-terminally truncated HtrA3S (ΔN-HtrA3S) were fully active at a wide range of temperatures and their substrate affinity was not impaired. This indicates that the PDZ domain is dispensable for HtrA3 activity. As determined by size exclusion chromatography, ΔN-HtrA3 formed stable trimers while both ΔN-HtrA3-ΔPDZ and ΔN-HtrA3S were monomeric. This suggests that the presence of the PDZ domain, unlike in HtrA1 and HtrA2, influences HtrA3 trimer formation. The unique C-terminal sequence of ΔN-HtrA3S appeared to have little effect on activity and oligomerization. Additionally, we examined the cleavage specificity of ΔN-HtrA3. Results reported in this paper provide new insights into the structure and function of ΔN-HtrA3, which seems to have a unique combination of features among human HtrA proteases.

  6. Progress of the HTR-10 project

    International Nuclear Information System (INIS)

    Zhong, D.; Xu, Y.

    1996-01-01

    This paper briefly introduces the main technical features and the design specifications of the HTR-10. Present status and main progress of the license applications, the design and manufacture of the main components and the engineering experiments as well as the construction of the HTR-10 are summarized. (author). 3 tabs

  7. The role of spent fuel test facilities in the fuel cycle strategy

    International Nuclear Information System (INIS)

    Huang, S. T.; Gross, D. L.; Snyder, N. W.; Woods, W. D.

    1988-01-01

    Disposal of commercial spent nuclear fuels in the major industrialized countries may be categorized into two broad approaches: a once-through policy which will dispose of spent fuels and recycle fissile materials. Within reprocess spent fuels and recycle fissile materials. Within each policy, various technical, licensing, institutional and public issues exist. These issues tend to complicate the formulation of an effective and acceptable fuel cycle strategy which will meet various cost, schedule, and legislative constraints. This paper examines overall fuel cycle strategies from the viewpoint of these underlying technical issues and assesses the roles of spent fuel test facilities in the overall fuel cycles steps. Basic functions of such test facilities are also discussed. The main emphasis is placed on the once-through policy although the reprocessing / recycle policy is also discussed. Benefits of utilizing test facilities in the fuel cycle strategies are explored. The results indicate that substantial benefits may be obtained in terms of minimizing programmatic risks, increasing public confidence, and more effective utilization of overall budgetary resources by structuring and highlighting the test facilities as an important element in the overall strategy

  8. Radioactive Waste Generation in Pyro-SFR Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Gao, Fanxing; Park, Byung Heung; Ko, Won Il

    2011-01-01

    Which nuclear fuel cycle option to deploy is of great importance in the sustainability of nuclear power. SFR fuel cycle employing pyroprocessing (named as Pyro- SFR Cycle) is one promising fuel cycle option in the near future. Radioactive waste generation is a key criterion in nuclear fuel cycle system analysis, which considerably affects the future development of nuclear power. High population with small territory is one special characteristic of ROK, which makes the waste management pretty important. In this study, particularly the amount of waste generation with regard to the promising advanced fuel cycle option was evaluated, because the difficulty of deploying an underground repository for HLW disposal requires a longer time especially in ROK

  9. Uncertainty Analyses of Advanced Fuel Cycles

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  10. 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.

  11. The benefits of longer fuel cycle lengths

    International Nuclear Information System (INIS)

    Kesler, D.C.

    1986-01-01

    Longer fuel cycle lengths have been found to increase generation and improve outage management. A study at Duke Power Company has shown that longer fuel cycles offer both increased scheduling flexibility and increased capacity factors

  12. Pu and MA Management in Thermal HTR, QUO VADIS? Insights from the Euratom PUMA project

    International Nuclear Information System (INIS)

    Kuijper, J.C.

    2013-01-01

    The results of this study demonstrate the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burn-up of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burn-up and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the “wallpaper” fuel does not have advantage over the standard fuel design in this respect

  13. International HTR activities

    International Nuclear Information System (INIS)

    Baust, E.; Weisbrodt, I.

    1989-01-01

    Asea Brown Boveri AG (ABB) and their subsidiary High Temperature Reactor Construction GmbH (HRB) have brought the pebble bed high temperature reactor to the edge of being ready for the market with the construction and operation of the AVR reactor at Juelich and the THTR 300 at Hamm-Uentrop. Siemens/Interatom have developed the HTR modular concept and, together with their partners HRB, KFA, Rheinbraun Bergbauforschung have taken the nuclear process heat project to its present advanced state of development. The further introduction of the HTR to the market is a long-term objective, due to the present market situation. ABB and Siemens AG have therefore agreed to collaborate by forming a joint company. (orig.)

  14. A New Dynamic Model for Nuclear Fuel Cycle System Analysis

    International Nuclear Information System (INIS)

    Choi, Sungyeol; Ko, Won Il

    2014-01-01

    The evaluation of mass flow is a complex process where numerous parameters and their complex interaction are involved. Given that many nuclear power countries have light and heavy water reactors and associated fuel cycle technologies, the mass flow analysis has to consider a dynamic transition from the open fuel cycle to other cycles over decades or a century. Although an equilibrium analysis provides insight concerning the end-states of fuel cycle transitions, it cannot answer when we need specific management options, whether the current plan can deliver these options when needed, and how fast the equilibrium can be achieved. As a pilot application, the government brought several experts together to conduct preliminary evaluations for nuclear fuel cycle options in 2010. According to Table 1, they concluded that the closed nuclear fuel cycle has long-term advantages over the open fuel cycle. However, it is still necessary to assess these options in depth and to optimize transition paths of these long-term options with advanced dynamic fuel cycle models. A dynamic simulation model for nuclear fuel cycle systems was developed and its dynamic mass flow analysis capability was validated against the results of existing models. This model can reflects a complex combination of various fuel cycle processes and reactor types, from once-through to multiple recycling, within a single nuclear fuel cycle system. For the open fuel cycle, the results of the developed model are well matched with the results of other models

  15. Several remarks on the fuel cycle economy

    International Nuclear Information System (INIS)

    Roman Kubin; Rudolf Vespalec

    2007-01-01

    Present paper deals with some aspects influencing significantly cost of nuclear fuel and possibilities of its usage in optimal fuel cycle technology. Our discussion is focused on the phase of fuel procurement that means financial parts of the contract as well as its technical Appendices. Typically the fuel fabrication price is taken as the main economy indicator; nevertheless also many other financial and technical features of the contract must be taken into account in order to reach the best price of electricity sold into public energy grid. Our experience from several international tenders shows that the consistent complex of commercial and technical parameters of the contract is necessary to achieve optimal economic results and prepare proper conditions for advanced fuel cycle technology. Among those essential characteristics are payment conditions and schedule and extent of vendor's services and assistance to the operator. Very important role play also technical parameters, as safety and operational limits, influencing loading pattern quality and operating flexibility. Obviously also a level of operator's fuel cycle technology is a crucial point that is necessary for usage of technical quality of the fuel at the power plant. The final electricity price, produced by the plant, and uranium consumption are the only objective criteria to evaluate economic level of the fuel contract and the fuel cycle at all (Authors)

  16. International Nuclear Fuel Cycle Fact Book

    International Nuclear Information System (INIS)

    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

  17. International Nuclear Fuel Cycle Fact Book

    International Nuclear Information System (INIS)

    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

  18. 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.

  19. Nuclear Fuel Cycle System Analysis (I)

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Kwon, Eun Ha; Kim, Ho Dong; Yoon, Ji Sup; Park, Seong Won

    2006-12-15

    As a nation develops strategies that provide nuclear energy while meeting its various objectives, it must begin with identification of a fuel cycle option that can be best suitable for the country. For such a purpose, this paper takes four different fuel cycle options - Once-through Cycle, DUPIC Recycle, Thermal Reactor Recycle and GEN-IV Recycle, and evaluates each option in terms of sustainability, environment-friendliness, proliferation-resistance and economics. The analysis shows that the GEN-IV Recycle appears to have an advantage in terms of sustainability, environment-friendliness and long-term proliferation-resistance, while it is expected to be more economically competitive, if uranium ore prices increase or costs of pyroprocessing and fuel fabrication decrease.

  20. A new impetus for developing industrial process heat applications of HTR in europe - HTR2008-58259

    International Nuclear Information System (INIS)

    Hittner, D.; De Groot, S.; Griffay, G.; Yvon, P.; Pienkowski, L.; Ruer, J.; Angulo, C.; Laquaniello, G.

    2008-01-01

    Due to its high operating temperature (up to 850 deg. C with present technologies, possibly higher in the longer term), and its power range (a few hundred MW), the modular HTR could address a larger scope of industrial process heat needs than other present nuclear systems. Even if HTR can contribute to competitive electricity generation, this potential for industrial heat applications is the main incentive for developing this type of reactor, as it could open to nuclear energy a large non-electricity market. However several issues must be addressed and solved successfully for HTR to actually enter the market of industrial process heat: 1) as an absolute prerequisite, to develop a strategic alliance of nuclear industry and R and D with process heat user industries. 2) to solve some key technical issues, as for instance the design of a reactor and of a coupling system flexible enough to reconcile a single reactor design with multiple applications and versatile requirements for the heat source, and the development of special adaptations of the application processes or even of new processes to fit with the assets and constraints of HTR heat supply, 3) to solve critical industrial issues such as economic competitiveness, availability and 4) to address the licensing issues raised by the conjunction of nuclear and industrial risks. In line with IAEA initiatives for supporting non-electric applications of nuclear energy and with the orientations of the SET-Plan of the European Commission, the (European) HTR Technology Network (HTR-TN) proposes a new project, together with industrial process heat user partners, to provide a first impetus to the strategic alliance between nuclear and non-nuclear industries. End user requirements will be expressed systematically on the basis of inputs from industrial partners on various types of process heat applications. These requirements will be confronted with the capabilities of the HTR heat source, in order to point out possible

  1. Outlook on to fuel cycle perspectives at WWER-440

    International Nuclear Information System (INIS)

    Stech, S.; Bajgl, J.

    2005-01-01

    Current internal fuel cycle in NPP Dukovany 4x440 MWe is shortly characterized with new types of fuel assemblies and advanced fuel cycles which have been introduced in the last years. The modernization activities accomplished until now might be extrapolated to the further period in fuel design - mechanic, thermal-hydraulic and neutronic respectively - with additional increase in fuel enrichments and burnups on the way to the 6-year cycle. Reaktor power up rating together with Unit thermal efficiency improvements could bring an increase in the electric output to the value nearly 500 MWe. The reasons are given for long-term cooperation with Fuel Supplier and Plant Designer in the area of fuel cycle as well as in Unit Design Basis. All innovations mentioned in the article including future fuel and fuel cycle changes might be a quite realistic perspective at the end of the first decade of the new century (Authors)

  2. Research reactors fuel cycle problems and dilemma

    International Nuclear Information System (INIS)

    Romano, R.

    2004-01-01

    During last 10 years, some problems appeared in different steps of research reactors fuel cycle. Actually the majority of these reactors have been built in the 60s and these reactors were operated during all this long period in a cycle with steps which were dedicated to this activity. Progressively and for reasons often economical, certain steps of the cycle became more and more difficult to manage due to closing of some specialised workshops in the activities of scraps recycling, irradiated fuel reprocessing, even fuel fabrication. Other steps of the cycle meet or will meet difficulties, in particular supplying of fissile raw material LEU or HEU because this material was mostly produced in enrichment units existing mainly for military reason. Rarefaction of fissile material lead to use more and more enriched uraniums said 'of technical quality', that is to say which come from mixing of varied qualities of enriched material, containing products resulting from reprocessing. Actually, problems of end of fuel cycle are increased, either consisting of intermediary storage on the site of reactor or on specialised sites, or consisting of reprocessing. This brief summary shows most difficulties which are met today by a major part of industrials of the fuel cycle in the exercise of their activities

  3. Research and development of thorium fuel cycle

    International Nuclear Information System (INIS)

    Oishi, Jun.

    1994-01-01

    Nuclear properties of thorium are summarized and present status of research and development of the use of thorium as nuclear fuel is reviewed. Thorium may be used for nuclear fuel in forms of metal, oxide, carbide and nitride independently, alloy with uranium or plutonium or mixture of the compound. Their use in reactors is described. The reprocessing of the spent oxide fuel in thorium fuel cycle is called the thorex process and similar to the purex process. A concept of a molten salt fuel reactor and chemical processing of the molten salt fuel are explained. The required future research on thorium fuel cycle is commented briefly. (T.H.)

  4. Factors influencing selection of a HTR for a developing country

    International Nuclear Information System (INIS)

    Karim, C.S.

    1989-01-01

    Consumption of commercial energy and electricity in Bangladesh has to grow rapidly in order to attain socio-economic development of the country. Nuclear power is considered to be an appropriate proposition due to the inadequacy of indigenous primary energy resources. A technical, economic and financial feasibility study of a 300-500 MWe nuclear power plant is underway now. Responses from different suppliers in SMPR range were enumerated jointly by the Consultants and BAEC under the feasibility study. Criteria for selection of technology and the factor influencing the selection of Modular HTR for Bangladesh are described in the paper. Some indicative results of cost economic calculations are included to help form an idea about various limiting conditions, under which a SMPR with the selected technology could become competitive with the other conventional alternatives. Problems in decision making associated with the uncertainties in estimating plant and fuel cycle costs are enumerated. The implications of not having a reference plant vis-a-vis the advantageous safety features are described to show how these aspects can influence the selection of a new technology like HTR for a developing country. Financing is identifiable as the major problem in implementing a nuclear power project in a developing country like Bangladesh. The entire scope of supplies and services may be broken down into components, so that the burden of financing could be shared by more than one exporting country. Some indicative ideas about the packaging of supplies and services are presented in the paper in order to identify different types of financing sources that could be explored for implementation of the project. Some salient features of the effect of joint-venture on the project financing and implementation are described in the paper. (author). 3 refs, 1 fig

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Le Biez, V. [Corps des Mines, 35 bis rue Saint-Sabin, F-75011 Paris (France); Machiels, A.; Sowder, A. [Electric Power Research Institute, Inc. 3420, Hillview Avenue, Palo Alto, CA 94304 (United States)

    2013-07-01

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

  7. Status of IFR fuel cycle demonstration

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; McFarlane, H.F.

    1993-01-01

    The next major step in Argonne's Integral Fast Reactor (IFR) Program is demonstration of the pyroprocess fuel cycle, in conjunction with continued operation of EBR-II. The Fuel Cycle Facility (FCF) is being readied for this mission. This paper will address the status of facility systems and process equipment, the initial startup experience, and plans for the demonstration program

  8. Safeguards operations in the integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Goff, K.M.; Benedict, R.W.; Brumbach, S.B.; Dickerman, C.E.; Tompot, R.W.

    1994-01-01

    Argonne National Laboratory is currently demonstrating the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The safeguards aspects of the fuel cycle demonstration must be approved by the United States Department of Energy, but a further goal of the program is to develop a safeguards system that could gain acceptance from the Nuclear Regulatory Commission and International Atomic Energy Agency. This fuel cycle is described with emphasis on aspects that differ from aqueous reprocessing and on its improved safeguardability due to decreased attractiveness and diversion potential of all process streams, including the fuel product

  9. Use of non-proliferation fuel cycles in the HTGR

    International Nuclear Information System (INIS)

    Baxter, A.M.; Merrill, M.H.; Dahlberg, R.C.

    1978-10-01

    All high-temperature gas-cooled reactors (HTGRs) built or designed to date utilize a uranium-thorium fuel cycle (HEU/Th) in which fully-enriched uranium (93% U-235) is the initial fuel and thorium is the fertile material. The U-233 produced from the thorium is recycled in subsequent loadings to reduce U-235 makeup requirements. However, the recent interest in proliferation-proof fuel cycles for fission reactors has prompted a review and evaluation of possible alternate cycles in the HTGR. This report discusses these alternate fuel cycles, defines those considered usable in an HTGR core, summarizes their advantages and disadvantages, and briefly describes the effect on core design of the most important cycles. Examples from design studies are also given. These studies show that the flexibility afforded by the HTGR coated-particle fuel design allows a variety of alternative cycles, each having special advantages and attractions under different circumstances. Moreover, these alternate cycles can all use the same fuel block, core layout, control scheme, and basic fuel zoning concept

  10. Worldwide status of HTR development

    International Nuclear Information System (INIS)

    1978-06-01

    The International Atomic Energy Agency convened a technical committee meeting on high temperature reactors (HTRs) from 12-14 Dec. 1977 at Agency Headquarters to provide a forum for the exchange of information on the status of HTR development programmes and to receive advice on the Agency programme in this field. The continuing high level of international interest in HTRs was evidenced by the participation from 11 countries and 2 organizations: Austria, Belgium, France, Federal Republic of Germany, Japan, Netherlands, Poland, Switzerland, Union of Soviet Socialist Republics, United Kingdom of Great Britain, United States of America, Commission of the European Communities, and the OECD Nuclear Energy Agency. In order to promote the continuing exchange of technical information through the offices of the IAEA, a recommendation was made that the Agency establish a standing International Working Group on High Temperature Reactors (IWGHTR). This recommendation is being implemented in 1978. Considerable information on recent progress in HTR development was present at the technical committee meeting in technical reports and in progress reports on HTR development programmes. Since this material will not be published, this summary report on the worldwide status of HTR development at the beginning of 1978 has been prepared, based primarily on information presented at the December 1977 meeting

  11. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

  12. A real-time data acquisition and processing system for the analytical laboratory automation of a HTR spent fuel reprocessing facility

    International Nuclear Information System (INIS)

    Watzlawik, K.H.

    1979-12-01

    A real-time data acquisition and processing system for the analytical laboratory of an experimental HTR spent fuel reprocessing facility is presented. The on-line open-loop system combines in-line and off-line analytical measurement procedures including data acquisition and evaluation as well as analytical laboratory organisation under the control of a computer-supported laboratory automation system. In-line measurements are performed for density, volume and temperature in process tanks and registration of samples for off-line measurements. Off-line computer-coupled experiments are potentiometric titration, gas chromatography and X-ray fluorescence analysis. Organisational sections like sample registration, magazining, distribution and identification, multiple data assignment and especially calibrations of analytical devices are performed by the data processing system. (orig.) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-19

    The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared to a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output

  15. Study on the shuffling scheme in HTR-10 MW test module

    International Nuclear Information System (INIS)

    Jing Xingqing; Zhang Xu; Luo Jingyu

    1993-01-01

    The shuffling ways, once through then out and multiple through then out, in HTR-10 MW Test Module are studied. Multiple through then out is better than once through with regard to rational use of the fuel and flattening the power. The behaviour of equilibrium core and loss of coolant accident is analyzed. The results indicate that characteristic features of the multiple through then out could be better to satisfy the demands of safety criterions

  16. WWER-1000 fuel cycles: current situation and outlook

    International Nuclear Information System (INIS)

    Kosourov, E.; Pavlov, V.; Pavlovichev, A.; Spirkin, E.; Shcherenko, A.

    2013-01-01

    Usage mode of nuclear fuel in WWER type reactor has been changed significantly till the moment of the first WWER-1000 commissioning. There are a lot of improvements, having an impact on the fuel cycle, have been implemented for units with WWER-1000. FA design and its constructional materials, FA fuel weight, burnable poison, usage mode of units and etc have been modified. As the result of development it has been designed a modern FA with rigid skeleton. As a whole it allows to use more efficient configurations of the core, to extend range of fuel cycle lengths and to provide good flexibility in the operation. In recent years there were in progress works on increasing FA uranium capacity. As the result there were developed two designs of the fuel rod: 1) the fuel column height of 3680 mm, diameters of the fuel pellet and its central hole of 7.6 and 1.2 mm respectively and 2) the fuel column height of 3530 mm, the fuel pellet diameter of 7.8 mm without the central hole. Such fuel rods have operating experience as a part of different FA designs. Positive operating experience was a base of new FA (TVS-4) development with the fuel column height of 3680 mm and the fuel pellet diameter of 7.8 mm without the central hole. The paper presents the overview of WWER-1000, AES-2006 and WWER-TOI fuel cycles based on FAs with fuel rod designs described above. There are demonstrated fuel cycle possibilities and its technical and economic characteristics. There are discussed problems of further fuel cycle improvements (fuel enrichment increase above 5 %, use of erbium as alternative burnable poison) and their impact on neutronics characteristics. (authors)

  17. Nuclear Fuel Cycle System Analysis (II)

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Kwon, Eun Ha; Yoon, Ji Sup; Park, Seong Won

    2007-04-15

    As a nation develops strategies that provide nuclear energy while meeting its various objectives, it must begin with identification of a fuel cycle option that can be best suitable for the country. For such a purpose, this paper takes four different fuel cycle options that are likely adopted by the Korean government, considering the current status of nuclear power generation and the 2nd Comprehensive Nuclear Energy Promotion Plan (CNEPP) - Once-through Cycle, DUPIC Recycle, Thermal Reactor Recycle and GEN-IV Recycle. The paper then evaluates each option in terms of sustainability, environment-friendliness, proliferation-resistance, economics and technologies. Like all the policy decision, however, a nuclear fuel cycle option can not be superior in all aspects of sustainability, environment-friendliness, proliferation-resistance, economics, technologies and so on, which makes the comparison of the options extremely complicated. Taking this into consideration, the paper analyzes all the four fuel cycle options using the Multi-Attribute Utility Theory (MAUT) and the Analytic Hierarchy Process (AHP), methods of Multi-Attribute Decision Making (MADM), that support systematical evaluation of the cases with multi- goals or criteria and that such goals are incompatible with each other. The analysis shows that the GEN-IV Recycle appears to be most competitive.

  18. Fuel Cycle Requirements Code (FLYER). Summary report

    International Nuclear Information System (INIS)

    Gift, E.H.; Goode, W.D.

    1976-01-01

    Planning for, and the analysis of, the fuel requirements of the nuclear industry requires the ability to evaluate contingencies in many areas of the nuclear fuel cycle. The areas of nuclear fuel utilization, both uranium and plutonium, and of separative work requirements are of particular interest. The Fuel Cycle Requirements (FLYER) model has been developed to provide a flexible, easily managed tool for obtaining a comprehensive analysis of the nuclear fuel cycle. The model allows analysis of the interactions among the nuclear capacity growth rate, reactor technology and mix, and uranium and plutonium recycling capabilities. The model was initially developed as a means of analyzing nuclear growth contingencies with particular emphasis on the uranium feed and separative work requirements. It served to provide the planning group with analyses similar to the OPA's NUFUEL code which has only recently become available for general use. The model has recently been modified to account for some features of the fuel cycle in a more explicit manner than the NUFUEL code. For instance, the uranium requirements for all reactors installed in a given year are calculated for the total lifetime of those reactors. These values are cumulated in order to indicate the total uranium committed for reactors installed by any given year of the campaign. Similarly, the interactions in the back end of the fuel cycle are handled specifically, such as, the impacts resulting from limitations on the industrial capacity for reprocessing and mixed oxide fabrication of both light water reactor and breeder fuels. The principal features of the modified FLYER code are presented in summary form

  19. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Fehrenbach, P.; Duffey, R.; Kuran, S.; Ivanco, M.; Dyck, G.R.; Chan, P.S.W.; Tyagi, A.K.; Mancuso, C.

    2006-01-01

    Over the last decade and a half AECL has established a strong record of delivering CANDU 6 nuclear power plants on time and at budget. Inherently flexible features of the CANDU type reactors, such as on-power fuelling, high neutron economy, fuel channel based heat transport system, simple fuel bundle configuration, two independent shut down systems, a cool moderator and a defence-in-depth based safety philosophy provides an evolutionary path to further improvements in design. The immediate milestone on this path is the Advanced CANDU ReactorTM** (ACRTM**), in the form of the ACR-1000TM**. This effort is being followed by the Super Critical Water Reactor (SCWR) design that will allow water-cooled reactors to attain high efficiencies by increasing the coolant temperature above 550 0 C. Adaptability of the CANDU design to different fuel cycles is another technology advantage that offers an additional avenue for design evolution. Thorium is one of the potential fuels for future reactors due to relative abundance, neutronics advantage as a fertile material in thermal reactors and proliferation resistance. The Thorium fuel cycle is also of interest to China, India, and Turkey due to local abundance that can ensure sustainable energy independence over the long term. AECL has performed an assessment of both CANDU 6 and ACR-1000 designs to identify systems, components, safety features and operational processes that may need to be modified to replace the NU or SEU fuel cycles with one based on Thorium. The paper reviews some of these requirements and the associated practical design solutions. These modifications can either be incorporated into the design prior to construction or, for currently operational reactors, during a refurbishment outage. In parallel with reactor modifications, various Thorium fuel cycles, either based on mixed bundles (homogeneous) or mixed channels (heterogeneous) have been assessed for technical and economic viability. Potential applications of a

  20. Nuclear power and the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1976-07-01

    The IAEA is organizing a major conference on nuclear power and the nuclear fuel cycle, which is to be held from 2 to 13 May 1977 in Salzburg, Austria. The programme for the conference was published in the preceding issue of the IAEA Bulletin (Vol.18, No. 3/4). Topics to be covered at the conference include: world energy supply and demand, supply of nuclear fuel and fuel cycle services, radioactivity management (including transport), nuclear safety, public acceptance of nuclear power, safeguarding of nuclear materials, and nuclear power prospects in developing countries. The articles in the section that follows are intended to serve as an introduction to the topics to be discussed at the Salzburg Conference. They deal with the demand for uranium and nuclear fuel cycle services, uranium supplies, a computer simulation of regional fuel cycle centres, nuclear safety codes, management of radioactive wastes, and a pioneering research project on factors that determine public attitudes toward nuclear power. It is planned to present additional background articles, including a review of the world nuclear fuel reprocessing situation and developments in the uranium enrichment industry, in future issues of the Bulletin. (author)

  1. The dupic fuel cycle synergism between LWR and HWR

    International Nuclear Information System (INIS)

    Lee, J.S.; Yang, M.S.; Park, H.S.; Lee, H.H.; Kim, K.P.; Sullivan, J.D.; Boczar, P.G.; Gadsby, R.D.

    1999-01-01

    The DUPIC fuel cycle can be developed as an alternative to the conventional spent fuel management options of direct disposal or plutonium recycle. Spent LWR fuel can be burned again in a HWR by direct refabrication into CANDU-compatible DUPIC fuel bundles. Such a linkage between LWR and HWR can result in a multitude of synergistic effects, ranging from savings of natural uranium to reductions in the amount of spent fuel to be buried in the earth, for a given amount of nuclear electricity generated. A special feature of the DUPIC fuel cycle is its compliance with the 'Spent Fuel Standard' criteria for diversion resistance, throughout the entire fuel cycle. The DUPIC cycle thus has a very high degree of proliferation resistance. The cost penalty due to this technical factor needs to be considered in balance with the overall benefits of the DUPIC fuel cycle. The DUPIC alternative may be able to make a significant contribution to reducing spent nuclear fuel burial in the geosphere, in a manner similar to the contribution of the nuclear energy alternative in reducing atmospheric pollution from fossil fuel combustion. (author)

  2. Criticality calculations of the HTR-10 pebble-bed reactor with SCALE6/CSAS6 and MCNP5

    International Nuclear Information System (INIS)

    Wang, Meng-Jen; Sheu, Rong-Jiun; Peir, Jinn-Jer; Liang, Jenq-Horng

    2014-01-01

    Highlights: • Comparisons of the HTR-10 criticality calculations with SCALE6/CSAS6 and MCNP5 were performed. • The DOUBLEHET unit-cell treatment provides the best k eff estimation among PBR criticality calculations using SCALE6. • The continuous-energy SCALE6 calculations present a non-negligible discrepancy with MCNP5 in three PBR cases. - Abstract: HTR-10 is a 10 MWt prototype pebble-bed reactor (PBR) that presents a doubly heterogeneous geometry for neutronics calculations. An appropriate unit-cell treatment for the associated fuel elements is vital for creating problem-dependent multigroup cross sections. Considering four unit-cell options for resonance self-shielding correction in SCALE6, a series of HTR-10 core models were established using the CSAS6 sequence to systematically investigate how they affected the computational accuracy and efficiency of PBR criticality calculations. Three core configurations, which ranged from simplified infinite lattices to a detailed geometry, were examined. Based on the same ENDF/B-VII.0 cross-section library, multigroup results were evaluated by comparing with continuous-energy SCALE6/CSAS6 and MCNP5 calculations. The comparison indicated that the INFHOMMEDIUM results overestimated the effective multiplication factor (k eff ) by about 2800 pcm, whereas the LATTICECELL and MULTIREGION treatments overestimated k eff values with similar biases at approximately 470–680 pcm. The DOUBLEHET results attained further improvement, reducing the k eff overestimation to approximately 280 pcm. The comparison yielded two unexpected problems from using SCALE6/CSAS6 in HTR-10 criticality calculations. In particular, the continuous-energy CSAS6 calculations in this study present a non-negligible discrepancy with MCNP5, potentially causing a k eff value overestimate of approximately 680 pcm. Notably, using a cell-weighted mixture instead of an explicit model of individual TRISO particles in the pebble fuel zone does not shorten the

  3. The transparency associated with the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2009-01-01

    This document presents the French national plan for the management of radioactive materials and wastes (PNGMDR - Plan national de gestion des matieres et dechets nucleaires), its elaboration process, its content in terms of nuclear fuel cycle. Then, it describes the control by the ASN of the nuclear fuel cycle, the associated installations, the concerned transports, the 'cycle consistency' approach and its limitations. Propositions are stated aiming at the improvement of the transparency associated with the fuel cycle: to use the PNGMDR, to extend the investigation on the cycle consistency to imported materials and wastes, to improve the transparency on radioactive material transport

  4. Perspectives for the french R and D program for high and very high temperature reactors - HTR2008-58172

    International Nuclear Information System (INIS)

    Yvon, P.; Hittner, D.; Delbecq, J. M.

    2008-01-01

    A R and D programme has been launched addressing the needs of the development of an indirect cycle flexible modular HTR operating at 850 deg. C for electricity generation and/or heat production for industrial processes. In the frame of this program, several significant technical challenges required to demonstrate the viability and performance of the system have been successfully addressed. Design and safety analysis needed the development of computational tools, therefore reactor physics, and thermo-fluid dynamics codes have been developed and are now in the process of being validated in the frame of international code-to-code and code to experiment benchmarks. Most importantly, the performance of the HTR/VHTR fuel identified as TRISO-coated particle must prove to be excellent in operating as well as accidental conditions. A manufacturing and quality control process has been developed and now fuel qualification based on irradiation and heating safety tests is being prepared on the basis of irradiation programs in France and in the frame of the GENERATION IV International Forum (GIF) as well as the development of fuel behaviour models including performance data, failure particle prediction and long-term integrity of the coating. Material and component technologies have been investigated in normal and accident conditions for V/HTR objectives. Significant progress has been made for vessel structures and reactor core structural elements. Major challenges still lie ahead for plate type compact intermediate heat exchangers, especially at temperatures above 850 deg. C, but an alternative solution with helical tubes is also being developed. In order to demonstrate that materials have adequate performance over long service life under impure helium environment and constraints, the research programme focuses on microstructural and mechanical property data, long-term irradiation behaviour, corrosion, modelling and codification of design rules as well as qualification of

  5. Quantities of actinides in nuclear reactor fuel cycles

    International Nuclear Information System (INIS)

    Ang, K.P.

    1975-01-01

    The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000 MW reactors of the following types: water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breeder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium, and recycled uranium. The radioactivity levels of plutonium, americium, and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the United States nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium processed in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing and fuel fabrication to eliminate the off-site transport of separated plutonium. (U.S.)

  6. Fuel performance and operation experience of WWER-440 fuel in improved fuel cycle

    International Nuclear Information System (INIS)

    Gagarinski, A.; Proselkov, V.; Semchenkov, Yu.

    2007-01-01

    The paper summarizes WWER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation (Authors)

  7. Economic evaluation of multilateral nuclear fuel cycle approach

    International Nuclear Information System (INIS)

    Takashima, Ryuta; Kuno, Yusuke; Omoto, Akira; Tanaka, Satoru

    2011-01-01

    Recently previous works have shown that multilateral nuclear fuel cycle approach has benefits not only of non-proliferation but also of cost effectiveness. This is because for most facilities in nuclear fuel cycle, there exist economies of scale, which has a significant impact on the costs of nuclear fuel cycle. Therefore, the evaluation of economic rationality is required as one of the evaluation factors for the multilateral nuclear fuel cycle approach. In this study, we consider some options with respect to multilateral approaches to nuclear fuel cycle in Asian-Pacific region countries that are proposed by the University of Tokyo. In particular, the following factors are embedded into each type: A) no involvement of assurance of services, B) provision of assurance of services including construction of new facility, without transfer of ownership, and C) provision of assurance of service including construction of new joint facilities with ownership transfer of facilities to multilateral nuclear fuel cycle approach. We show the overnight costs taking into account install and operation of nuclear fuel cycle facilities for each option. The economic parameter values such as uranium price, scale factor, and market output expansion influences the total cost for each option. Thus, we show how these parameter values and economic risks affect the total overnight costs for each option. Additionally, the international facilities could increase the risk of transportation for nuclear material compared to national facilities. We discuss the potential effects of this transportation risk on the costs for each option. (author)

  8. Non-proliferation and safeguards aspects of alternative fuel cycle concepts

    International Nuclear Information System (INIS)

    Persiani, P.J.

    1997-01-01

    Timely visibility on the development, evaluation and optimization of fuel cycle concepts with respect to nonproliferation characteristics should be emphasized in the early stage of planning a civilian nuclear power program, by fuel cycle developers, reviewers and decision makers. Fuel cycle technologies have inherently differing levels of nonproliferation characteristic profiles. Institutional and/or multi-national arrangements have been effective in reducing the nonproliferation concerns. The implementation of international safeguards further reduces these concerns by the timely detection of a possible physical diversion of SNM from fuel cycle facilities. Fuel cycles are safeguardable, but the nonproliferation characteristics of fuel cycle concepts differ significantly with consequent impacts on the international level of technical safeguards measures. The paper comments on characteristics of some of the fuel cycle concepts for the purpose of exploring the need to develop advanced nonproliferation and safeguards measures. (author)

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

    KAUST Repository

    Pasunurthi, Shyamsundar

    2017-03-28

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

  10. NPP fuel cycle and assessment of possible options for long-term fuel supply

    International Nuclear Information System (INIS)

    Ignatenko, E.I.; Lebedev, V.M.; Davidenko, N.N.

    1999-01-01

    The purpose of this paper is to present some results of the analysis of the possible options for Russian NPPs fuel supply. In the classical consideration these are four fuel cycles: uranium cycle based on natural uranium, this cycle has several economical advantages with the use of CANDU type reactors with a heavy-water moderator; uranium cycle based on enriched uranium, it is a basis for the current and future nuclear power; uranium-thorium fuel cycle with capabilities which are very promising but unfortunately difficult to implement in practice; plutonium-uranium cycle, in terms of its potential capabilities it is an excellent option, but it is extremely difficult to implement it in practice due to a high activity and toxicity of nuclear materials under recycle. The nuclear power of Russia is currently aimed at using the cheapest fuel resources, that is first of all, uranium reprocessed from industrial reactor fuel and slag-heaps accumulated on the past in isotope-separation plant sites. These resources are enough for the Russian large-scale nuclear power to be developed [ru

  11. Fuel Cycle Services The Heart of Nuclear Energy

    International Nuclear Information System (INIS)

    Soedyartomo-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 CO 2 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)

  12. 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.

  13. Recent developments in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Wunderer, A.

    1984-01-01

    There is a description of the present situation in each individual area of the nuclear fuel cycle. Further topics are: risk and safety factors and emissions from the fuel cycle, availability and disruptions, waste disposal and the storage of radioactive waste. (UA) [de

  14. Closing the fuel cycle: A superior option for India

    International Nuclear Information System (INIS)

    Balu, K.; Purushotham, D.S.C.; Kakodkar, A.

    1999-01-01

    The closed fuel cycle option with reprocessing and recycle of uranium and plutonium (U and Pu) for power generation allows better utilization of the uranium resources. On its part, plutonium is a unique energy source. During the initial years of nuclear fuel cycle activities, reprocessing and recycle of uranium and plutonium for power generation was perceived by many countries to be among the best of long term strategies for the management of spent fuel. But, over the years, some of the countries have taken a position that once-through fuel cycle is both economical and proliferation-resistant. However, such perceptions do vary as a function of economic growth and energy security of a given country. This paper deals with techno-economic perspectives of reprocessing and recycling in the Indian nuclear power programme. Experience of developing Mixed Oxide UO 2 -PuO 2 (MOX) fuel and its actual use in a power reactor (BWR) is presented. The paper further deals with the use of MOX in PHWRs in the future and current thinking, in the Indian context, in respect of advanced fuel cycles for the future. From environmental safety considerations, the separation of long-lived isotopes and minor actinides from high level waste (HLW) would enhance the acceptability of reprocessing and recycle option. The separated actinides are suitable for recycling with MOX fuel. However, the advanced fuel cycles with such recycling of Uranium and transuranium elements call for additional sophisticated fuel cycle activities which are yet to be mastered. India is interested in both uranium and thorium fuel cycles. This paper describes the current status of the Indian nuclear power scenario with reference to the program on reactors, reprocessing and radioactive waste management, plutonium recycle options, thorium-U233 fuel cycle studies and investigations on partitioning of actinides from Purex HLW as relevant to PHWR spent fuels. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-01

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

  16. Dynamic modeling and analysis of alternative fuel cycle scenarios in Korea

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Choi, Hang Bok

    2007-01-01

    The Korean nuclear fuel cycle was modeled by the dynamic analysis method, which was applied to the once-through and alternative fuel cycles. First, the once-through fuel cycle was analyzed based on the Korean nuclear power plant construction plan up to 2015 and a postulated nuclear demand growth rate of zero after 2015. Second, alternative fuel cycles including the direct use of spent pressurized water reactor fuel in Canada deuterium reactors (DUPIC), a sodium-cooled fast reactor and an accelerator driven system were assessed and the results were compared with those of the once-through fuel cycle. The once-through fuel cycle calculation showed that the nuclear power demand would be 25 GWe and the amount of the spent fuel will be ∼65000 tons by 2100. The alternative fuel cycle analyses showed that the spent fuel inventory could be reduced by more than 30% and 90% through the DUPIC and fast reactor fuel cycles, respectively, when compared with the once-through fuel cycle. The results of this study indicate that both spent fuel and uranium resources can be effectively managed if alternative reactor systems are timely implemented along with the existing reactors

  17. Risk assessment of small-sized HTR with pebble-bed core

    International Nuclear Information System (INIS)

    Kroeger, W.; Mertens, J.; Wolters, J.

    1987-01-01

    Two recent concepts of small-sized HTR's (HTR-Modul and HTR-100) were analysed regarding their safety concepts and risk protection. In neither case do core cooling accidents contribute to the risk because of the low induced core temperatures. Water ingress accidents dominate the risk in both cases by detaching deposited fission products which can be released into the environment. For these accident sequences no early fatalities and practically no lethal case of cancer were computed. Both HTR concepts include adequate precautionary measures and an infinitely small risk according to the usual standards. The safety concepts make express use of the specific inherent safety features of pebble-bed HTR's. (orig.)

  18. Concept for fuel-cycle based safeguards

    International Nuclear Information System (INIS)

    deMontmollin, J.M.; Higinbotham, W.A.; Gupta, D.

    1985-01-01

    Although the guidelines for NPT safeguards specify that the State's fuel cycle and degree of international independence are to be taken into account, the same model approach and absolute-quantity inspection goals are applied to all similar facilities, irrespective of the State's fuel cycle, and the findings are reported in those terms. A concept whereby safeguards might more effectively and efficiently accomplish the purposes of NPT safeguards is explored. The principal features are: (1) division of the fuel cycle into three zones, each containing material having a different degree of significance for safeguards; (2) closing a verified material balance around each zone, supplementing the present MBA balances for more sensitive facilities and replacing them for others; (3) maintenance by the IAEA of a current book inventory for each facility by means of immediate, abbreviated reporting of interfacility transfers; (4) near real-time analysis of material flow patterns through the fuel cycle; and (5) a periodic statement of the findings for the entire State that takes the form that there is assurance that all nuclear materials under safeguards are accounted for to some stated degree of uncertainty

  19. Fast breeder fuel cycle

    International Nuclear Information System (INIS)

    1978-07-01

    This contribution is prepared for the answer to the questionnaire of working group 5, subgroup B. B.1. is the short review of the fast breeder fuel cycles based on the reference large commercial Japanese LMFBR. The LMFBRs are devided into two types. FBR-A is the reactor to be used before 2000, and its burnup and breeding ratio are relatively low. The reference fuel cycle requirement is calculated based on the FBR-A. FBR-B is the one to be used after 2000, and its burnup and breeding ratio are relatively high. B.2. is basic FBR fuel reprocessing scheme emphasizing the differences with LWR reprocessing. This scheme is based on the conceptual design and research and development work on the small scale LMFBR reprocessing facility of Japan. The facility adopts a conventional PUREX process except head end portions. The report also describes the effects of technical modifications of conventional reprocessing flow sheets, and the problems to be solved before the adoption of these alternatives

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-25

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

  1. Transparency associated with the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2009-01-01

    This document presents the different fuel cycle stages with which the CEA is associated, the annual flow of materials and wastes produced at these different stages, and the destiny of these produced materials and wastes. These information are given for the different CEA R and D activities: experimentation hot laboratories (activities, fuel cycle stages, list of laboratories, tables giving annual flows for each of them), research reactors (types of reactors, fuel usage modes, annual flows of nuclear materials for each reactor), spent fuel management (different types of used materials), spent fuels and radioactive wastes with a foreign origin (quantities, processes)

  2. Lessons Learned From Dynamic Simulations of Advanced Fuel Cycles

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  3. International issue: the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    In this special issue a serie of short articles of informations are presented on the following topics: the EEC's medium term policy regarding the reprocessing and storage of spent fuel, France's natural uranium supply, the Pechiney Group in the nuclear field, zircaloy cladding for nuclear fuel elements, USSI: a major French nuclear engineering firm, gaseous diffusion: the only commercial enrichment process, the transport of nuclear materials in the fuel cycle, Cogema and spent fuel reprocessing, SGN: a leader in the fuel cycle, quality control of mechanical, thermal and termodynamic design in nuclear engineering, Sulzer's new pump testing station in Mantes, the new look of the Ateliers et Chantiers de Bretagne, tubes and piping in nuclear power plants, piping in pressurized water reactor. All these articles are written in English and in French [fr

  4. Thorium fuel cycle analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, K [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    1980-07-01

    Systems analysis of the thorium cycle, a nuclear fuel cycle accomplished by using thorium, is reported in this paper. Following a brief review on the history of the thorium cycle development, analysis is made on the three functions of the thorium cycle; (1) auxiliary system of U-Pu cycle to save uranium consumption, (2) thermal breeder system to exert full capacity of the thorium resource, (3) symbiotic system to utilize special features of /sup 233/U and neutron sources. The effects of the thorium loading in LWR (Light Water Reactor), HWR (Heavy Water Reactor) and HTGR (High Temperature Gas-cooled Reactor) are considered for the function of auxiliary system of U-Pu cycle. Analysis is made to find how much uranium is saved by /sup 233/U recycling and how the decrease in Pu production influences the introduction of FBR (Fast Breeder Reactor). Study on thermal breeder system is carried out in the case of MSBR (Molten Salt Breeder Reactor). Under a certain amount of fissile material supply, the potential system expansion rate of MSBR, which is determined by fissile material balance, is superior to that of FBR because of the smaller specific fissile inventory of MSBR. For symbiotic system, three cases are treated; i) nuclear heat supply system using HTGR, ii) denatured fuel supply system for nonproliferation purpose, and iii) hybrid system utilizing neutron sources other than fission reactor.

  5. Long-term alternatives for nuclear fuel cycles

    International Nuclear Information System (INIS)

    Vira, J.; Vieno, T.

    1981-07-01

    Several technical alternatives have been proposed to the nuclear spent fuel management but the practical experience on any of these is small or totally lacking. Since the management method is also connected with the composition of fresh fuel, the comparison of the alternatives must include the whole fuel cycle of a nuclear power plant. In the planning of the nuclear fuel cycle over a time range of several decades a consideration must be given, in addition, to the potential of the new reactor types with increased efficiency of uranium utilization. For analyses and mutual comparisons of the fuel cycle alternatives a number of computer models have been designed and implemented at the Technical Research Centre of Finland. Given the estimated boundary conditions the models can be used to study the impact of different goals and requirements on the fuel cycle decisions. Further, they facilitate cost predictions and display information on the role of the intrinsic uncertainties in the decision-making. The conclusions of the study are tied to the questions of price and availability of uranium. Hence, for instance, the benefits from the reprocessing of spent fuel might prove to be small when compared to the costs required, especially as the current reprocessing contracts do not allow the custemer to dismiss the duty of building the final disposal facilities for high level radioactive waste. For a few decades the final decisions can be postponed by extending the interim storage period. Farther in the future the decisions in the nuclear fuel cycle arrangements will more link to the introduction of the fast breeder reactors. (author)

  6. Radioactive characteristics of spent fuels and reprocessing products in thorium fueled alternative cycles

    International Nuclear Information System (INIS)

    Maeda, Mitsuru

    1978-09-01

    In order to provide one fundamental material for the evaluation of Th cycle, compositions of the spent fuels were calculated with the ORIGEN code on following fuel cycles: (1) PWR fueled with Th- enriched U, (2) PWR fueled with Th-denatured U, (3) CANDU fueled with Th-enriched U and (4) HTGR fueled with Th-enriched U. Using these data, product specifications on radioactivity for their reprocessing were calculated, based on a criterion that radioactivities due to foreign elements do not exceed those inherent in nuclear fuel elements, due to 232 U in bred U or 228 Th in recovered Th, respectively. Conclusions are as the following: (1) Because of very high contents of 232 U and 228 Th in the Th cycle fuels from water moderated reactors, especially from PWR, required decontamination factors for their reprocessing will be smaller by a factor of 10 3 to 10 4 , compared with those from U-Pu fueled LWR cycle. (2) These less stringent product specifications on the radioactivity of bred U and recovered Th will justify introduction of some low decontaminating process, with additional advantage of increased proliferation resistance. (3) Decontamination factors required for HTGR fuel will be 10 to 30 times higher than for the other fuels, because of less 232 U and 228 Th generation, and higher burn-up in the fuel. (author)

  7. Advanced fuel cycles: a rationale and strategy for adopting the low-enriched-uranium fuel cycle

    International Nuclear Information System (INIS)

    James, R.A.

    1980-01-01

    A two-year study of alternatives to the natural uranium fuel cycle in CANDU reactors is summarized. The possible advanced cycles are briefly described. Selection criteria for choosing a cycle for development include resource utilization, economics, ease of implementaton, and social acceptability. It is recommended that a detailed study should be made with a view to the early implementation of the low-enriched uranium cycle. (LL)

  8. The DUPIC alternative for backend fuel cycle

    International Nuclear Information System (INIS)

    Lee, J.S.; Choi, J.W.; Park, H.S.; Boczar, P.; Sullivan, J.; Gadsby, R.D.

    1997-01-01

    From the early nineties, a research programme, called DUPIC (Direct Use of Spent PWR Fuel in CANDU) has been undertaken in an international exercise involving Korea, Canada, the U.S. and later the IAEA. The basic idea of this fuel cycle alternative is that the spent fuel from LWR contains enough fissile remnant to be burnt again in CANDUs thanks to its excellent neutron economy. A systematic R and D plan has now gained a full momentum to verify experimentally the DUPIC fuel cycle concept. 4 refs

  9. Nuclear-fuel-cycle education: Module 10. Environmental consideration

    International Nuclear Information System (INIS)

    Wethington, J.A.; Razvi, J.; Grier, C.; Myrick, T.

    1981-12-01

    This educational module is devoted to the environmental considerations of the nuclear fuel cycle. Eight chapters cover: National Environmental Policy Act; environmental impact statements; environmental survey of the uranium fuel cycle; the Barnwell Nuclear Fuel Reprocessing Plant; transport mechanisms; radiological hazards in uranium mining and milling operations; radiological hazards of uranium mill tailings; and the use of recycle plutonium in mixed oxide fuel

  10. Waste management and the holistic fuel cycle

    International Nuclear Information System (INIS)

    Holmes, R.G.G.; Robbins, R.A.; Eilbeck, A.

    1996-01-01

    This paper outlines a holistic approach to the nuclear fuel cycle and the impact that waste management can have on the holistic approach. The philosophy includes regarding irradiated fuel as a resource rather than a waste that can be used as a source of fissile material to be recycled, either Uranium returned to fuel or Plutonium in mixed oxide fuels (MOX) for fast and impact of those compounds that leave the cycle (solid waste, liquid effluent and gaseous effluent) are minimized. This can only be achieved by applying a full life cycle analysis of process benefits. The paper describes some of the work in waste management but notes that waste and its generation must be seen as an integral part of any developed strategy. (authors)

  11. Increased fuel burn-up and fuel cycle equilibrium

    International Nuclear Information System (INIS)

    Debes, M.

    2001-01-01

    Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO 2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  13. New technology and fuel cycles

    International Nuclear Information System (INIS)

    Mooradian, A.J.

    1979-06-01

    The means of improving uranium utilization in nuclear power reactors are reviewed with respect to economic considerations, assurance of adequate fuel supplies and risk of weapons proliferation. Reference is made to what can be done to improve fuel economy in existing reactor systems operating on a once-through fuel cycle and the potential for improvement offered by fuel recycle in those systems. The state of development of new reactor systems that offer significant savings in uranium utilization is also reviewed and conclusions are made respecting the policy implications of the search for fuel economy. (author)

  14. Fuel cycle math - part two

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This article is Part 2 of a two part series on simple mathematics associated with the nuclear fuel cycle. While not addressing any of the financial aspects of the fuel cycle, this article does discuss the following: conversion between English and metric systems; uranium content expressed in equivalent forms, such as U3O8, and the method of determining these equivalencies; the uranium conversion process, considering different input and output compounds; and the enrichment process, including feed, tails, and product assays, as well as SWU and feed requirements

  15. Report of Nuclear Fuel Cycle Subcommittee

    International Nuclear Information System (INIS)

    1982-01-01

    In order to secure stable energy supply over a long period of time, the development and utilization of atomic energy have been actively promoted as the substitute energy for petroleum. Accordingly, the establishment of nuclear fuel cycle is indispensable to support this policy, and efforts have been exerted to promote the technical development and to put it in practical use. The Tokai reprocessing plant has been in operation since the beginning of 1981, and the pilot plant for uranium enrichment is about to start the full scale operation. Considering the progress in the refining and conversion techniques, plutonium fuel fabrication and son on, the prospect to technically establish the nuclear fuel cycle in Japan has been bright. The important problem for the future is to put these techniques in practical use economically. The main point of technical development hereafter is the enlargement and rationalization of the techniques, and the cooperation of the government and the people, and the smooth transfer of the technical development results in public corporations to private organization are necessary. The important problems for establishing the nuclear fuel cycle, the securing of enriched uranium, the reprocessing of spent fuel, unused resources, and the problems related to industrialization, location and fuel storing are reported. (Kako, I.)

  16. Nuclear Fuel Cycle Analysis and Simulation Tool (FAST)

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Kwon, Eun Ha; Kim, Ho Dong

    2005-06-15

    This paper describes the Nuclear Fuel Cycle Analysis and Simulation Tool (FAST) which has been developed by the Korea Atomic Energy Research Institute (KAERI). Categorizing various mix of nuclear reactors and fuel cycles into 11 scenario groups, the FAST calculates all the required quantities for each nuclear fuel cycle component, such as mining, conversion, enrichment and fuel fabrication for each scenario. A major advantage of the FAST is that the code employs a MS Excel spread sheet with the Visual Basic Application, allowing users to manipulate it with ease. The speed of the calculation is also quick enough to make comparisons among different options in a considerably short time. This user-friendly simulation code is expected to be beneficial to further studies on the nuclear fuel cycle to find best options for the future all proliferation risk, environmental impact and economic costs considered.

  17. Regulation at nuclear fuel cycle

    International Nuclear Information System (INIS)

    2002-01-01

    This bulletin contains information about activities of the Nuclear Regulatory Authority of the Slovak Republic (UJD). In this leaflet the role of the UJD in regulation at nuclear fuel cycle is presented. The Nuclear Fuel Cycle (NFC) is a complex of activities linked with production of nuclear fuel for nuclear reactors as a source of energy used for production of electricity and heat, and of activities linked with spent nuclear fuel handling. Activities linked with nuclear fuel (NF) production, known as the Front-End of Nuclear Fuel Cycle, include (production of nuclear fuel from uranium as the most frequently used element). After discharging spent nuclear fuel (SNF) from nuclear reactor the activities follow linked with its storage, reprocessing and disposal known as the Back-End of Nuclear Fuel Cycle. Individual activity, which penetrates throughout the NFC, is transport of nuclear materials various forms during NF production and transport of NF and SNF. Nuclear reactors are installed in the Slovak Republic only in commercial nuclear power plants and the NFC is of the open type is imported from abroad and SNF is long-term supposed without reprocessing. The main mission of the area of NFC is supervision over: - assurance of nuclear safety throughout all NFC activities; - observance of provisions of the Treaty on Non-Proliferation of Nuclear Weapons during nuclear material handling; with an aim to prevent leakage of radioactive substances into environment (including deliberated danage of NFC sensitive facilities and misuse of nuclear materials to production of nuclear weapons. The UJD carries out this mission through: - assessment of safety documentation submitted by operators of nuclear installations at which nuclear material, NF and SNF is handled; - inspections concentrated on assurance of compliance of real conditions in NFC, i.e. storage and transport of NF and SNF; storage, transport and disposal of wastes from processing of SNF; with assumptions of the safety

  18. Uranium Resource Availability Analysis of Four Nuclear Fuel Cycle Options

    International Nuclear Information System (INIS)

    Youn, S. R.; Lee, S. H.; Jeong, M. S.; Kim, S. K.; Ko, W. I.

    2013-01-01

    Making the national policy regarding nuclear fuel cycle option, the policy should be established in ways that nuclear power generation can be maintained through the evaluation on the basis of the following aspects. To establish the national policy regarding nuclear fuel cycle option, that must begin with identification of a fuel cycle option that can be best suited for the country, and the evaluation work for that should be proceeded. Like all the policy decision, however, a certain nuclear fuel cycle option cannot be superior in all aspects of sustain ability, environment-friendliness, proliferation-resistance, economics, technologies, which make the comparison of the fuel cycle options very complicated. For such a purpose, this paper set up four different fuel cycle of nuclear power generation considering 2nd Comprehensive Nuclear Energy Promotion Plan(CNEPP), and analyzed material flow and features in steady state of all four of the fuel cycle options. As a result of an analysis on material flow of each nuclear fuel cycle, it was analyzed that Pyro-SFR recycling is most effective on U resource availability among four fuel cycle option. As shown in Figure 3, OT cycle required the most amount of U and Pyro-SFR recycle consumed the least amount of U. DUPIC recycling, PWR-MOX recycling, and Pyro-SFR recycling fuel cycle appeared to consumed 8.2%, 12.4%, 39.6% decreased amount of uranium respectively compared to OT cycle. Considering spent fuel can be recycled as potential energy resources, U and TRU taken up to be 96% is efficiently used. That is, application period of limited uranium natural resources can be extended, and it brings a great influence on stable use of nuclear energy

  19. Fuel cycles - a key to future CANDU success

    International Nuclear Information System (INIS)

    Kuran, S.; Hopwood, J.; Hastings, I.J.

    2011-01-01

    Globally, fuel cycles are being evaluated as ways of extending nuclear fuel resources, addressing security of supply and reducing back-end spent-fuel management. Current-technology thermal reactors and future fast reactors are the preferred platform for such fuel cycle applications and as an established thermal reactor with unique fuel-cycle capability, CANDU will play a key role in fulfilling such a vision. The next step in the evolution of CANDU fuel cycles will be the introduction of Recovered Uranium (RU), derived from conventional reprocessing. A low-risk RU option applicable in the short term comprises a combination of RU and Depleted Uranium (DU), both former waste streams, giving a Natural Uranium Equivalent (NUE) fuel. This option has been demonstrated in China, and all test bundles have been removed from the Qinshan 1 reactor. Additionally, work is being done on an NUE full core, a Thorium demonstration irradiation and an Advanced Fuel CANDU Reactor(AFCR). AECL is developing other fuel options for CANDU, including actinide waste burning. AECL has developed the Enhanced CANDU 6 (EC6) reactor, upgraded from its best-performing CANDU 6 design. High neutron economy, on-power refueling and a simple fuel bundle provide the EC6 with the flexibility to accommodate a range of advanced fuels, in addition to its standard natural uranium. (author)

  20. 77 FR 19278 - Informational Meeting on Nuclear Fuel Cycle Options

    Science.gov (United States)

    2012-03-30

    ... DEPARTMENT OF ENERGY Informational Meeting on Nuclear Fuel Cycle Options AGENCY: Office of Fuel... activities leading to a comprehensive evaluation and screening of nuclear fuel cycle options in 2013. At this... fuel cycle options developed for the evaluation and screening provides a comprehensive representation...

  1. Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

    International Nuclear Information System (INIS)

    Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L.; Saito, M.

    2003-01-01

    Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, 237 Np, 238 Pu, 231 Pa, 232 U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations

  2. Advanced Fuel Cycle Cost Basis – 2017 Edition

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-29

    This report, commissioned by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the DOE Nuclear Technology Research and Development (NTRD) Program (previously the Fuel Cycle Research and Development (FCRD) and the Advanced Fuel Cycle Initiative (AFCI)). The report describes the NTRD cost basis development process, reference information on NTRD cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for numerous fuel cycle cost modules (modules A-O) as well as cost modules for a number of reactor types (R modules). The fuel cycle cost modules were developed in the areas of natural uranium mining and milling, thorium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, managed decay storage, recycled product storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste. Since its inception, this report has been periodically updated. The last such internal document was published in August 2015 while the last external edition was published in December of 2009 as INL/EXT-07-12107 and is available on the Web at URL: www.inl.gov/technicalpublications/Documents/4536700.pdf. This current report (Sept 2017) is planned to be reviewed for external release, at which time it will replace the 2009 report as an external publication. This information is used in the ongoing evaluation of nuclear fuel cycles by the NE NTRD program.

  3. Burnup effect on nuclear fuel cycle cost using an equilibrium model

    International Nuclear Information System (INIS)

    Youn, S. R.; Kim, S. K.; Ko, W. I.

    2014-01-01

    The degree of fuel burnup is an important technical parameter to the nuclear fuel cycle, being sensitive and progressive to reduce the total volume of process flow materials and eventually cut the nuclear fuel cycle costs. This paper performed the sensitivity analysis of the total nuclear fuel cycle costs to changes in the technical parameter by varying the degree of burnups in each of the three nuclear fuel cycles using an equilibrium model. Important as burnup does, burnup effect was used among the cost drivers of fuel cycle, as the technical parameter. The fuel cycle options analyzed in this paper are three different fuel cycle options as follows: PWR-Once Through Cycle(PWR-OT), PWR-MOX Recycle, Pyro-SFR Recycle. These fuel cycles are most likely to be adopted in the foreseeable future. As a result of the sensitivity analysis on burnup effect of each three different nuclear fuel cycle costs, PWR-MOX turned out to be the most influenced by burnup changes. Next to PWR-MOX cycle, in the order of Pyro-SFR and PWR-OT cycle turned out to be influenced by the degree of burnup. In conclusion, the degree of burnup in the three nuclear fuel cycles can act as the controlling driver of nuclear fuel cycle costs due to a reduction in the volume of spent fuel leading better availability and capacity factors. However, the equilibrium model used in this paper has a limit that time-dependent material flow and cost calculation is impossible. Hence, comparative analysis of the results calculated by dynamic model hereafter and the calculation results using an equilibrium model should be proceed. Moving forward to the foreseeable future with increasing burnups, further studies regarding alternative material of high corrosion resistance fuel cladding for the overall

  4. Nuclear Fuel Cycle Objectives

    International Nuclear Information System (INIS)

    2013-01-01

    . The four Objectives publications include Nuclear General Objectives, Nuclear Power Objectives, Nuclear Fuel Cycle Objectives, and Radioactive Waste management and Decommissioning Objectives. This publication sets out the objectives that need to be achieved in the area of the nuclear fuel cycle to ensure that the Nuclear Energy Basic Principles are satisfied. Within each of these four Objectives publications, the individual topics that make up each area are addressed. The five topics included in this publication are: resources; fuel engineering and performance; spent fuel management and reprocessing; fuel cycles; and the research reactor nuclear fuel cycle

  5. Nuclear Fuel Cycle Evaluation and Screening Findings on Partitioning and Transmutation

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Taiwo, T.A.; Gehin, J.C.; Jubin, R.; Todosow, M.

    2015-01-01

    A Nuclear Fuel Cycle Evaluation and Screening (E and S) study has recently been completed in the United States. The study considered the entire fuel cycle, included considerations for both once-through and recycle fuel cycle options, evaluated a set of 40 fuel cycles that allowed a comprehensive assessment of fuel cycle performance, identified a relatively small number of promising fuel cycle options that have the potential for achieving substantial improvements compared to the current nuclear fuel cycle in the United States, and allowed the identification of research and development (R and D) activities needed to support the development of the promising fuel cycle options. Nine high-level criteria (Nuclear Waste Management, Proliferation Risk, Nuclear Material Security Risk, Safety, Environmental Impact, Resource Utilisation, Development and Deployment Risk, Institutional Issues, and Financial Risk and Economics) and associated metrics were used in the study to compare the performance of nuclear fuel cycle options to that of the current fuel cycle practiced in the United States. The study also evaluated a number of fuel cycle characteristics that may have the potential to impact future R and D directions. These included for example: 1) The fuel resources used, i. e., uranium and/or thorium. 2) Impact of extremely high burnup fuels. 3) Minor actinide recycle. 4) The impact of losses during separations (partitioning). 5) Critical versus subcritical (externally-driven) systems for material irradiation. 6) Impact of spectrum of irradiation system, i.e., fast, thermal or intermediate. 7) Waste generation reduction, all of which were quantified in the study. The E and S study has implemented a framework that can be used now and in the future to objectively inform on the potential of alternative nuclear fuel cycles, providing decision-makers and others with perspective on fuel cycle capabilities. (authors)

  6. Fuel-cycle assessment of selected bioethanol production

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  7. Fuel-cycle assessment of selected bioethanol production.

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-01-31

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

  8. Nonproliferation norms in civilian nuclear fuel cycle

    International Nuclear Information System (INIS)

    Kawata, Tomio

    2005-01-01

    For sustainable use of nuclear energy in large scale, it seems inevitable to choose a closed cycle option. One of the important questions is, then, whether we can really achieve the compatibility between civilian nuclear fuel cycle and nonproliferation norms. In this aspect, Japan is very unique because she is now only one country with full-scope nuclear fuel cycle program as a non-nuclear weapon state in NPT regime. In June 2004 in the midst of heightened proliferation concerns in NPT regime, the IAEA Board of Governors concluded that, for Japanese nuclear energy program, non-diversion of declared nuclear material and the absence of undeclared nuclear material and activities were verified through the inspections and examinations under Comprehensive Safeguards and the Additional Protocol. Based on this conclusion, the IAEA announced the implementation of Integrated Safeguards in Japan in September 2004. This paper reviews how Japan has succeeded in becoming the first country with full-scope nuclear fuel cycle program to qualify for integrated Safeguards, and identifies five key elements that have made this achievement happen: (1) Obvious need of nuclear fuel cycle program, (2) Country's clear intention for renunciation of nuclear armament, (3) Transparency of national nuclear energy program, (4) Record of excellent compliance with nonproliferation obligations for many decades, and (5) Numerous proactive efforts. These five key elements will constitute a kind of an acceptance model for civilian nuclear fuel cycle in NNWS, and may become the basis for building 'Nonproliferation Culture'. (author)

  9. 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.

  10. Physics challenges for advanced fuel cycle assessment

    Energy Technology Data Exchange (ETDEWEB)

    Salvatores, Massimo; Aliberti, Gerardo; Palmiotti, Giuseppe

    2014-06-17

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

  11. The nuclear fuel cycle in the 21st century

    International Nuclear Information System (INIS)

    Todreas, Neil E.

    2004-01-01

    As we enter the 21st century and contemplate the deployment of Generation III+ machines and the development of Generation IV systems, the fuel cycle within which these reactors are to operate has become a predominant consideration. The four challenges to nuclear development of the 21st century of economics, safety, sustainability through spent fuel management and efficient fuel utilization, and proliferation resistance increasingly involve the front and back ends of the fuel cycle equally if not more than the design of the reactor which has reached a far higher level of maturity. It is tempting to accept the closed cycle with its promise of effective waste management as inevitable. The central questions, however, are the characteristics of the desired closed cycle, the relative advantages of thermal versus fast spectrum closed cycles, the character and pace of the transition to a closed cycle, and finally the most central question as to whether the closed cycle is indeed more desirable a choice than is an open cycle. The desired closed fuel cycle for the long term around which this paper is based is full actinide recycle with natural uranium feed and only fission products discharged to an ultimate waste repository. It is concluded that a major international research and development program to achieve this fuel cycle is important to pursue. However, the need to decide for the closed cycle and deploy it is not pressing for the next several decades. (author)

  12. Benefits of cycle stretchout in pressurized water reactor extended-burnup fuel cycles

    International Nuclear Information System (INIS)

    Matzie, R.A.; Leung, D.C.; Liu, Y.; Beekmann, R.W.

    1981-01-01

    Nuclear reactors are inherently capable of operating for a substantial period beyond their nominal end of cycle (EOC) as a result of negative moderator and fuel temperature coefficients and the decrease in xenon poisoning with lower core power levels. This inherent capability can be used to advantage to reduce annual uranium makeup requirements and cycle energy costs by the use of planned EOC stretchout. This paper discusses the fuel utilization efficiency and economics of both the five-batch, extended-burnup cycle and the three-batch, standard-burnup cycle, which can be improved by employing planned EOC (end of cycle) stretchout. 11 refs

  13. Fuel cell hybrid taxi life cycle analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-15

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

  14. Fuel cell hybrid taxi life cycle analysis

    International Nuclear Information System (INIS)

    Baptista, Patricia; Ribau, Joao; Bravo, Joao; Silva, Carla; Adcock, Paul; Kells, Ashley

    2011-01-01

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

  15. The Proliferation Resistance of a Nuclear Fuel Cycle Using Fuel Recovered from the Electrolytic Reduction of Pressurized Water Reactor Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Jung Min; Cochran, Thomas; Mckinzie, Matthew [NRDC, Washington, (United States)

    2016-05-15

    At some points in the fuel cycle, a level of intrinsic or technical proliferation-resistance can be provided by radiation barriers that surround weapons-usable materials. In this report we examine some aspects of intrinsic proliferation resistance of a fuel cycle for a fast neutron reactor that uses fuel recovered from the electrolytic reduction process of pressurized water reactor spent fuel, followed by a melt-refining process. This fuel cycle, proposed by a nuclear engineer at the Korea Advanced Institute of Science and Technology (KAIST), is being examined with respect to its potential merits of higher fuel utilization, lower production of radioactive byproducts, and better economics relative to a pyroprocesing-based fuel cycle. With respect to intrinsic proliferation resistance, however, we show that since europium is separated out during the electrolytic reduction process, this fuel cycle has little merit beyond that of a pyroprocessing-based fuel cycle because of the lower radiation barrier of its recovered materials containing weapons-usable actinides. Unless europium is not separated following voloxidation, the proposed KAIST fuel cycle is not intrinsically proliferation resistant and in this regard does not represent a significant improvement over pyroprocessing. We suggest further modification of the proposed KAIST fuel cycle, namely, omitting electrolytic reduction and melt reduction, and producing the fast reactor fuel directly following voloxidation.

  16. The nuclear fuel cycle light and shadow

    International Nuclear Information System (INIS)

    Giraud, A.

    1977-01-01

    The nuclear fuel cycle industry has a far reaching effect on future world energy developments. The growth in turnover of this industry follows a known patterm; by 1985 this turnover will have reached a figure of 2 billion dollars. Furthermore, the fuel cycle plays a determining role in ensuring the physical continuity of energy supplies for countries already engaged in the nuclear domain. Finally, the development of this industry is subject to economic and political constraints which imply the availability of raw materials, technological know-how, and production facilities. Various factors which could have an adverse influence on the cycle: technical, economic, or financial difficulties, environmental impact, nuclear safety, theft or diversion of nuclear materials, nuclear weapon, proliferation risks, are described, and the interaction between the development of the cycle, energy independance, and the fulfillment of nuclear energy programs is emphasized. It is concluded that the nuclear fuel cycle industry is confronted with difficulties due to its extremely rapid growth rate (doubling every 5 years); it is a long time since such a growth rate has been experienced by any heavy industry. The task which lays before us is difficult, but the fruit is worth the toil, as it is the fuel cycle which will govern the growth of the nuclear industry [fr

  17. Evaluation and optimization of LWR fuel cycles

    International Nuclear Information System (INIS)

    Akbas, T.; Zabunoglu, O.; Tombakoglu, M.

    2001-01-01

    There are several options in the back-end of the nuclear fuel cycle. Discharge burn-up, length of interim storage period, choice of direct disposal or recycling and method of reprocessing in case of recycling affect the options and determine/define the fuel cycle scenarios. These options have been evaluated in viewpoint of some tangible (fuel cycle cost, natural uranium requirement, decay heat of high level waste, radiological ingestion and inhalation hazards) and intangible factors (technological feasibility, nonproliferation aspect, etc.). Neutronic parameters are calculated using versatile fuel depletion code ORIGEN2.1. A program is developed for calculation of cost related parameters. Analytical hierarchy process is used to transform the intangible factors into the tangible ones. Then all these tangible and intangible factors are incorporated into a form that is suitable for goal programming, which is a linear optimization technique and used to determine the optimal option among alternatives. According to the specified objective function and constraints, the optimal fuel cycle scenario is determined using GPSYS (a linear programming software) as a goal programming tool. In addition, a sensitivity analysis is performed for some selected important parameters

  18. Verification test of control rod system for HTR-10

    International Nuclear Information System (INIS)

    Zhou Huizhong; Diao Xingzhong; Huang Zhiyong; Cao Li; Yang Nianzu

    2002-01-01

    There are 10 sets of control rods and driving devices in 10 MW High Temperature Gas-cooled Test Reactor (HTR-10). The control rod system is the controlling and shutdown system of HTR-10, which is designed for reactor criticality, operation, and shutdown. In order to guarantee technical feasibility, a series of verification tests were performed, including room temperature test, thermal test, test after control rod system installed in HTR-10, and test of control rod system before HTR-10 first criticality. All the tests data showed that driving devices working well, control rods running smoothly up and down, random position settling well, and exactly position indicating

  19. On the problems of the fuel cycles of nuclear fuels

    International Nuclear Information System (INIS)

    Schmidt-Kuester, W.J.; Wagner, H.F.

    1976-01-01

    A secured procurement with nuclear energy can be only achieved if a completely closed fuel cycle will be established. In the Federal Republic of Germany efforts are concentrated on the front end as well as on the back end of the fuel cycle. At the front end the main tasks are to secure uranium supply and to establish the necessary enrichment capacity. The German concept for the back end of the fuel cycle will provide for an integrated and co-located system for all necessary facilities including reprocessing, plutonium fuel fabrication, treatment, interim storage and final disposal of the radioactive wastes to be operational in the mid-80's. Responsibilities for establishing this system are shared between government and private industry. Government will provide for final waste disposal, industry will built and operate the other facilities. Another important point for the introduction of nuclear energy is to solve reliably the problems of protection of fissionable material, radioactive waste and nuclear facilities. German government has initiated respective activities and has started appropriate R+D-work. (orig.) [de

  20. OECD/NEA Ongoing activities related to the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Cornet, S.M.; McCarthy, K.; Chauvin, N.

    2013-01-01

    As part of its role in encouraging international collaboration, the OECD Nuclear Energy Agency is coordinating a series of projects related to the Nuclear Fuel Cycle. The Nuclear Science Committee (NSC) Working Party on Scientific Issues of the Nuclear Fuel Cycle (WPFC) comprises five different expert groups covering all aspects of the fuel cycle from front to back-end. Activities related to fuels, materials, physics, separation chemistry, and fuel cycles scenarios are being undertaken. By publishing state-of-the-art reports and organizing workshops, the groups are able to disseminate recent research advancements to the international community. Current activities mainly focus on advanced nuclear systems, and experts are working on analyzing results and establishing challenges associated to the adoption of new materials and fuels. By comparing different codes, the Expert Group on Advanced Fuel Cycle Scenarios is aiming at gaining further understanding of the scientific issues and specific national needs associated with the implementation of advanced fuel cycles. At the back end of the fuel cycle, separation technologies (aqueous and pyrochemical processing) are being assessed. Current and future activities comprise studies on minor actinides separation and post Fukushima studies. Regular workshops are also organized to discuss recent developments on Partitioning and Transmutation. In addition, the Nuclear Development Committee (NDC) focuses on the analysis of the economics of nuclear power across the fuel cycle in the context of changes of electricity markets, social acceptance and technological advances and assesses the availability of the nuclear fuel and infrastructure required for the deployment of existing and future nuclear power. The Expert Group on the Economics of the Back End of the Nuclear Fuel Cycle (EBENFC), in particular, is looking at assessing economic and financial issues related to the long term management of spent nuclear fuel. (authors)