Advanced Research Reactor Fuel Development

Energy Technology Data Exchange (ETDEWEB)

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

2006-04-15

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

Pyrometallurgical processing of Integral Fast Reactor metal fuels

International Nuclear Information System (INIS)

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

1991-01-01

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

A proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the Integral Fast Reactor

International Nuclear Information System (INIS)

Burris, L.; Steindler, M.; Miller, W.

1984-01-01

The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps -- a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented

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

Gaseous fuel reactors for power systems

International Nuclear Information System (INIS)

Helmick, H.H.; Schwenk, F.C.

1978-01-01

The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

Research reactor fuel - an update

International Nuclear Information System (INIS)

Finlay, M.R.; Ripley, M.I.

2003-01-01

In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

Remote Handling Devices for Disposition of Enriched Uranium Reactor Fuel Using Melt-Dilute Process

International Nuclear Information System (INIS)

Heckendorn, F.M.

2001-01-01

Remote handling equipment is required to achieve the processing of highly radioactive, post reactor, fuel for the melt-dilute process, which will convert high enrichment uranium fuel elements into lower enrichment forms for subsequent disposal. The melt-dilute process combines highly radioactive enriched uranium fuel elements with deleted uranium and aluminum for inductive melting and inductive stirring steps that produce a stable aluminum/uranium ingot of low enrichment

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)

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

International Nuclear Information System (INIS)

Morrell, Douglas

2011-01-01

The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

Target-fueled nuclear reactor for medical isotope production

Science.gov (United States)

Coats, Richard L.; Parma, Edward J.

2017-06-27

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels

International Nuclear Information System (INIS)

Olsen, A.R.; Judkins, R.R.

1979-12-01

The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O 2 fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required

Processing of FRG high-temperature gas-cooled reactor fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements

International Nuclear Information System (INIS)

Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Drake, R.N.

1981-11-01

The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects of gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment development section of the agreement, both FRG mixed uranium/ thorium and low-enriched uranium fuel spheres have been processed in the Department of Energy-sponsored cold pilot plant for high-temperature gas-cooled reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles suitable for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated certain modifications to the US HTGR fuel burining process necessary for FRG fuel treatment. Results of the tests will be used in the design of a US/FRG joint prototype headend facility for HTGR fuel

Fast reactor fuel reprocessing. An Indian perspective

International Nuclear Information System (INIS)

Natarajan, R.; Raj, Baldev

2005-01-01

The Department of Atomic Energy (DAE) envisioned the introduction of Plutonium fuelled fast reactors as the intermediate stage, between Pressurized Heavy Water Reactors and Thorium-Uranium-233 based reactors for the Indian Nuclear Power Programme. This necessitated the closing of the fast reactor fuel cycle with Plutonium rich fuel. Aiming to develop a Fast Reactor Fuel Reprocessing (FRFR) technology with low out of pile inventory, the DAE, with over four decades of operating experience in Thermal Reactor Fuel Reprocessing (TRFR), had set up at the India Gandhi Center for Atomic Research (IGCAR), Kalpakkam, R and D facilities for fast reactor fuel reprocessing. After two decades of R and D in all the facets, a Pilot Plant for demonstrating FRFR had been set up for reprocessing the FBTR (Fast Breeder Test Reactor) spent mixed carbide fuel. Recently in this plant, mixed carbide fuel with 100 GWd/t burnup fuel with short cooling period had been successfully reprocessed for the first time in the world. All the challenging problems encountered had been successfully overcome. This experience helped in fine tuning the designs of various equipments and processes for the future plants which are under construction and design, namely, the DFRP (Demonstration Fast reactor fuel Reprocessing Plant) and the FRP (Fast reactor fuel Reprocessing Plant). In this paper, a comprehensive review of the experiences in reprocessing the fast reactor fuel of different burnup is presented. Also a brief account of the various developmental activities and strategies for the DFRP and FRP are given. (author)

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)

Fissile fuel doubling time characteristics for reactor lifetime fuel logistics

International Nuclear Information System (INIS)

Heindler, M.; Harms, A.A.

1978-01-01

The establishment of nuclear fuel requirements and their efficient utilization requires a detailed knowledge of some aspects of fuel dynamics and processing during the reactor lifetime. It is shown here that the use of the fuel stockpile inventory concept can serve effectively for this fuel management purpose. The temporal variation of the fissile fuel doubling time as well as nonequilibrium core conditions are among the characteristics which thus become more evident. These characteristics - rather than a single figure-of-merit - clearly provide an improved description of the expansion capacity and/or fuel requirements of a nuclear reactor energy system

Waste removal in pyrochemical fuel processing for the Integral Fast Reactor

International Nuclear Information System (INIS)

Ackerman, J.P.; Johnson, T.R.; Laidler, J.J.

1994-01-01

Electrorefining in a molten salt electrolyte is used in the Integral Fast Reactor fuel cycle to recover actinides from spent fuel. Processes that are being developed for removing the waste constituents from the electrorefiner and incorporating them into the waste forms are described in this paper. During processing, halogen, chalcogen, alkali, alkaline earth, and rare earth fission products build up in the molten salt as metal halides and anions, and fuel cladding hulls and noble metal fission products remain as metals of various particle sizes. Essentially all transuranic actinides are collected as metals on cathodes, and are converted to new metal fuel. After processing, fission products and other waste are removed to a metal and a mineral waste form. The metal waste form contains the cladding hulls, noble metal fission products, and (optionally) most rare earths in a copper or stainless steel matrix. The mineral waste form contains fission products that have been removed from the salt into a zeolite or zeolite-derived matrix

Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

Energy Technology Data Exchange (ETDEWEB)

Wegst, Ulrike G.K. [Dartmouth College, Hanover, NH (United States). Thayer School of Engineering; Allen, Todd [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States)

2014-04-07

Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

International Nuclear Information System (INIS)

Wegst, Ulrike G.K.; Sridharan, Kumar

2014-01-01

Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

Recent BWR fuel management reactor physics advances

International Nuclear Information System (INIS)

Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

1982-01-01

Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

Development of fuel cycle technology for molten-salt reactor systems

International Nuclear Information System (INIS)

Uhlir, J.

2006-01-01

Full text: Full text: The Molten-Salt Reactor (MSR) represents one of promising advanced reactor type assigned to the GEN IV reactor systems. It can be operated either as thorium breeder within the Th -133U fuel cycle or as actinide transmuter incinerating transuranium fuel. Essentially the main advantage of MSR comes out from the prerequisite, that this reactor type should be directly connected with the 'on-line' reprocessing of circulating liquid (molten-salt) fuel. This principle should allow very effective extraction of freshly constituted fissile material (233U). Besides, the on-line fuel salt clean up is necessary within a long run to keep the reactor in operation. As a matter of principle, it permits to clear away typical reactor poisons like xenon, krypton, lanthanides etc. and possibly also other products of burned plutonium and transmuted minor actinides. The fuel salt clean up technology should be linked with the fresh MSR fuel processing to continuously refill the new fuel (thorium or transuranics) into the reactor system. On the other hand, the technologies of fresh transuranium molten-salt fuel processing from the current LWR spent fuel and of the on-line reprocessing of MSR fuel represent two killing points of the whole MSR technology, which have to be successfully solved before MSR deployment in the future. There are three main pyrochemical partitioning techniques proposed for processing and/or reprocessing of MSR fuel: Fluoride volatilization processes, Molten salt / liquid metal extraction processes and Electrochemical separation processes. Two of them - Fluoride Volatility Method and Electrochemical separation process from fluoride media are under development in the Nuclear Research Institute Rez pic. R and D in the field of Fluoride Volatility Method is concentrated to the development and verification of experimental semi-pilot technology for LWR spent fuel reprocessing, which may result in a product the form and composition of which might be

Waste management in IFR [Integral Fast Reactor] fuel cycle

International Nuclear Information System (INIS)

Johnson, T.R.; Battles, J.E.

1991-01-01

The fuel cycle of the Integral Fast Reactor (IFR) has important potential advantage for the management of high-level wastes. This sodium-cooled, fast reactor will use metal fuels that are reprocessed by pyrochemical methods to recover uranium, plutonium, and the minor actinides from spent core and blanket fuel. More than 99% of all transuranic (TRU) elements will be recovered and returned to the reactor, where they are efficiently burned. The pyrochemical processes being developed to treat the high-level process wastes are capable of producing waste forms with low TRU contents, which should be easier to dispose of. However, the IFR waste forms present new licensing issues because they will contain chloride salts and metal alloys rather than glass or ceramic. These fuel processing and waste treatment methods can also handle TRU-rich materials recovered from light-water reactors and offer the possibility of efficiently and productively consuming these fuel materials in future power reactors

Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Sease, J.D.; Primm, R.T. III; Miller, J.H.

2007-01-01

The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the

Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Sease, J.D.; Primm, R.T. III; Miller, J.H.

2007-09-30

The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are

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.)

Fuel deposits, chemistry and CANDU® reactor operation

International Nuclear Information System (INIS)

Roberts, J.G.

2014-01-01

'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU® reactor, the first being the Nuclear Power Demonstration - 2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channelled to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5. The difference being that during 'hot conditioning' of CANDU® heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

Fossil fuel furnace reactor

Science.gov (United States)

Parkinson, William J.

1987-01-01

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

Pretreated Landfill Gas Conversion Process via a Catalytic Membrane Reactor for Renewable Combined Fuel Cell-Power Generation

Directory of Open Access Journals (Sweden)

Zoe Ziaka

2013-01-01

Full Text Available A new landfill gas-based reforming catalytic processing system for the conversion of gaseous hydrocarbons, such as incoming methane to hydrogen and carbon oxide mixtures, is described and analyzed. The exit synthesis gas (syn-gas is fed to power effectively high-temperature fuel cells such as SOFC types for combined efficient electricity generation. The current research work is also referred on the description and design aspects of permreactors (permeable reformers carrying the same type of landfill gas-reforming reactions. Membrane reactors is a new technology that can be applied efficiently in such systems. Membrane reactors seem to perform better than the nonmembrane traditional reactors. The aim of this research includes turnkey system and process development for the landfill-based power generation and fuel cell industries. Also, a discussion of the efficient utilization of landfill and waste type resources for combined green-type/renewable power generation with increased processing capacity and efficiency via fuel cell systems is taking place. Moreover, pollution reduction is an additional design consideration in the current catalytic processors fuel cell cycles.

Fuel Development For Gas-Cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

M. K. Meyer

2006-06-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

Energy Technology Data Exchange (ETDEWEB)

NONE

1985-07-01

Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements.

Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

International Nuclear Information System (INIS)

1985-01-01

Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements

Risk assessment of a fusion-reactor fuel-processing system

International Nuclear Information System (INIS)

Bruske, S.Z.; Holland, D.F.

1983-07-01

The probabilistic risk assessment (PRA) methodology provides a means to systematically examine the potential for accidents that may result in a release of hazardous materials. This report presents the PRA for a typical fusion reactor fuel processing system. The system used in the analysis is based on the Tritium Systems Test Assembly, which is being operated at the Los Alamos National Laboratory. The results of the evaluation are presented in a probability-consequence plot that describes the probability of various accidental tritium release magnitudes

Safety considerations in the fast reactor fuel cycle

International Nuclear Information System (INIS)

Baker, A.R.; Burton, W.R.; Taylor, H.A.

1977-01-01

The fuel cycle safety problems for fast reactors, as compared with thermal reactors, are enhanced by the higher fissile content and heat rating of the fuel. Additionally recycling leads to the build up of substantial isotopes which contribute to the alpha and neutron hazards. The plutonium arisings in a nuclear power reactor programme extending into the next century are discussed. A requirement is to be able to return the product plutonium to a reactor about 9 months after the end of irradiation and it is anticipated that progress will be made slowly towards this fuel cycle, having regard to the necessity for maintaining safe and reliable operations. Consideration of the steps in the fuel cycle has indicated that it will be best to store the irradiated fuel on the reactor sites while I131 decays and decay heat falls before transporting and a suitable transport flask is being developed. Reprocessing development work is aimed at the key area of fuel breakdown, the inter-relation of the fuel characteristics on the dissolution of the plutonium and a solvent extract cycle leading to a product suitable for a co-located fabrication plant. Because of the high activity of recycled fuel it is considered that fabrication must move to a fully remote operation as is already the case for reprocessing, and a gel precipitation process producing a vibro compacted fuel is under development for this purpose. The waste streams from the processing plants must be minimised, processed for recovery of plutonium where applicable and then conditioned so that the final products released from the processing cycle are acceptable for ultimate disposal. The safety aspects reviewed cover protection of operators, containment of radioactive materials, criticality and regulation of discharges to the environment

Fuel transfer system for a nuclear reactor

International Nuclear Information System (INIS)

Katz, L.R.; Marshall, J.R.; Desmarchais, W.E.

1977-01-01

Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

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

Future fuel cycle and reactor strategies

International Nuclear Information System (INIS)

Meneley, D.A.

1999-01-01

Within the framework of the 1997 IAEA Symposium 'Future Fuel Cycle and Reactor Strategies Adjusting to New Realities', Working Group No.3 produced a Key Issues paper addressing the title of the symposium. The scope of the Key Issues paper included those factors that are expected to remain or become important in the time period from 2015 to 2050, considering all facets of nuclear energy utilization from ore extraction to final disposal of waste products. The paper addressed the factors influencing the choice of reactor and fuel cycle. It then addressed the quantitatively largest category of reactor types expected to be important during the period; that is, thermal reactors burning uranium and plutonium fuel. The fast reactor then was discussed both as a stand-alone technology and as might be used in combination with thermal reactors. Thorium fuel use was discussed briefly. The present paper includes of a digest of the Key Issues Paper. Some comparisons arc made between the directions suggested in that paper and those indicated by the Abstracts of this Technical Committee Meeting- Recommendations are made for work which might be undertaken in the short and medium time frames, to ensure that fuel cycle technologies and processes established by the year 2050 will support the continuation of nuclear energy applications in the long term. (author)

Nuclear fuel, with emphasis on its utilization in pressurized water reactor

International Nuclear Information System (INIS)

Khazaneh, R.; Roshanzamir, M.

1997-01-01

Production processes of nuclear fuel on one hand and using nuclear fuels in reactors, particularly PWR Type reactors on the other hand is investigated. The first chapter reviews the relationship between fuel and reactors; The principals of reactor physics in relation with fuel are described shortly. The second chapter reviews uranium exploration and extraction as well as production of uranium concentrate and uranium dioxides. The third chapter is specified to the different procedures of uranium enrichment. In the fourth chapter, processing of uranium dioxide powder and fuel pellet is described. In the fifth chapter fabrication of fuel rod and fuel assemblies is explained thoroughly. The sixth chapter devoted to the different phenomena which occur ed in fuel structure and can during operational time of reactor; damage to fuel rods and developing theoretical models to describe these phenomena and analysis of fuel structure. The seventh chapter discusses how fuel rods are to be experimented during fabrication, operation and development of technology. The eighth chapter explains different fuels such as uranium compounds and mixed oxide fuel of uranium Gadolinium and uranium plutonium and the process of fabrication of zircaloy. In the tenth chapter, fuel reprocessing is investigated and the difficulties of developing this technology is referred

Proposed high throughput electrorefining treatment for spent N- Reactor fuel

International Nuclear Information System (INIS)

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

1996-01-01

A high-throughput electrorefining process is being adapted to treat spent N-Reactor fuel for ultimate disposal in a geologic repository. Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the type of fragmentation necessary to provide fuel segments suitable for this process. Based on these tests, a conceptual design was produced of a plant-scale electrorefiner. In this design, the diameter of an electrode assembly is about 1.07 m (42 in.). Three of these assemblies in an electrorefiner would accommodate a 3-metric-ton batch of N-Reactor fuel that would be processed at a rate of 42 kg of uranium per hour

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

Cermet fuel reactors

International Nuclear Information System (INIS)

Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

1987-09-01

Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

Process and equipment for locating defective fuel rods of a reactor fuel element

International Nuclear Information System (INIS)

Jester, A.; Honig, H.

1977-01-01

By this equipment, well-known processes for determining defective fuel rods of a reactor fuel element are improved in such a fashion that defective fuel rods can be located individually, so that it is possible to replace them. The equipment consists of a cylindrical test vessel open above, which accommodates the element to be tested, so that an annular space is left between the latter's external circumference and the wall of the vessel, and so that the fuel rods project above the vessel. A bell in the shape of a frustrum of a cone is inverted over the test vessel, which has an infra-red measuring equipment at a certain distance above the tops of the fuel rods. The fuel element to be tested together with the test vessel and hood are immersed in a basin full of water, which displaces water by means of gas from the hood. The post-shutdown heat increases the temperature in the water space of the test vessel, which is stabilised at 100 0 C. In each defective fuel rod the water which has penetrated the defective fuel rod previously, or does so now, starts to boil. The steam rising in the fuel rod raises the temperature of the defective fuel rod compared to all the sound ones. The subsequent measurement easily determines this. Where one can expect interference with the measurement by appreciable amounts of gamma rays, the measuring equipment is removed from the path of radiation by mirror deflection in a suitably shaped measuring hood. (FW) [de

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

Qualification process of dispersion fuels in the IEAR1 research reactor

International Nuclear Information System (INIS)

Domingos, D.B.; Silva, A.T.; Silva, J.E.R.

2010-01-01

Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a miniplate irradiation device (MID) to be placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 -Al dispersion fuels, LEU type (19,9% of 235 U) with uranium densities of, respectively, 3.0 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and TWODB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer code LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. This paper also presents a system designed for fuel swelling evaluation. The determination of the fuel swelling will be performed by means of the fuel miniplate thickness measurements along the irradiation time. (author)

Back-end of the research reactor fuel cycle

International Nuclear Information System (INIS)

Gruber, Gehard J.

1996-01-01

This paper outlines the status of topics and issues related to: (1) Research Reactor Spent Nuclear Fuel Return to the U.S., including policy, shipments and ports of entry, management sites, fees, storage technologies, contracts, actual shipment, and legal process, (2) UKAEA: MTR Spent Nuclear Fuel Reprocessing, (3) COGEMA: MTR Spent Nuclear Fuel Reprocessing, and (4) Intermediate Storage + Direct Disposal for Research Reactors. (author)

Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

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

1993-01-01

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

Reprocessing of fast neutron reactor fuel

International Nuclear Information System (INIS)

Bourgeois, M.

1981-05-01

A PUREX process specially adapted to fast neutron reactor fuels is employed. The results obtained indicate that the aqueous process can be applied to this type of fuel: almost 10 years operation at the AT 1 plant which processes fuel from RAPSODIE; the good results obtained at the MARCOULE pilot plant on large batches of reference fuels. The CEA is continuing its work to transfer this technology onto an industrial scale. Industrial prototypes and the launching of the TOR (traitement d'oxydes rapides) project will facilitate this transfer. In 1984, it is expected that fast fuels will be able to be processed on a significant scale and that supplementary R and D facilities will be available [fr

Nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Sakurai, Shungo; Ogiya, Shunsuke.

1990-01-01

In a fuel assembly, if the entire fuels comprise mixed oxide fuels, reactivity change in cold temperature-power operation is increased to worsen the reactor shutdown margin. The reactor shutdown margin has been improved by increasing the burnable poison concentration thereby reducing the reactivity of the fuel assembly. However, since unburnt poisons are present at the completion of the reactor operation, the reactivity can not be utilized effectively to bring about economical disadvantage. In view of the above, the reactivity change between lower temperature-power operations is reduced by providing a non-boiling range with more than 9.1% of cross sectional area at the inside of a channel at the central portion of the fuel assembly. As a result, the amount of the unburnt burnable poisons is decreased, the economy of fuel assembly is improved and the reactor shutdown margin can be increase. (N.H.)

The thermodynamics of pyrochemical processes for liquid metal reactor fuel cycles

International Nuclear Information System (INIS)

Johnson, I.

1987-01-01

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

Alternative fuels for the French fast breeder reactors programme

International Nuclear Information System (INIS)

Bailly, H.; Bernard, H.; Mansard, B.

1989-01-01

French fast breeder reactors use mixed oxide as reference fuel. A great deal of experience has been gained in the behaviour and manufacture of oxide fuel, which has proved to be the most suitable fuel for future commercial breeder reactors. However, France is maintaining long-term alternative fuels programme, in order to be in a position to satisfy eventually new future reactor design and operational requirements. Initially, the CEA in France developed a carbide-based, sodium-bonded fuel designed for a high specific power. The new objective of the alternative fuels programme is to define a fuel which could replace the oxide without requiring any significant changes to the operating conditions, fuel cycle processes or facilities. The current program concentrates on a nitride-based, helium-bonded fuel, bearing in mind the carbide solution. The paper describes the main characteristics required, the manufacturing process as developed, the inspection methods, and the results obtained. Present indications are that the industrial manufacture of mixed nitride is feasible and that production costs for nitride and oxide fuels would be not significantly different. (author) 8 refs., 2 figs

Research reactor de-fueling and fuel shipment

International Nuclear Information System (INIS)

Ice, R.D.; Jawdeh, E.; Strydom, J.

1998-01-01

Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures

Fast breeder reactor fuel reprocessing in France

International Nuclear Information System (INIS)

Bourgeois, M.; Le Bouhellec, J.; Eymery, R.; Viala, M.

1984-08-01

Simultaneous with the effort on fast breeder reactors launched several years ago in France, equivalent investigations have been conducted on the fuel cycle, and in particular on reprocessing, which is an indispensable operation for this reactor. The Rapsodie experimental reactor was associated with the La Hague reprocessing plant AT1 (1 kg/day), which has reprocessed about one ton of fuel. The fuel from the Phenix demonstration reactor is reprocessed partly at the La Hague UP2 plant and partly at the Marcoule pilot facility, undergoing transformation to reprocess all the fuel (TOR project, 5 t/y). The fuel from the Creys Malville prototype power plant will be reprocessed in a specific plant, which is in the design stage. The preliminary project, named MAR 600 (50 t/y), will mobilize a growing share of the CEA's R and D resources, as the engineering needs of the UP3 ''light water'' plant begins to decline. Nearly 20 tonnes of heavy metals irradiated in fast breeder reactors have been processed in France, 17 of which came from Phenix. The plutonium recovered during this reprocessing allowed the power plant cycle to be closed. This power plant now contains approximately 140 fuel asemblies made up with recycled plutonium, that is, more than 75% of the fuel assemblies in the Phenix core

Progress and status of the integral fast reactor (IFR) fuel cycle development

International Nuclear Information System (INIS)

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

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. The Integral Fast Reactor (IFR) fuel cycle, is based on the use of a metallic fuel alloy (U-Pu-Zr) that permits use of an innovative method for processing of spent fuel. This method, a combination of pyrometallurgical and electrochemical processes, has been termed pyroprocessing. It offers the advantages of a simple, compact processing system and limited volumes of stabilized high-level wastes. This translates to an economically viable system that is likely to receive favorable public response, particularly when combined with the other attributes of the Integral Fast Reactor. Substantial progress has been made in the development of the IFR pyroprocessing method. A comprehensive demonstration of the process will soon begin at the Argonne National Laboratory Idaho site, using spent fuel from the EBR-II reactor. An important advantage of the IFR is its ability to recycle fuel in the process of power generation, extending fuel resources by a considerable amount and assuring the continued viability of nuclear power stations by reducing dependence on external fuel supplies. Pyroprocessing is the means whereby the recycle process is accomplished. It can also be applied to the recovery of fuel constituents from spent fuel generated in the process of operation of conventional light water reactor power plants, offering the means to recover the valuable fuel resources remaining in that material

Nuclear reactor using fuel sphere for combustion and fuel spheres for breeding

International Nuclear Information System (INIS)

Yamashita, Kiyonobu.

1995-01-01

The present invention concerns a pebble bed-type reactor which can efficiently convert parent nuclides to fission nuclides. Fuel spheres for combustion having fission nuclides as main fuels, and fuel spheres for breeding having parent nuclides as main fuels are used separately, in the pebble bed-type reactor. According to the present invention, fuel spheres for breeding can be stayed in a reactor core for a long period of time, so that parent nuclides can be sufficiently converted into fission nuclides. In addition, since fuel spheres for breeding are loaded repeatedly, the amount thereof to be used is reduced. Therefore, the amount of the fuel spheres for breeding is small even when they are re-processed. On the other hand, since the content of the fission nuclides in the fuel spheres for breeding is not great, they can be put to final storage. This is attributable that although the fuel spheres for breeding contain fission nuclides generated by conversion, the fission nuclides are annihilated by nuclear fission reactions at the same time with the generation thereof. (I.S.)

Fuel deposits, chemistry and CANDU reactor operation

International Nuclear Information System (INIS)

Roberts, J.G.

2013-01-01

'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU reactor, the first being the Nuclear Power Demonstration-2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channel led to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5, and subsequently utilized for each CANDU unit since. The difference being that during 'hot conditioning' of CANDU heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

Preparations for the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Lineberry, M.J.; Phipps, R.D.

1989-01-01

Modifications to the Hot Fuel Examination Facility-South (HFEF/S) have been in progress since mid-1988 to ready the facility for demonstration of the unique Integral Fast Reactor (IFR) pyroprocess fuel cycle. This paper updates the last report on this subject to the American Nuclear Society and describes the progress made in the modifications to the facility and in fabrication of the new process equipment. The IFR is a breeder reactor, which is central to the capability of any reactor concept to contribute to mitigation of environmental impacts of fossil fuel combustion. As a fast breeder, fuel of course must be recycled in order to have any chance of an economical fuel cycle. The pyroprocess fuel cycle, relying on a metal alloy reactor fuel rather than oxide, has the potential to be economical even at small-scale deployment. Establishing this quantitatively is one important goal of the IFR fuel cycle demonstration

Cermet fuel reactors

International Nuclear Information System (INIS)

Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.

1987-01-01

Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. Key features of the cermet fueled reactor design are 1) the ability to achieve very high coolant exit temperatures, and 2) thermal shock resistance during rapid power changes, and 3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, there is a potential for achieving a long operating life because of 1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and 2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core

Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

International Nuclear Information System (INIS)

Weaver, K.D.; Sterbentz, J.; Meyer, M.; Lowden, R.; Hoffman, E.; Wei, T.Y.C.

2004-01-01

The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO 2 ) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

Uranium-plutonium fuel for fast reactors

International Nuclear Information System (INIS)

Antipov, S.A.; Astafiev, V.A.; Clouchenkov, A.E.; Gustchin, K.I.; Menshikova, T.S.

1996-01-01

Technology was established for fabrication of MOX fuel pellets from co-precipitated and mechanically blended mixed oxides. Both processes ensure the homogeneous structure of pellets readily dissolvable in nitric acid upon reprocessing. In order to increase the plutonium charge in a reactor-burner a process was tested for producing MOX fuel with higher content of plutonium and an inert diluent. It was shown that it is feasible to produce fuel having homogeneous structure and the content of plutonium up to 45% mass

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-07-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Plant-scale anodic dissolution of unirradiated N-Reactor fuel

International Nuclear Information System (INIS)

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

1995-01-01

Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electrorefiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electrorefiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electrorefiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour

Revisit of analytical methods for the process and plant control analyses during reprocessing of fast reactor fuels

International Nuclear Information System (INIS)

Subba Rao, R.V.

2016-01-01

CORAL (COmpact facility for Reprocessing of Advanced fuels in Lead cell) is an experimental facility for demonstrating the reprocessing of irradiated fast reactor fuels discharged from the Fast Breeder Test Reactor (FBTR). The objective of the reprocessing plant is to achieve nuclear grade plutonium and uranium oxides with minimum process waste volumes. The process flow sheet for the reprocessing of spent Fast Reactor Fuel consists of Transport of spent fuel, Chopping, Dissolution, Feed conditioning, Solvent Extraction cycle, Partitioning Cycle and Re-conversion of Plutonium nitrate and uranium nitrate to respective oxides. The efficiency and performance of the plant to achieve desired objective depends on the analyses of various species in the different steps adopted during reprocessing of fuels. The analytical requirements in the plant can be broadly classified as 1. Process control Analyses (Analyses which effect the performance of the plant- PCA); 2. Plant control Analyses (Analyses which indicates efficiency of the plant-PLCA); 3. Nuclear Material Accounting samples (Analyses which has bearing on nuclear material accounting in the plant - NUMAC) and Quality control Analyses (Quality of the input bulk chemicals as well as products - QCA). The analytical methods selected are based on the duration of analyses, precision and accuracies required for each type analytical requirement classified earlier. The process and plant control analyses requires lower precision and accuracies as compared to NUMAC analyses, which requires very high precision accuracy. The time taken for analyses should be as lower as possible for process and plant control analyses as compared to NUMAC analyses. The analytical methods required for determining U and Pu in process and plant samples from FRFR will be different as compared to samples from TRFR (Thermal Reactor Fuel Reprocessing) due to higher Pu to U ratio in FRFR as compared TRFR and they should be such that they can be easily

Licensing process characteristics of Small Modular Reactors and spent nuclear fuel repository

Energy Technology Data Exchange (ETDEWEB)

Söderholm, Kristiina, E-mail: kristiina.soderholm@fortum.com [Fortum Power (Finland); Tuunanen, Jari, E-mail: jari.tuunanen@fortum.com [Fortum Power (Finland); Amaba, Ben, E-mail: baamaba@us.ibm.com [IBM Complex Systems (United States); Bergqvist, Sofia, E-mail: sofia.bergqvist@se.ibm.com [IBM Rational Software (Sweden); Lusardi, Paul, E-mail: plusardi@nuscalepower.com [NuScale Power (United States)

2014-09-15

Highlights: • We examine the licensing process challenges of modular nuclear facilities. • We compare the features of Small Modular Reactors and spent nuclear fuel repository. • We present the need of nuclear licensing simplification. • Part of the licensing is proposed to be internationally applicable. • Systems engineering and requirements engineering benefits are presented. - Abstract: This paper aims to increase the understanding of the licensing processes characteristics of Small Modular Reactors (SMR) compared with licensing of spent nuclear fuel repository. The basis of the SMR licensing process development lies in licensing processes used in Finland, France, the UK, Canada and the USA. These countries have been selected for this study because of their various licensing processes and recent actions in the new NPP construction. Certain aspects of the aviation industry licensing process have also been studied and selected practices have been investigated as possibly suitable for use in nuclear licensing. Suitable features for SMR licensing are emphasized and suggested. The licensing features of the spent nuclear fuel deep repository along with similar features of SMR licensing are discussed. Since there are similar types of challenges of lengthy licensing time frames, as well as modular features to be taken into account in licensing, these two different nuclear industry fields can be compared. The main SMR features to take into account in licensing are: • Standardization of the design. • Modularity. • Mass production. • Serial construction. Modularity can be divided into two different categories: the first category is simply a single power plant unit constructed of independently engineered modules (e.g. construction process for Westinghouse AP-1000 NPP) and the second one a power plant composed of many reactor modules, which are manufactured in factories and installed as needed (e.g. NuScale Power SMR design). The deep underground repository

Reactor transients tests for SNR fuel elements in HFR reactor

International Nuclear Information System (INIS)

Plitz, H.

1989-01-01

In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

Detection system for location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing

Energy Technology Data Exchange (ETDEWEB)

Liu, Hongbing, E-mail: liuhb07@mails.tsinghua.edu.cn [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); Du, Dong, E-mail: dudong@tsinghua.edu.cn [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); Huang, An; Chang, Baohua; Han, Zandong [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); He, Ayada [Shanghai Electric Power Generation Group Shanghai Generator Works, Shanghai 200240 (China)

2016-08-15

Highlights: • A detection system for locations of pebbles transported in pipes is introduced. • The detection system is based on vibration signal processing, which is original. • The characteristics of the vibration signals of the pipe are analyzed. • The experiment shows that the detection results are accurate. • The research provides an important basis for the design of the reactor. - Abstract: Pebble-bed high temperature gas-cooled reactors have many advantages such as inherent safety, high efficiency, etc., and have been considered as a candidate for Generation IV nuclear reactors. During the operation of the reactor, there are thousands of fuel pebbles transported in the pipes outside the core by gravity and helium flow. The pattern of the pipes which consist of straight and arc sections is very complex. When a fuel pebble is transported, it will constantly collide with the pipes, especially in the arc sections. The collisions will lead to the vibration of the pipes. This paper aims to provide a detection system for the location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing. Before the reactor is running, the system acquires the vibration signals of several key sections by sensors. Then the frequency characteristics of the signals are obtained by joint time–frequency analysis. When the reactor is running, the system detects the signals and processes them based on their frequency characteristics in real time. According to the results of the processing, the system can correctly judge whether the fuel pebble has passed through the section and records the time of the passing. The experiment validates the accuracy and reliability of the detection results. In this way, the operational condition of the reactor can be monitored so that the normal running of the reactor can be ensured. Additionally, the detection data are of great significance to the evaluation and optimization of the reactor performance

Detection system for location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing

International Nuclear Information System (INIS)

Liu, Hongbing; Du, Dong; Huang, An; Chang, Baohua; Han, Zandong; He, Ayada

2016-01-01

Highlights: • A detection system for locations of pebbles transported in pipes is introduced. • The detection system is based on vibration signal processing, which is original. • The characteristics of the vibration signals of the pipe are analyzed. • The experiment shows that the detection results are accurate. • The research provides an important basis for the design of the reactor. - Abstract: Pebble-bed high temperature gas-cooled reactors have many advantages such as inherent safety, high efficiency, etc., and have been considered as a candidate for Generation IV nuclear reactors. During the operation of the reactor, there are thousands of fuel pebbles transported in the pipes outside the core by gravity and helium flow. The pattern of the pipes which consist of straight and arc sections is very complex. When a fuel pebble is transported, it will constantly collide with the pipes, especially in the arc sections. The collisions will lead to the vibration of the pipes. This paper aims to provide a detection system for the location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing. Before the reactor is running, the system acquires the vibration signals of several key sections by sensors. Then the frequency characteristics of the signals are obtained by joint time–frequency analysis. When the reactor is running, the system detects the signals and processes them based on their frequency characteristics in real time. According to the results of the processing, the system can correctly judge whether the fuel pebble has passed through the section and records the time of the passing. The experiment validates the accuracy and reliability of the detection results. In this way, the operational condition of the reactor can be monitored so that the normal running of the reactor can be ensured. Additionally, the detection data are of great significance to the evaluation and optimization of the reactor performance

Ceramics as nuclear reactor fuels

International Nuclear Information System (INIS)

Reeve, K.D.

1975-01-01

Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

Research reactor spent fuel management in Argentina

International Nuclear Information System (INIS)

Audero, M.A.; Bevilacqua, A.M.; Mehlich, A.M.; Novara, O.

2002-01-01

The research reactor spent fuel (RRSF) management strategy will be presented as well as the interim storage experience. Currently, low-enriched uranium RRSF is in wet interim storage either at reactor site or away from reactor site in a centralized storage facility. High-enriched uranium RRSF from the centralized storage facility has been sent to the USA in the framework of the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The strategy for the management of the RRSF could implement the encapsulation for interim dry storage. As an alternative to encapsulation for dry storage some conditioning processes are being studied which include decladding, isotopic dilution, oxidation and immobilization. The immobilized material will be suitable for final disposal. (author)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

Calculation device for fuel power history in BWR type reactors

International Nuclear Information System (INIS)

Sakagami, Masaharu.

1980-01-01

Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

Fuel transfer cask concept design for reactor TRIGA PUSPATI (RTP)

International Nuclear Information System (INIS)

Ahmad Nabil Ab Rahim; Phongsakorn Prak; Tonny Lanyau; Mohd Fazli Zakaria

2010-01-01

Reactor Triga PUSPATI (RTP) has been operated since 1982 till now. For such long period, the organization feels the need to upgrade the power from 1 MW to 3 MW which involved changing new fuels. Spent fuels will be stored in a Spent Fuel Pool. The process of transferring spent fuels into Spent Fuels Pool required a fuel transfer cask. This paper discussed the design concept for the fuel transfer cast which is essential equipment for reactor upgrading mission. (author)

Fuel exchanger in FBR type reactor

International Nuclear Information System (INIS)

Shinden, Kazuhiko; Tanaka, Osamu.

1990-01-01

The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

Reactor fuel charging equipment

International Nuclear Information System (INIS)

Wade, Elman.

1977-01-01

In many types of reactor fuel charging equipment, tongs or a grab, attached to a trolley, housed in a guide duct, can be used for withdrawing from the core a selected spent fuel assembly or to place a new fuel assembly in the core. In these facilities, the trolley may have wheels that roll on rails in the guide duct. This ensures the correct alignment of the grab, the trolley and fuel assembly when this fuel assembly is being moved. By raising or lowering such a fuel assembly, the trolley can be immerged in the coolant bath of the reactor, whereas at other times it can be at a certain level above the upper surface of the coolant bath. The main object of the invention is to create a fuel handling apparatus for a sodium cooled reactor with bearings lubricated by the sodium coolant and in which the contamination of these bearings is prevented [fr

Plant-scale anodic dissolution of unirradiated N-Reactor fuel

International Nuclear Information System (INIS)

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

1995-01-01

Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electro-refiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electro-refiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electro-refiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour. (author)

Hydrogen production with fully integrated fuel cycle gas and vapour core reactors

International Nuclear Information System (INIS)

Anghaie, S.; Smith, B.

2004-01-01

This paper presents results of a conceptual design study involving gas and vapour core reactors (G/VCR) with a combined scheme to generate hydrogen and power. The hydrogen production schemes include high temperature electrolysis as well as two dominant thermochemical hydrogen production processes. Thermochemical hydrogen production processes considered in this study included the calcium-bromine process and the sulphur-iodine processes. G/VCR systems are externally reflected and moderated nuclear energy systems fuelled by stable uranium compounds in gaseous or vapour phase that are usually operated at temperatures above 1500 K. A gas core reactor with a condensable fuel such as uranium tetrafluoride (UF 4 ) or a mixture of UF 4 and other metallic fluorides (BeF 2 , LiF, KF, etc.) is commonly known as a vapour core reactor (VCR). The single most relevant and unique feature of gas/vapour core reactors is that the functions of fuel and coolant are combined into one. The reactor outlet temperature is not constrained by solid fuel-cladding temperature limits. The maximum fuel/working fluid temperature in G/VCR is only constrained by the reactor vessel material limits, which is far less restrictive than the fuel clad. Therefore, G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Gas and vapour fuel reactors feature very low fuel inventory and fully integrated fuel cycle that provide for exceptional sustainability and safety characteristics. With respect to fuel utilisation, there is no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to the flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from completely burner to breeder is achievable. The continuous recycling of fuel in G/VCR systems allow for complete burning of actinides without removing and reprocessing of the fuel. The only waste products at the back

EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

Directory of Open Access Journals (Sweden)

S. M. Dmitriev

2016-01-01

Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

Power from plutonium: fast reactor fuel

International Nuclear Information System (INIS)

Bishop, J.F.W.

1981-01-01

Points of similarity and of difference between fast reactor fuel and fuels for AGR and PWR plants are established. The flow of uranium and plutonium in fast and thermal systems is also mentioned, establishing the role of the fast reactor as a plutonium burner. A historical perspective of fast reactors is given in which the substantial experience accumulated in test and prototype is indicated and it is noted that fast reactors have now entered the commercial phase. The relevance of the data obtained in the test and prototype reactors to the behaviour of commercial fast reactor fuel is considered. The design concepts employed in fuel are reviewed, including sections on core support styles, pin support and pin detail. This is followed by a discussion of current issues under the headings of manufacture, performance and reprocessing. This section includes a consideration of gel fuel, achievable burn-up, irradiation induced distortions and material choices, fuel form, and fuel failure mechanisms. Future development possibilities are also discussed and the Paper concludes with a view on the logic of a UK fast reactor strategy. (U.K.)

Radionuclide release from research reactor spent fuel

Energy Technology Data Exchange (ETDEWEB)

Curtius, H., E-mail: h.curtius@fz-juelich.de [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany); Kaiser, G.; Mueller, E.; Bosbach, D. [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany)

2011-09-01

Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO{sub 2} fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in {sup 235}U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO{sub 2}-fuel (LWR fuel, enrichment in {sup 235}U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Juelich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl{sub 2}-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl{sub x}-Al and U{sub 3}Si{sub 2}-Al) was studied in 400 mL MgCl{sub 2}-rich salt brine in the presence of Fe{sup 2+} under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH){sub 3}(s) and Eu(OH){sub 3}(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

Unified fuel elements development for research reactors

International Nuclear Information System (INIS)

Vatulin, A.; Stetsky, Y.; Dobrikova, I.

1998-01-01

Square cross-section rod type fuel elements have been developed for russian pool-type research reactors. new fuel elements can replace the large nomenclature of tubular fuel elements with around, square and hexahedral cross-sections and to solve a problem of enrichment reduction. the fuel assembly designs with rod type fuel elements have been developed. The overall dimensions of existing the assemblies are preserved in this one. the experimental-industrial fabricating process of fuel elements, based on a joint extrusion method has been developed. The fabricating process has been tested in laboratory conditions, 150 experimental fuel element samples of the various sizes were produced. (author)

Construction of a sputtering reactor for the coating and processing of monolithic U-Mo nuclear fuel

International Nuclear Information System (INIS)

Schmid, Wolfgang

2011-01-01

In the presented thesis sputter deposition was used for the first time to coat monolithic U-Mo nuclear fuel foils with diffusion inhibitive materials. The intention of these coatings is to prevent the formation of an interdiffusion layer between U-Mo and Al cladding during the use of the fuel. A small sputtering reactor was built, in which the method was tested and processing parameters were investigated. In parallel a larger sputtering reactor was constructed, that allows to coat full size monolithic U-Mo nuclear fuel foils and was used to test an industrial application of the technique. As a result a method based on sputter deposition and erosion can be presented, that allows to clean as well as to coat the surface of monolithic U-Mo nuclear fuel foils in excellent quality. It can be included at any time into the manufacturing chain for U-Mo fuel elements, which is currently being developed.

Optimization of a fuel bundle within a CANDU supercritical water reactor

International Nuclear Information System (INIS)

Schofield, M.E.

2009-01-01

The supercritical water reactor is one of six nuclear reactor concepts being studied under the Generation IV International Forum. Generation IV nuclear reactors will improve the metrics of economics, sustainability, safety and reliability, and physical protection and proliferation resistance over current nuclear reactor designs. The supercritical water reactor has specific benefits in the areas of economics, safety and reliability, and physical protection. This work optimizes the fuel composition and bundle geometry to maximize the fuel burnup, and minimize the surface heat flux and the form factor. In optimizing these factors, improvements can be achieved in the areas of economics, safety and reliability of the supercritical water reactor. The WIMS-AECL software was used to model a fuel bundle within a CANDU supercritical water reactor. The Gauss' steepest descent method was used to optimize the above mentioned factors. Initially the fresh fuel composition was optimized within a 43-rod CANFLEX bundle and a 61-rod bundle. In both the 43-rod and 61-rod bundle scenarios an online refuelling scheme and non-refuelling scheme were studied. The geometry of the fuel bundles was then optimized. Finally, a homogeneous mixture of thorium and uranium fuel was studied in a 60-rod bundle. Each optimization process showed definitive improvements in the factors being studied, with the most significant improvement being an increase in the fuel burnup. The 43-rod CANFLEX bundle was the most successful at being optimized. There was little difference in the final fresh fuel content when comparing an online refuelling scheme and non-refuelling scheme. Through each optimization scenario the ratio of the fresh fuel content between the annuli was a significant determining cause in the improvements in the factors being optimized. The geometry optimization showed that improvement in the design of a fuel bundle is indeed possible, although it would be more advantageous to pursue it

Asymptotic estimation of reactor fueling optimal strategy

International Nuclear Information System (INIS)

Simonov, V.D.

1985-01-01

The problem of improving the technical-economic factors of operating. and designed nuclear power plant blocks by developino. internal fuel cycle strategy (reactor fueling regime optimization), taking into account energy system structural peculiarities altogether, is considered. It is shown, that in search of asymptotic solutions of reactor fueling planning tasks the model of fuel energy potential (FEP) is the most ssuitable and effective. FEP represents energy which may be produced from the fuel in a reactor with real dimensions and power, but with hypothetical fresh fuel supply, regime, providing smilar burnup of all the fuel, passing through the reactor, and continuous overloading of infinitely small fuel portion under fule power, and infinitely rapid mixing of fuel in the reactor core volume. Reactor fuel run with such a standard fuel cycle may serve as FEP quantitative measure. Assessment results of optimal WWER-440 reactor fresh fuel supply periodicity are given as an example. The conclusion is drawn that with fuel enrichment x=3.3% the run which is 300 days, is economically justified, taking into account that the cost of one energy unit production is > 3 cop/KW/h

Development of alternative fuel for pressurized water reactors

International Nuclear Information System (INIS)

Cardoso, P.E.; Ferreira, R.A.N.; Ferraz, W.B.; Lameiras, F.S.; Santos, A.; Assis, G. de; Doerr, W.O.; Wehner, E.L.

1984-01-01

The utilization of alternative fuel cycles in Pressurized Water Reactors (PWR) such as Th/U and Th/Pu cycles can permit a better utilization of uranium reserves without the necessity of developing new power reactor concepts. The development of the technology of alternative fuels for PWR is one of the objectives of the 'Program on Thorium Utilization in Pressurized Water Reactors' carried out jointly by Empresas Nucleares Brasileiras S.A. (NUCLEBRAS), through its Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) and by German institutions, the Julich Nuclear Research Center (KFA), the Kraftwerk Union A.G. (KWU) and NUKEM GmbH. This paper summarizes the results so far obtained in the fuel technology. The development of a fabrication process for PWR fuel pellets from gel-microspheres is reported as well as the design, the specification, and the fabrication of prototype fuel rods for irradiation tests. (Author) [pt

Process for changing fuel elements of a water-cooled nuclear reactor

International Nuclear Information System (INIS)

Fleischmann, R.; Rau, P.

1986-01-01

In order to change fuel elements, a water-filled duct can be installed between the rector pressure vessel and a space for accommodating the fuel elements. The fuel elements are transported there under water by a fuelling machine. The duct is installed as watertight connection closed on all sides between the reactor pressure vessel and a fuel element transport container brought close to it. The fuelling machine works in this duct. (orig./HP) [de

A preliminary assessment of thorium as a fuel for thermal reactors

International Nuclear Information System (INIS)

Duret, M.F.; Halsall, M.J.

1965-08-01

This report presents a preliminary survey of the contributing factors to total fuel cost, for a variety of potentially economic fuel cycles. U 238 and Th 232 are considered as fertile materials with enrichments of U 235 , U 233 and plutonium. The total fuel cost comprises: (a) a fuel supply cost which is derived from a manufacturing cost and the fuel burnup, (b) as spent fuel credit, and (c) inventory charges. The interrelation of these three factors and their dependence on reactor flux, fuel burnup, the heat rating of the fuel and the interest rate are considered for each combination of fertile and fissile material. Results show inventory charges, spent fuel credit (hence processing costs) and fabrication costs have an important effect on the fuelling cost. In addition, using highly rated thorium-based fuels will affect reactor capital costs. Further, more detailed studies of reactor design, fuel design and fuel management schemes are now required to verify the attraction of using thorium as a reactor fuel. (author)

Fuel assembly and reactor core

International Nuclear Information System (INIS)

Aoyama, Motoo; Koyama, Jun-ichi; Uchikawa, Sadao; Bessho, Yasunori; Nakajima, Akiyoshi; Maruyama, Hiromi; Ozawa, Michihiro; Nakamura, Mitsuya.

1990-01-01

The present invention concerns fuel assemblies charged in a BWR type reactor and the reactor core. The fuel assembly comprises fuel rods containing burnable poisons and fuel rods not containing burnable poisons. Both of the highest and the lowest gadolinia concentrations of the fuel rods containing gadolinia as burnable poisons are present in the lower region of the fuel assembly. This can increase the spectral shift effect without increasing the maximum linear power density. (I.N.)

Method of reactor fueling

International Nuclear Information System (INIS)

Saito, Toshiro.

1983-01-01

Purpose: To decrease the cost and shorten the working time by saving fueling neutron detectors and their components. Method: Incore drive tubes for the neutron source range monitor (SRM) and intermediate range monitor (IRM) are disposed respectively within in a reactor core and a SRM detector assembly is inserted to the IRM incore drive tube which is most nearest to the neutron source upon reactor fueling. The reactor core reactivity is monitored by the SRM detector assembly. The SRM detector asesembly inserted into the IRM drive tube is extracted at the time of charging fuels up to the frame connecting the SRM and, thereafter, IRM detection assembly is inserted into the IRM drive tube and the SRM detector assembly is inserted into the SRM drive tube respectively for monitoring the reactor core. (Sekiya, K.)

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Gaseous fuel reactors for power systems

Science.gov (United States)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

Actinide recycle potential in the integral fast reactor (IFR) fuel cycle

International Nuclear Information System (INIS)

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

1991-01-01

In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The use of metallic fuel in the IFR allows a radically improved fuel cycle technology. Based on the recent IFR process development, a preliminary assessment has been made to investigate the feasibility of further adapting pyrochemical processes to directly extract actinides from LWR spent fuel. The results of this assessment indicate very promising potential and two most promising flowsheet options have been identified for further research and development. This paper also summarizes current thinking on the rationale for actinide recycle, its ramifications on the geologic repository and the current high-level waste management plans, and the necessary development programs

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Butterfield, C.E.; Waite, E.

1982-01-01

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

Fast reactor fuel design and development

International Nuclear Information System (INIS)

Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

1977-01-01

Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

Parametric studies on the fuel salt composition in thermal molten salt breeder reactors

International Nuclear Information System (INIS)

Nagy, K.; Kloosterman, J.L.; Lathouwers, D.; Van der Hagen, T.H.J.J.

2008-01-01

In this paper the salt composition and the fuel cycle of a graphite moderated molten salt self-breeder reactor operating on the thorium cycle is investigated. A breeder molten salt reactor is always coupled to a fuel processing plant which removes the fission products and actinides from the core. The efficiency of the removal process(es) has a large influence on the breeding capacity of the reactor. The aim is to investigate the effect on the breeding ratio of several parameters such as the composition of the molten salt, moderation ratio, power density and chemical processing. Several fuel processing strategies are studied. (authors)

Experience on Russian military origin plutonium conversion into fast reactor nuclear fuel

International Nuclear Information System (INIS)

Grachev, A.F.; Skiba, O.V.; Bychkov, A.V.; Mayorshin, A.A.; Kisly, V.A.; Bobrov, D.A.; Osipenko, A.G.; Babikov, L.G.; Mishinev, V.B.

2001-01-01

According to the Concept of Russian Minatom on military plutonium excess utilization, the State Scientific Center of Russian Federation ''Research Institute of Atomic Reactors'' (Dimitrovgrad) has begun study on possibility of technological processing of the metal military plutonium into MOX fuel. The Program and the stages of its realization are submitted in the paper. During 1998-2000 the first stage of the Program was fulfilled and 50 kg of military origin metallic plutonium was converted to MOX fuel for the BOR-60 and BN-600 reactor. The plutonium conversion into MOX fuel is carried out under the original technology developed by SSC RIAR. It includes pyro-electrochemical process for production of fuel on the domestic equipment with the subsequent fuel pins manufacturing for the fast reactors by the vibro-packing method. The produced MOX fuel is purified from alloy additives (Ga) and corresponds to the vibro-packed fuel standard for fast reactors. The fuel pins manufacturing for BOR-60 and BN-600 reactors are carried out by the vibro-packing method on a standard procedure, which is used in SSC RIAR more than 20 years. (author)

Fuel assembly for FBR type reactor and reactor core thereof

International Nuclear Information System (INIS)

Kobayashi, Kaoru.

1998-01-01

The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

Fuel management at the Petten high flux reactor

International Nuclear Information System (INIS)

Thijssen, P.J.M.

1999-01-01

Several years ago the shipment of spent fuel of the High Flux Reactor (HFR) at Petten has come to a standstill resulting in an ever growing stock of fuel elements that are labelled 'fully burnt up'. Examination of those elements showed that a reasonably number of them have a relatively high 235 U mass left. A reactor physics analysis showed that the use of such elements in the peripheral core zone allows the loading of four instead of five fresh fuel elements in many cycle cores. For the assessment of safety and performance parameters of HFR cores a new calculational tool is being developed. It is based on AEA Technology's Reactor physics code suite Winfrith Improved Multigroup Scheme (WIMS). NRG produced pre- and post-processing facilities to feed input data into WIMS's 2D transport code CACTUS and to extract relevant parameters from the output. The processing facilities can be used for many different types of application. (author)

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.

Cost targets for at-reactor spent fuel rod consolidation

International Nuclear Information System (INIS)

Macnabb, W.V.

1985-01-01

The high-level nuclear waste management system in the US currently envisions the disposal of spent fuel rods that have been removed from their assemblies and reconfigured into closely packed arrays. The process of fuel rod removal and packaging, referred to as rod consolidation, can occur either at reactors or at an integrated packaging facility, monitored retrievable storage (MRS). Rod consolidation at reactors results in cost savings down stream of reactors by reducing needs for additional storage, reducing the number of shipments, and reducing (eliminating, in the extreme) the amount of fuel handling and consolidation at the MRS. These savings accrue to the nuclear waste fund. Although private industry is expected to pay for at-reactor activities, including rod consolidation, it is of interest to estimate cost savings to the waste system if all fuel were consolidated at reactors. If there are savings, the US Department of Energy (DOE) may find it advantageous to pay for at-reactor rod consolidation from the nuclear waste fund. This paper assesses and compares the costs of rod consolidation at reactors and at the MRS in order to determine at what levels the former could be cost competitive with the latter

Shifting to non-explosive fuels for research reactors

International Nuclear Information System (INIS)

Gilinsky, Victor

1984-01-01

The RERTR program is not just an American program, it is an international program and it can succeed only with wide support and participation. Excellent work underway at research laboratories in several countries is making a vital contribution. The cooperative spirit shown by all participants is particularly gratifying. Some practical difficulties may be encountered with the safety regulatory agencies in different countries when modifying the reactors. The US NRC intends to demonstrate that conversion to low enriched fuel is not a difficult process by taking steps to enable domestic reactors operating in the United States to convert to low enriched fuels. A proposed regulation is being prepared limiting the use of highly enriched uranium in domestic reactors. In this connection, The US NRC will be prepared tp cooperate with reactor operators from other countries in the safety area as it relates to the conversion process

Use of freeze-casting in advanced burner reactor fuel design

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

Comparison of fuel assemblies in lead cooled fast reactors

Energy Technology Data Exchange (ETDEWEB)

Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G., E-mail: alejandria.peval@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

2016-09-15

This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

Comparison of fuel assemblies in lead cooled fast reactors

International Nuclear Information System (INIS)

Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G.

2016-09-01

This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

Waste disposal from the light water reactor fuel cycle

International Nuclear Information System (INIS)

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

1981-05-01

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

Fuel exchange device for FBR type reactor

International Nuclear Information System (INIS)

Onuki, Koji.

1993-01-01

The device of the present invention can provide fresh fuels with a rotational angle aligned with the direction in the reactor core, so that the fresh fuels can be inserted being aligned with apertures of the reactor core even if a self orientation mechanism should fail to operate. That is, a rotational angle detection means (1) detects the rotational angle of fresh fuels before insertion to the reactor core. A fuel rotational angle control means (2) controls the rotational angle of the fresh fuels by comparing the detection result of the means (1) and the data for the insertion position of the reactor core. A fuel rotation means (3) compensates the rotational angel of the fresh fuels based on the control signal from the means (2). In this way, when the fresh fuels are inserted to the reactor core, the fresh fuels set at the same angle as that for the aperture of the reactor core. Accordingly, even if the self orientation mechanism should not operate, the fresh fuels can be inserted smoothly. As a result, it is possible to save loss time upon fuel exchange and mitigate operator's burden during operation. (I.S.)

Actinide recycle potential in the Integral Fast Reactor (IFR) fuel cycle

International Nuclear Information System (INIS)

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

1990-01-01

In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The use of metallic fuel in the IFR allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. In the electrorefining operation, uranium and plutonium are selectively transported from an anode to a cathode, leaving impurity elements, mainly fission products, either in the anode compartment or in a molten salt electrolyte. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management, because these actinides are automatically recycled back into the reactor for in-situ burning. Based on the recent IFR process development, a preliminary assessment has also been made to investigate the feasibility of further adapting the pyrochemical processes to directly extract actinides from LWR spent fuel. The results of this assessment indicate very promising potential and two most promising flowsheet options have been identified for further research and development. This paper also summarizes current thinking on the rationale for actinide recycle, its ramifications on the geologic repository and the current high-level waste management plans, and the necessary development programs. 5 refs., 4 figs., 4 tabs

Fast reactor fuel reprocessing in the UK

International Nuclear Information System (INIS)

Allardice, R.H.; Williams, J.; Buck, C.

1977-01-01

Enriched uranium metal fuel irradiated in the Dounreay Fast Reactor has been reprocessed and refabricated in plants specifically designed for the purpose in the U.K. since 1961. Efficient and reliable fuel recycle is essential to the development of a plutonium based fast reactor system and the importance of establishing at an early stage fast reactor fuel reprocessing has been reinforced by current world difficulties in reprocessing high burn-up thermal reactor oxide fuel. In consequence, the U.K. has decided to reprocess irradiated fuel from the 250 MW(E) Prototype Fast Reactor as an integral part of the fast reactor development programme. Flowsheet and equipment development work for the small scale fully active demonstration plant have been carried out over the past 5 years and the plant will be commissioned and ready for active operation during 1977. In parallel, a comprehensive waste management system has been developed and installed. Based on this development work and the information which will arise from active operation of the plant a parallel development programme has been initiated to provide the basis for the design of a large scale fast reactor fuel reprocessing plant to come into operation in the late 1980s to support the projected U.K. fast reactor installation programme. The paper identifies the important differences between fast reactor and thermal reactor fuel reprocessing technologies and describes some of the development work carried out in these areas for the small scale P.F.R. fuel reprocessing operation. In addition, the development programme in aid of the design of a larger scale fast reactor fuel reprocessing plant is outlined and the current design philosophy is discussed

Status of the Integral Fast Reactor fuel cycle demonstration and waste management practices

International Nuclear Information System (INIS)

Benedict, R.W.; Goff, K.M.; McFarlane, H.F.

1994-01-01

Over the past few years, Argonne National Laboratory has been preparing for the demonstration of 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 and operations, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle, which will be demonstrated at Argonne-West in Idaho, employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The required facility modifications and process equipment for the demonstration are nearing completion. Their status and the results from initial fuel fabrication work, including the waste management aspects, are presented. Additionally, estimated compositions of the various process waste streams have been made, and characterization and treatment methods are being developed. The status of advanced waste processing equipment being designed and fabricated is described

Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

Energy Technology Data Exchange (ETDEWEB)

NONE

2009-06-15

facilities. - 3. Advances in Water Reactor Fuel Technology: Advances in fuel, rod, spacer grids, and assembly design; fuel processing and manufacturing; cladding and structural alloy development; MOX fuel design and manufacturing; advances in fuel pellet development; fuel design for improved thermal hydraulics, mechanical, and corrosion-resistant behavior; irradiation experience in test reactors. - 4. Concepts for Transportation and Interim Storage of Spent Fuels and Conditioned Waste (Shared with Global 2009): Industrial experience and ongoing developments. - 5. Innovative Fuel Design and Core Management: Future development and trends in fuel for the next thirty years; Goals and perspectives for nuclear fuel; Long term improvement in fissile material management; Use of composite material; Innovative microstructure and material under development; Future core management.

Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

International Nuclear Information System (INIS)

2009-06-01

Reactor Fuel Technology: Advances in fuel, rod, spacer grids, and assembly design; fuel processing and manufacturing; cladding and structural alloy development; MOX fuel design and manufacturing; advances in fuel pellet development; fuel design for improved thermal hydraulics, mechanical, and corrosion-resistant behavior; irradiation experience in test reactors. - 4. Concepts for Transportation and Interim Storage of Spent Fuels and Conditioned Waste (Shared with Global 2009): Industrial experience and ongoing developments. - 5. Innovative Fuel Design and Core Management: Future development and trends in fuel for the next thirty years; Goals and perspectives for nuclear fuel; Long term improvement in fissile material management; Use of composite material; Innovative microstructure and material under development; Future core management

Fuel assembly for FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki.

1995-01-01

Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

Simulated first operating campaign for the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Park, K.H.; Ackerman, J.P.

1993-01-01

This report discusses the Integral Fast Reactor (IFR) which is an innovative liquid-metal-cooled reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid-metal cooling to offer significant improvements in reactor safety, operation, fuel cycle-economics, environmental protection, and safeguards. Over the next few years, the IFR fuel cycle will be demonstrated at Argonne-West in Idaho. Spent fuel from the Experimental Breeder Reactor II (EBR-II) win be processed in its associated Fuel Cycle Facility (FCF) using a pyrochemical method that employs molten salts and liquid metals in an electrorefining operation. As part of the preparation for the fuel cycle demonstration, a computer code, PYRO, was developed at Argonne to model the electrorefining operation using thermodynamic and empirical data. This code has been used extensively to evaluate various operating strategies for the fuel cycle demonstration. The modeled results from the first operating campaign are presented. This campaign is capable of processing more than enough material to refuel completely the EBR-II core

Advanced research reactor fuel development

Energy Technology Data Exchange (ETDEWEB)

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

2000-05-01

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

A Nuclear Reactor and Chemical Processing Design for Production of Molybdenum-99 with Crystalline Uranyl Nitrate Hexahydrate Fuel

Science.gov (United States)

Stange, Gary Michael

Medical radioisotopes are used in tens of millions of procedures every year to detect and image a wide variety of maladies and conditions in the human body. The most widely-used diagnostic radioisotope is technetium-99m, a metastable isomer of technetium-99 that is generated by the radioactive decay of molybdenum-99. For a number of reasons, the supply of molybdenum-99 has become unreliable and the techniques used to produce it have become unattractive. This has spurred the investigation of new technologies that avoid the use of highly enriched uranium to produce molybdenum-99 in the United States, where approximately half of the demand originates. The first goal of this research is to develop a critical nuclear reactor design powered by solid, discrete pins of low enriched uranium. Analyses of single-pin heat transfer and whole-core neutronics are performed to determine the required specifications. Molybdenum-99 is produced directly in the fuel of this reactor and then extracted through a series of chemical processing steps. After this extraction, the fuel is left in an aqueous state. The second goal of this research is to describe a process by which the uranium may be recovered from this spent fuel solution and reconstituted into the original fuel form. Fuel recovery is achieved through a crystallization step that generates solid uranyl nitrate hexahydrate while leaving the majority of fission products and transuranic isotopes in solution. This report provides background information on molybdenum-99 production and crystallization chemistry. The previously unknown thermal conductivity of the fuel material is measured. Following this is a description of the modeling and calculations used to develop a reactor concept. The operational characteristics of the reactor core model are analyzed and reported. Uranyl nitrate crystallization experiments have also been conducted, and the results of this work are presented here. Finally, a process flow scheme for uranium

Thermal hydraulic analysis of gas-cooled reactors with annular fuel rods

International Nuclear Information System (INIS)

Han, Kyu Hyun; Chang, Soon Heung

2005-01-01

More than half of the world's energy is used in industrial processes and for heating applications which have hardly been touched by the nuclear industry. Nuclear power could be brought into a wide range of applications for industrial processes, provided that gas outlet temperatures of gascooled reactors are sufficiently high. The most limiting core design requirement which controls the core outlet temperature is the maximum acceptable fuel compact temperature. An innovative fuel design is required for a significant decrease in the fuel temperature. This study investigated the possibilities of implementing internally and externally cooled annular fuel rods in a gas-cooled reactor

Dose rate analyses for fast reactor fuel manufacturers

International Nuclear Information System (INIS)

Smith, R.C.; Strode, J.N.; Brackenbush, L.W.; Faust, L.G.

1976-01-01

An early appraisal of the radiation exposure situation in the fabrication of plutonium enriched mixed-oxide fuels for fast reactors is presented. Radiation data are presented on fuel process operations measured under actual operating conditions using plutonium containing up to 19 wt percent 240 Pu

Russian research reactor fuel return program starts shipping fuel to Russia

International Nuclear Information System (INIS)

Dedik, T.; Bolshinsky, I.; Krass, A.

2003-01-01

For almost four years the United States (U.S), the Russian Federation (R.F.), and the International Atomic Energy Agency (IAEA) have been discussing an initiative to return Soviet/Russian-origin research reactor fuel to the Russian Federation. In a series of bilateral and trilateral meetings in Vienna and Moscow, considerable progress has been made toward defining the Russian Research Reactor Fuel Return Program as well as obtaining the necessary technical data to facilitate the return. More than 20 research reactors in 17 countries that have Soviet- or Russian-supplied fuel have identified. Most of these reactors have stocks of both fresh and irradiated HEU fuel that must be carefully stored and managed for many years to come. On September 21, 2003 the Russian Research Reactor Fuel Return program shipped 14 kg of fresh Russian-origin HEU fuel from Romania to the nuclear fuel fabrication facility in Russia, which represented the beginning of the practical implementation of the program. (author)

Fuel assembly and reactor core

International Nuclear Information System (INIS)

Yuchi, Yoko; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro; Koyama, Jun-ichi.

1996-01-01

In a fuel assembly of a BWR type reactor, a region substantially containing burnable poison is divided into an upper region and a lower region having different average concentrations of burnable poison along a transverse cross section perpendicular to the axial direction. The ratio of burnable poison contents of both regions is determined to not more than 80%, and the average concentration of the burnable poison in the lower region is determined to not less than 9% by weight. An infinite multiplication factor at an initial stage of the burning of the fuel assembly is controlled effectively by the burnable poisons. Namely, the ratio of the axial power can be controlled by the distribution of the enrichment degree of uranium fuels and the distribution of the burnable poison concentration in the axial direction. Since the average enrichment degree of the reactor core has to be increased in order to provide an initially loaded reactor core at high burnup degree. Distortion of the power distribution in the axial direction of the reactor core to which fuel assemblies at high enrichment degree are loaded is flattened to improve thermal margin, to extend continuous operation period and increase a burnup degree upon take-out thereby improving fuel economy without worsening the reactor core characteristics of the initially loaded reactor core. (N.H.)

Facilities of fuel transfer for nuclear reactors

International Nuclear Information System (INIS)

Wade, E.E.

1977-01-01

This invention relates to sodium cooled fast breeder reactors. It particularly concerns facilities for the transfer of fuel assemblies between the reactor core and a fuel transfer area. The installation is simple in construction and enables a relatively small vessel to be used. In greater detail, the invention includes a vessel with a head, fuel assemblies housed in this vessel, and an inlet and outlet for the coolant covering these fuel assemblies. The reactor has a fuel transfer area in communication with this vessel and gear inside the vessel for the transfer of these fuel assemblies. These facilities are borne by the vessel head and serve to transfer the fuel assemblies from the vessel to the transfer area; whilst leaving the fuel assemblies completely immersed in a continuous mass of coolant. A passageway is provided between the vessel and this transfer area for the fuel assemblies. Facilities are provided for closing off this passageway so that the inside of the reactor vessel may be isolated as desired from this fuel transfer area whilst the reactor is operating [fr

MOX fuel fabrication, in reactor performance and improvement

International Nuclear Information System (INIS)

Vliet, J. van; Deramaix, P.; Nigon, J.L.; Fournier, W.

1998-01-01

In Europe, MOX fuel for light water reactors (LWRs) has first been manufactured in Belgium and Germany. Belgonucleaire (BN) loaded the first MOX assembly in the BR3 Pressurised Water Reactor (PWR) in 1963. In June 1998, more than 750 tHM LWR MOX fuel assemblies were manufactured on a industrial scale in Europe without any particular difficulty relating to fuel fabrication, reactor operation or fuel behaviour. So, today plutonium recycling through MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants. In this field, COGEMA and BELGONUCLEAIRE are the main actors by operating simultaneously three complete multidesign fuel production plants: MELOX plant (in Marcoule), CADARACHE plant and P0 plant (in Dessel, Belgium). Present MOX production capacity available to COGEMA and BN fits 175 tHM per year and is to be extended to reach about 325 tHM in the year 2000. This will represent 75% of the total MOX fabrication capacity in Europe. The industrial mastery and the high production level in MOX fabrication assured by high technology processes confer to these companies a large expertise for Pu recycling. This allows COGEMA and BN to be major actors in Pu-based fuels in the coming second nuclear era with advanced fuel cycles. (author)

BR2 Reactor: Irradiation of fuels

International Nuclear Information System (INIS)

Verwimp, A.

2005-01-01

Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

Integrating the fuel cycle at IFR [Integral Fast Reactor

International Nuclear Information System (INIS)

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

1992-01-01

During the past few years Argonne National Laboratory has been developing the Integral Fast Reactor (IFR), an advanced liquid metal reactor. Much of the IFR technology stems from Argonne National Laboratory's experience with the Experimental Breeder Reactors, EBR 1 and 2. The unique aspect of EBR 2 is its success with high-burnup metallic fuel. Irradiation tests of the new U-Pu-Zr fuel for the IFR have now reached a burnup level of 20%. The results to date have demonstrated excellent performance characteristics of the metallic fuel in both steady-state and off-normal operating conditions. EBR 2 is now fully loaded with the IFR fuel alloys and fuel performance data are being generated. In turn, metallic fuel becomes the key factor in achieving a high degree of passive safety in the IFR. These characteristics were demonstrated dramatically by two landmark tests conducted at EBR 2 in 1986: loss of flow without scram; and loss of heat sink without scram. They demonstrated that the combination of high heat conductivity of metallic fuel and thermal inertia of the large sodium pool can shut the reactor down during potentially severe accidents without depending on human intervention or the operation of active engineered components. The IFR metallic fuel is also the key factor in compact pyroprocessing. Pyroprocessing uses high temperatures, molten salt and metal solvents to process metal fuels. The result is suitable for fabrication into new fuel elements. Feasibility studies are to be conducted into the recycling of actinides from light water reactor spent fuel in the IFR using the pyroprocessing approach to extract the actinides (author)

Build-up and decay of fuel actinides in the fuel cycle of nuclear reactors

International Nuclear Information System (INIS)

Tasaka, Kanji; Kikuchi, Yasuyuki; Shindo, Ryuichi; Yoshida, Hiroyuki; Yasukawa, Shigeru

1976-05-01

For boiling water reactors, pressurized light-water reactors, pressure-tube-type heavy water reactors, high-temperature gas-cooled reactors, and sodium-cooled fast breeder reactors, uranium fueled and mixed-oxide fueled, each of 1000 MWe, the following have been studied: (1) quantities of plutonium and other fuel actinides built up in the reactor, (2) cooling behaviors of activities of plutonium and other fuel actinides in the spent fuels, and (3) activities of plutonium and other fuel actinides in the high-level reprocessing wastes as a function of storage time. The neutron cross section and decay data of respective actinide nuclides are presented, with their evaluations. For effective utilization of the uranium resources and easy reprocessing and high-level waste management, a thermal reactor must be fueled with uranium; the plutonium produced in a thermal reactor should be used in a fast reactor; and the plutonium produced in the blanket of a fast reactor is more appropriate for a fast reactor than that from a thermal reactor. (auth.)

Water reactor fuel element fabrication, with special emphasis on its effects on fuel performance

International Nuclear Information System (INIS)

1979-01-01

Full text: The performance of nuclear fuel has improved over the years and is now a minor cause of outages and of power limitations in nuclear power plants. On the other hand, an increasing number of countries are in the process of developing or implementing their own capability for manufacturing fuel elements. In this context, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) advised that a symposium be organized devoted to the relationship between fuel fabrication and performance The Czechoslovak Atomic Energy Commission agreed to co-operate in the organization of this symposium and to host it in Prague. Those factors which influence fuel fabrication requirements are now well ascertained: as little reactor primary circuit contamination as possible, the tendency to increased burnups, reactor manoeuverability to match power grid demands, the desirability of an autonomous fabrication capability. It is the general experience of fuel element suppliers that fuel quality and performance has increased over the years, the importance of quality assurance and process monitoring has been decisive in this respect The ever increasing mass-production aspect of nuclear fuel leads to some processing steps being revised and alternatives being developed. The relation between fabrication processes and fuel performance characteristics, although generally well perceived, are still the subject of a large amount of experiment and assessment in most countries, both industrial and developing This evidence is most encouraging; it means indeed that nuclear power, which is already amongst the cheapest and safest sources of energy, will continue to be improved. The performance of Zircaloy fuel cladding - presently the material used in most water reactors - is under particular consideration. Better understanding of this quite recent alloy will pave the way for broader fuel utilization limits in the future. The panel discussion, which noted some

Fuel Fabrication and Nuclear Reactors

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-02

The uranium from the enrichment plant is still in the form of UF₆. UF₆ is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF₆ is converted into UO₂ fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

CANDU reactor experience: fuel performance

International Nuclear Information System (INIS)

Truant, P.T.; Hastings, I.J.

1985-07-01

Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

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

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

International Nuclear Information System (INIS)

2000-07-01

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

Fuel handling system of nuclear reactor plants

International Nuclear Information System (INIS)

Faulstich, D.L.

1991-01-01

This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

Integrated scheme of long-term for spent fuel management of power nuclear reactors

International Nuclear Information System (INIS)

Ramirez S, J. R.; Palacios H, J. C.; Martinez C, E.

2015-09-01

After of irradiation of the nuclear fuel in the reactor core, is necessary to store it for their cooling in the fuel pools of the reactor. This is the first step in a processes series before the fuel can reach its final destination. Until now there are two options that are most commonly accepted for the end of the nuclear fuel cycle, one is the open nuclear fuel cycle, requiring a deep geological repository for the fuel final disposal. The other option is the fuel reprocessing to extract the plutonium and uranium as valuable materials that remaining in the spent fuel. In this study the alternatives for the final part of the fuel cycle, which involves the recycling of plutonium and the minor actinides in the same reactor that generated them are shown. The results shown that this is possible in a thermal reactor and that there are significant reductions in actinides if they are recycled into reactor fuel. (Author)

Status of power reactor fuel reprocessing in India

International Nuclear Information System (INIS)

Kansra, V.P.

1999-01-01

Spent fuel reprocessing in India started with the commissioning of the Trombay Plutonium Plant in 1964. This plant was intended for processing spent fuel from the 40 MWth research reactor CIRUS and recovering plutonium required for the research and development activities of the Indian Atomic Energy programme. India's nuclear energy programme aims at the recycle of plutonium in view of the limited national resources of natural uranium and abundant quantities of thorium. This is based on the approach which aims at separating the plutonium from the power reactor spent fuel, use it in the fast reactors to breed 233 U and utilise the 233 U generated to sustain a virtually endless source of power through thorium utilisation. The separated plutonium is also being utilised to fabricate MOX fuel for use in thermal reactors. Spent fuel treatment and extracting plutonium from it makes economic sense and a necessity for the Indian nuclear power programme. This paper describes the status and trends in the Indian programme for the reprocessing of power reactor fuels. The extraction of plutonium can also be seen as a far more positive approach to long-term waste management. The closed cycle approach visualised and pursued by the pioneers in the field is now steadily moving India towards the goal of a sustainable source of power through nuclear energy. The experience in building, operating and refurbishing the reprocessing facilities for uranium and thorium has resulted in acquiring the technological capability for designing, constructing, operating and maintaining reprocessing plants to match India's growing nuclear power programme. (author)

The Canadian research reactor spent fuel situation

International Nuclear Information System (INIS)

Ernst, P.C.

1996-01-01

This paper summarizes the present research reactor spent fuel situation in Canada. The research reactors currently operating are listed along with the types of fuel that they utilize. Other shut down research reactors contributing to the storage volume are included for completeness. The spent fuel storage facilities associated with these reactors and the methods used to determine criticality safety are described. Finally the current inventory of spent fuel and where it is stored is presented along with concerns for future storage. (author). 3 figs

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

Gas release from a failed fuel pin after reactor shut-down

International Nuclear Information System (INIS)

Pshenichnikov, B.V.

1975-01-01

A mathematical model of gassing from a hypothetical core fuel element in the active zone of a stopped water-moderated reactor was analysed to investigate the process of liberation of gaseous fission products from an unpressurized fuel element. A one-dimensional problem was obtained as a result of the accepted hypotheses. A fault was assumed to have occured during reactor operation; at the same time, a vapour-gas mixture was considered to be present under the envelope at reactor working pressure by the moment of stoppage. An approximative estimation was made of the retardation time of pressure balancing at the open end of the fuel element, and also of the amount of total gas remaining in the gap under the fuel element envelope after pressure drop in the reactor. Estimation of retardation time permitted to conclude that pressure in the nonhermetic fuel element envelope follows pressure fluctuation in the reactor in the course of cooling, the retardation time of pressure balancing outside and inside the fuel element lasting but a few seconds

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)

Plans for the development of the IFR [Integral Fast Reactor] fuel cycle

International Nuclear Information System (INIS)

Johnson, T.R.

1986-01-01

The Integral Fast Reactor (IFR) is a concept for a self-contained facility in which several sodium-cooled fast reactors of moderate size are located at the same site along with complete fuel-recycle and waste-treatment facilities. After the initial core loading with enriched uranium or plutonium, only natural or depleted uranium is shipped to the plant, and only wastes in final disposal forms are shipped out. The reactors have driver and blanket fuels of uranium-plutonium-zirconium alloys in stainless steel cladding. The use of metal alloy fuels is central to the IFR concept, contributing to the inherent safety of the reactor, the ease of reprocessing, and the relatively low capital and operating costs. Discharged fuels are recovered in a pyrochemical process that consists of two basic steps: an electrolytic process to separate fission products from actinides, and halide slagging to separate plutonium from uranium

Evolution of processing of GE fuel clad tubing for corrosion resistance in boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Williams, C.D. [GE Nuclear Energy, Wilmington, NC (United States); Adamson, R.B. [GE Nuclear Energy, Wilmington, NC (United States); Marlowe, M.O. [GE Nuclear Energy, Wilmington, NC (United States); Plaza-Meyer, E. [GE Nuclear Energy, Wilmington, NC (United States); Proebstle, R.A. [GE Nuclear Energy, Wilmington, NC (United States); White, D.W. [GE Nuclear Energy, Wilmington, NC (United States)

1996-05-01

The current modification of the primary GE in-process solution-quench heat treatment, an (alpha+beta) solution-quench carried out at a tube diameter requiring only two subsequent reduction and anneal cycles, is applicable to Zr barrier fuel clad tubing, to non-barrier fuel clad tubing, and to the TRICLAD tubing product. A combination of good in-reactor corrosion performance and degradation resistance is anticipated for these products, based on knowledge of metallurgical characteristics and supported by the demonstrated performance capability of the Zircaloy-2 materials used. (orig.)

Nuclear reactors and fuel cycle

International Nuclear Information System (INIS)

2014-01-01

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

Nuclear reactors and fuel cycle

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-07-01

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

Status of research reactor spent fuel world-wide

International Nuclear Information System (INIS)

Ritchie, I.G.

2004-01-01

Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialised and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author)

United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

International Nuclear Information System (INIS)

Morrell, Douglas

2008-01-01

The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room. Reactor

Reactor-specific spent fuel discharge projections, 1987-2020

International Nuclear Information System (INIS)

Walling, R.C.; Heeb, C.M.; Purcell, W.L.

1988-03-01

The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs

Water Reactor Fuel Performance Meeting 2008

International Nuclear Information System (INIS)

2008-10-01

This meeting contains articles of the Water Reactor Fuel Performance Meeting 2008 of Korean Nuclear Society, Atomic Energy Society of Japan, Chinese Nuclear Society, European Nuclear Society and American Nuclear Society. It was held on Oct. 19-23, 2008 in Seoul, Korea and subject of Meeting is 'New Clear' Fuel - A green energy solution. This proceedings is comprised of 5 tracks. The main topic titles of track are as follows: Advances in water reactor fuel technology, Fuel performance and operational experience, Transient fuel behavior and safety-related issues, Fuel cycle, spent fuel storage and transportations and Fuel modeling and analysis. (Yi, J. H.)

Reproduction of the RA reactor fuel element fabrication

International Nuclear Information System (INIS)

Novakovic, M.

1961-12-01

This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals [sr

Advanced spent fuel processing technologies for the United States GNEP programme

International Nuclear Information System (INIS)

Laidler, J.J.

2007-01-01

Spent fuel processing technologies for future advanced nuclear fuel cycles are being developed under the scope of the Global Nuclear Energy Partnership (GNEP). This effort seeks to make available for future deployment a fissile material recycling system that does not involve the separation of pure plutonium from spent fuel. In the nuclear system proposed by the United States under the GNEP initiative, light water reactor spent fuel is treated by means of a solvent extraction process that involves a group extraction of transuranic elements. The recovered transuranics are recycled as fuel material for advanced burner reactors, which can lead in the long term to fast reactors with conversion ratios greater than unity, helping to assure the sustainability of nuclear power systems. Both aqueous and pyrochemical methods are being considered for fast reactor spent fuel processing in the current US development programme. (author)

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

Space reactor fuels performance and development issues

International Nuclear Information System (INIS)

Wewerka, E.M.

1984-01-01

Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

Future fuel cycle development for CANDU reactors

International Nuclear Information System (INIS)

Hatcher, S.R.; McDonnell, F.N.; Griffiths, J.; Boczar, P.G.

1987-01-01

The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

Method of fueling for a nuclear reactor

International Nuclear Information System (INIS)

Igarashi, Takao.

1983-01-01

Purpose: To enable the monitoring of reactor power with sufficient accuracy, upon starting even without existence of neutron source in case of a low average burnup degree in the reactor core. Constitution: Each of fuel assemblies is charged such that neutron source region monitors for the start-up system in a reactor core neutron instrumentation system having nuclear fuel assemblies and a neutron instrumentation system are surrounded with 4 or 16 fuel assemblies of a low burnup degree. Then, the average burnup degree of the fuel assemblies surrounding the neutron source region monitors are increased than the reactor core burnup degree, whereby neutrons released from the peripheral fuels are increased, sufficient number of neutron counts can be obtained even with no neutron sources upon start-up and the reactor power can be monitored at a sufficient accuracy. (Sekiya, K.)

Fuel assemblies for nuclear reactor

International Nuclear Information System (INIS)

Nishi, Akihito.

1987-01-01

Purpose: To control power-up rate at the initial burning stage of new fuel assemblies due to fuel exchange in a pressure tube type power reactor. Constitution: Burnable poisons are disposed to a most portion of fuel pellets in a fuel assembly to such a low concentration as the burn-up rate changes with time at the initial stage of the burning. The most portion means substantially more than one-half part of the pellets and gadolinia is used as burn-up poisons to be dispersed and the concentration is set to less than about 0.2 %. Upon elapse of about 15 days after the charging, the burnable poisons are eliminated and the infinite multiplication factors are about at 1.2 to attain a predetermined power state. Since the power-up rate of the nuclear reactor fuel assembly is about 0.1 % power/hour and the power-up rate of the fuel assembly around the exchanged channel is lower than that, it can be lowered sufficiently than the limit for the power-up rate practiced upon reactor start-up thereby enabling to replace fuels during power operation. (Horiuchi, T.)

Inter renewal travelling wave reactor with rotary fuel columns

International Nuclear Information System (INIS)

Terai, Yuzo

2016-01-01

To realize the COP21 decision, this paper proposes Inter Renewal Travelling Wave Reactor that bear high burn-up rate 50% and product TRU fuel efficiently. The reactor is based on 4S Fast Reactor and has Reactor Fuel Columns as fuel assemblies that equalize temperature in the fuel assembly so that fewer structure is need to restrain thermal transformation. To equalize burn-up rate of all fuel assemblies in the reactor, each rotary fuel column has each motor-lifter. The rotary fuel column has two types (Cylinder type and Heat Pipe type using natrium at 15 kPa which supply high temperature energy for Ultra Super Critical power plant). At 4 years cycle all rotary fuel columns of the reactor are renewed by the metallurgy method (vacuum re-smelting) and TRU fuel is gotten from the water fuel. (author)

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

Removal of spent fuel from the TVR reactor for reprocessing and proposals for the RA reactor spent fuel handling

International Nuclear Information System (INIS)

Volkov, E.B.; Konev, V.N.; Shvedov, O.V.; Bulkin, S.Yu; Sokolov, A.V.

2002-01-01

The 2,5 MW heavy-water moderated and cooled research reactor TVR was located at the Moscow Institute for Theoretical and Experimental Physics site. In 1990 the final batch of spent nuclear fuel (SNF) from the TVR reactor was transported for reprocessing to Production Association (PA) 'Mayak'. This transportation of the SNF was a part of TVR reactor decommissioning. The special technology and equipment was developed in order to fulfill the preparation of TVR SNF for transportation. The design of the TVR reactor and the fuel elements used are similar to the design and fuel elements of the RA reactor. Two different ways of RA spent fuel elements for transportation to reprocessing plant are considered: in aluminum barrels, and in additional cans. The experience and equipment used for the preparing TVR fuel elements for transportation can help the staff of RA reactor to find the optimal way for these technical operations. (author)

Performance of metallic fuels in liquid-metal fast reactors

International Nuclear Information System (INIS)

Seidel, B.R.; Walters, L.C.; Kittel, J.H.

1984-01-01

Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

Fast-reactor fuel reprocessing in the United Kingdom

International Nuclear Information System (INIS)

Allardice, R.H.; Buck, C.; Williams, J.

1977-01-01

Enriched uranium metal fuel irradiated in the Dounreay Fast Reactor has been reprocessed and refabricated in plants specifically designed for the purpose in the United Kingdom since 1961. Efficient and reliable fuel recycle is essential to the development of a plutonium-based fast-reactor system, and the importance of establishing at an early stage fast-reactor fuel reprocessing has been reinforced by current world difficulties in reprocessing high-burnup thermal-reactor oxide fuel. The United Kingdom therefore decided to reprocess irradiated fuel from the 250MW(e) Prototype Fast Reactor (PFR) as an integral part of the fast reactor development programme. Flowsheet and equipment development work for the small-scale fully active demonstration plant has been carried out since 1972, and the plant will be commissioned and ready for active operation during 1977. In parallel, a comprehensive waste-management system has been developed and installed. Based on this development work and the information which will arise from active operation of the plant, a parallel development programme has been initiated to provide the basis for the design of a large-scale fast-reactor fuel-reprocessing plant to come into operation in the late 1980s to support the projected UK fast-reactor installation programme. The paper identifies the important differences between fast-reactor and thermal-reactor fuel-reprocessing technologies and describes some of the development work carried out in these areas for the small-scale PFR fuel-reprocessing operation. In addition, the development programme in aid of the design of a larger scale fast-reactor fuel-reprocessing plant is outlined and the current design philosophy discussed. (author)

Advanced fuel in the Budapest research reactor

International Nuclear Information System (INIS)

Hargitai, T.; Vidovsky, I.

1997-01-01

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

Dry process fuel performance technology development

International Nuclear Information System (INIS)

Kang, Kweon Ho; Kim, K. W.; Kim, B. K.

2006-06-01

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

Dry process fuel performance technology development

Energy Technology Data Exchange (ETDEWEB)

Kang, Kweon Ho; Kim, K. W.; Kim, B. K. (and others)

2006-06-15

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

An engineering design of reactor with NPP spent fuels

International Nuclear Information System (INIS)

Yuan Luzheng; Shen Feng; Yang Changjiang; Dai Changnian; Jin Huajin; Li Yulun

2005-01-01

Study has proven that it is of practical significance to design a reactor in suitable low parameters using the spent fuels of nuclear power plant. This kind of reactor will supply, safely and economically, a clean energy for desalination of sea- water and heating supply for city residents. Based on listing main problems required to be solved when designing a reactor in suitable low parameters directly using NPP spent fuels, a preliminary design scheme with engineering feasibility is given. Some significant efforts and attempts have been made for this scheme on its core structure and main processing systems design, adopting inherent safety characteristics to the full, making the reactor as a 'foolish type' one with easy operation, safe and reliable merit to the best. (authors)

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1984-01-01

The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

Evaluation of Metal-Fueled Surface Reactor Concepts

International Nuclear Information System (INIS)

Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Hiatt, Matthew T.; Amiri, Benjamin W.

2007-01-01

Surface fission power systems for use on the Moon and Mars may provide the first use of near-term reactor technology in space. Most near-term surface reactor concepts specify reactor temperatures <1000 K to allow the use of established material and power conversion technology and minimize the impact of the in-situ environment. Metal alloy fuels (e.g. U-10Zr and U-10Mo) have not traditionally been considered for space reactors because of high-temperature requirements, but they might be an attractive option for these lower temperature surface power missions. In addition to temperature limitations, metal fuels are also known to swell significantly at rather low fuel burnups (∼1 a/o), but near-term surface missions can mitigate this concern as well, because power and lifetime requirements generally keep fuel burnups <1 a/o. If temperature and swelling issues are not a concern, then a surface reactor concept may be able to benefit from the high uranium density and relative ease of manufacture of metal fuels. This paper investigates two reactor concepts that utilize metal fuels. It is found that these concepts compare very well to concepts that utilize other fuels (UN, UO2, UZrH) on a mass basis, while also providing the potential to simplify material safeguards issues

Irradiation behavior of metallic fast reactor fuels

International Nuclear Information System (INIS)

Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

1991-01-01

Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985

Evaluation of research reactor fuel reliability in support of regulatory requirements

International Nuclear Information System (INIS)

Sokolov, Eugene N.

2005-01-01

This standards, codes and practices survey is devoted to the problem of reliability of R and D especially research reactor fuel (RRF) performance-related processes. Regulatory R and D evaluations were based on one standard and just few of them provide correlation to other relative standards whereas synthetic process approach reflects actual status of particular R and D practices. Fuel performance regulatory parameters are based on quality standards. A reliability process-based method similar to PSA/FMEA is proposed to evaluate RRF performance- related parameters in terms of reactor safety. (author)

Evaluation of research reactor fuel reliability in support of regulatory requirements

Energy Technology Data Exchange (ETDEWEB)

Sokolov, Eugene N [Chalk River Laboratories, AECL, Chalk River, ON, K0J 1J0 (Canada)

2005-07-01

This standards, codes and practices survey is devoted to the problem of reliability of R and D especially research reactor fuel (RRF) performance-related processes. Regulatory R and D evaluations were based on one standard and just few of them provide correlation to other relative standards whereas synthetic process approach reflects actual status of particular R and D practices. Fuel performance regulatory parameters are based on quality standards. A reliability process-based method similar to PSA/FMEA is proposed to evaluate RRF performance- related parameters in terms of reactor safety. (author)

Nuclear reactor application for high temperature power industrial processes

International Nuclear Information System (INIS)

Dollezhal', N.A.; Zaicho, N.D.; Alexeev, A.M.; Baturov, B.B.; Karyakin, Yu.I.; Nazarov, E.K.; Ponomarev-Stepnoj, N.N.; Protzenko, A.M.; Chernyaev, V.A.

1977-01-01

This report gives the results of considerations on industrial heat and technology processes (in chemistry, steelmaking, etc.) from the point of view of possible ways, technical conditions and nuclear safety requirements for the use of high temperature reactors in these processes. Possible variants of energy-technological diagrams of nuclear-steelmaking, methane steam-reforming reaction and other processes, taking into account the specific character of nuclear fuel are also given. Technical possibilities and economic conditions of the usage of different types of high temperature reactors (gas cooled reactors and reactors which have other means of transport of nuclear heat) in heat processes are examined. The report has an analysis of the problem, that arises with the application of nuclear reactors in energy-technological plants and an evaluation of solutions of this problem. There is a reason to suppose that we will benefit from the use of high temperature reactors in comparison with the production based on high quality fossil fuel [ru

Preliminary study of the economics of enriching PWR fuel with a fusion hybrid reactor

International Nuclear Information System (INIS)

Kelly, J.L.

1978-09-01

This study is a comparison of the economics of enriching uranium oxide for pressurized water reactor (PWR) power plant fuel using a fusion hybrid reactor versus the present isotopic enrichment process. The conclusion is that privately owned hybrid fusion reactors, which simultaneously produce electrical power and enrich fuel, are competitive with the gaseous diffusion enrichment process if spent PWR fuel rods are reenriched without refabrication. Analysis of irradiation damage effects should be performed to determine if the fuel rod cladding can withstand the additional irradiation in the hybrid and second PWR power cycle. The cost competitiveness shown by this initial study clearly justifies further investigations

The integral fast reactor fuels reprocessing laboratory at Argonne National Laboratory, Illinois

International Nuclear Information System (INIS)

Wolson, R.D.; Tomczuk, Z.; Fischer, D.F.; Slawecki, M.A.; Miller, W.E.

1986-09-01

The processing of Integral Fast Reactor (IFR) metal fuel utilizes pyrochemical fuel reprocessing steps. These steps include separation of the fission products from uranium and plutonium by electrorefining in a fused salt, subsequent concentration of uranium and plutonium for reuse, removal, concentration, and packaging of the waste material. Approximately two years ago a facility became operational at Argonne National Laboratory-Illinois to establish the chemical feasibility of proposed reprocessing and consolidation processes. Sensitivity of the pyroprocessing melts to air oxidation necessitated operation in atmosphere-controlled enclosures. The Integral Fast Reactor Fuels Reprocessing Laboratory is described

Issues of high-burnup fuel for advanced nuclear reactors

International Nuclear Information System (INIS)

Belac, J.; Milisdoerfer, L.

2004-12-01

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

Development of tritium fuel processing system using electrolytic reactor for ITER

International Nuclear Information System (INIS)

Yamanishi, T.; Kawamura, Y.; Iwai, Y.

2001-01-01

The system composed of a palladium diffuser and an electrolytic reactor was proposed, and was developed for a Fuel Cleanup system of ITER. The performance of the system was studied in a stand-alone test in detail. A fuel simulation loop of ITER was constructed by connecting the developed Fuel Cleanup and Hydrogen Isotope Separation systems; and the function of each system in the loop was demonstrated. For the tritium recovery from the exhaust gas at He glow discharge cleaning of vacuum chamber of ITER, a cryogenic molecular sieve bed system was proposed and demonstrated. (author)

Development of tritium fuel processing system using electrolytic reactor for ITER

International Nuclear Information System (INIS)

Yamanishi, Toshihiko; Kawamura, Y.; Iwai, Y.

1999-01-01

The system composed of a palladium diffuser and an electrolytic reactor was proposed, and was developed for a Fuel Cleanup system of ITER. The performance of the system was studied in a stand-alone test in detail. A fuel simulation loop of ITER was constructed by connecting the developed Fuel Cleanup and Hydrogen Isotope Separation systems; and the function of each system in the loop was demonstrated. For the tritium recovery from the exhaust gas at He glow discharge cleaning of vacuum chamber of ITER, a cryogenic molecular sieve bed system was proposed and demonstrated. (author)

R and D on fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Subba Rao, R.V.; Vijaya Kumar, V.; Natarajan, R.

2012-01-01

Development of Fast Reactor Fuel Reprocessing technology, with low out of pile inventory, is carried out at the Indira Gandhi Centre for Atomic Research (IGCAR). Based on the successful R and D programme which addressed specific issues of fast reactor fuels, a pilot plant called CORAL was set up. This plant is operational since 2003 and several reprocessing campaigns with spent FBTR fuels of varying burnups have been carried out. Based on the valuable operating experience of CORAL, the design of demonstration fast reactor fuel reprocessing plant (DFRP) and the commercial reprocessing plant, FRP have been taken up. Concurrently R and D efforts are continuing for improving the process and equipment performance apart from reducing the waste volumes and the radiation exposures to the operating personnel. Some important R and D efforts are highlighted in the paper. Reducing the dissolution time is one of the vital area of investigation especially for the high plutonium bearing MOX fuels which are known to dissolve slowly. To address this as well as criticality issues, continuous dissolvers are being developed. Solvent extraction based process is employed for getting highly pure nuclear grade uranium and plutonium. In view of the lower cooling time the fission product activity in the spent fuel is higher, formulation of process flowsheet with reduced number of solvent extraction cycles to improve the decontamination of ruthenium and zirconium without the formation of second organic phase due to plutonium loading, is under investigation. Retention of plutonium in lean organic is another issue to be addressed as otherwise it would lead to further deterioration of the solvent on storage. Several reagents to effectively wash the lean solvent have been investigated and flowsheets have been formulated to recover the retained plutonium with minimum secondary wastes. Partitioning of uranium and plutonium is an important step and methods reported in the literature have inherent

Fast reactors with axial arrangement of oxide and metal fuels in the core

International Nuclear Information System (INIS)

Troyanov, M.F.; Ilyunin, V.G.; Matveev, V.I.; Murogov, V.M.; Proshkin, A.A.; Rudneva, V.Ya.; Shmelev, A.N.

1980-01-01

Problems of using metal fuel in fast reactor (FR) core are discussed Results are given of the calculation of two-dimentional (R-Z) FR version having a composed core with the combined usage of oxide and metal fuels having parameters close to optimal from the point of view of fuel breeding rate, an oxide subzone having increased enrichment and a decreased proper conversion ratio. A reactor is considered where metallic fuel elements are placed from the side of ''cold'' coolant inlet (400-480 deg C), and oxide fuel elements - in the region where the coolant has a higher temperature (500-560 deg C). It is shown that the new fuel breeding rate in such a reactor can be increased by 20-30% as compared with an oxide fuel reactor. Growth of the total conversion ratio is mainly stipulated with the increase of the inner conversion ratio of the core (CRC) which is important not only from the point of view of nuclear fuel breeding rate but also the optimization of the mode of powerful fast reactor operation with provision for the change in reactivity in the process of its continuous operation. The fact, that the core version under investigation has a CRC value slightly exceeding unit, stipulates considerably less reactivity change as compared with the oxide version in the process of the reactor operation and permits at a constant reactor control system power to significantly increase the time between reloadings and, therefore, to increase the NPP load factor which is of great importance both from the point of view of economy and the improvement of operation conditions as well as of reactor operation reliability. It is concluded on the base of the analysis of the results obtained that FRs with the combined usage of oxide and metal fuels having an increased specific load and increased conversion ratio as compared with the oxide fuel FRs provide a higher rate of development of the whole nuclear power balanced with respect to the fuel [ru

Aspects regarding the fuel management for PHWR nuclear reactors

International Nuclear Information System (INIS)

Dragusin, O.; Bobolea, A.; Voicu, A.

2001-01-01

Fuel management for PHWR nuclear reactors is completely different from the PWR reactors fuel management. PHWR reactor fuel loading procedures are repeated after an interval of time, as defined and specified in the project documentation, using a fuel machine that can be attached to the terminal fittings of horizontal pressure tubes while the reactor is a full power. Another aspect of fuel management policy is related to the possibility of bi-directional loading of the reactor, with the primary advantage of uniform and symmetrical characteristics. (authors)

Method of monitoring fuel-rod vibrations in a nuclear fuel reactor

International Nuclear Information System (INIS)

Kawamura, Makoto; Takai, Katsuaki.

1985-01-01

Purpose: To monitor the vibration modes of fuel rods continuously and on real time during operation of a PWR type nuclear reactor. Method: Vibrations of fuel rods during reactor operation are mainly caused by the lateral flow of coolants flowing through the gaps at the joints of reactor core buffle plates into a reactor core and fretting damages may possibly be caused to the fuel rod support portions due to the vibrations. In view of the above, self-powered detectors are disposed at a plurality of axial positions for the respective peripheral fuel assemblies in adjacent with the buffle plates and the detection signals from neutron detectors, that is, the fluctuations in neutrons are subjected to a frequency analysis during the operation period. The neutron detectors are disposed at the periphery of the reactor core, because the fuel assemblies disposed at the peripheral portion directly undergo the lateral flow from the joints of the buffle plates and vibrates most violently. Thus, the vibration situations can be monitored continuously, in a three demensional manner and on real time. (Moriyama, K.)

Progress of the DUPIC fuel compatibility analysis (I) - reactor physics

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Jeong, Chang Joon; Roh, Gyu Hong; Rhee, Bo Wook; Park, Jee Won

2003-12-01

Since 1992, the direct use of spent pressurized water reactor fuel in CANada Deuterium Uranium (CANDU) reactors (DUPIC) has been studied as an alternative to the once-through fuel cycle. The DUPIC fuel cycle study is focused on the technical feasibility analysis, the fabrication of DUPIC fuels for irradiation tests and the demonstration of the DUPIC fuel performance. The feasibility analysis was conducted for the compatibility of the DUPIC fuel with existing CANDU-6 reactors from the viewpoints of reactor physics, reactor safety, fuel cycle economics, etc. This study has summarized the intermediate results of the DUPIC fuel compatibility analysis, which includes the CANDU reactor physics design requirements, DUPIC fuel core physics design method, performance of the DUPIC fuel core, regional overpower trip setpoint, and the CANDU primary shielding. The physics analysis showed that the CANDU-6 reactor can accommodate the DUPIC fuel without deteriorating the physics design requirements by adjusting the fuel management scheme if the fissile content of the DUPIC fuel is tightly controlled.

Inspecting fuel pellets for nuclear reactor

International Nuclear Information System (INIS)

Wilks, R.S.; Sternheim, E.; Breakey, G.A.; Sturges, R.H.; Taleff, A.; Castner, R.P.

1982-01-01

An improved method of controlling the inspection, sorting and classifying of nuclear reactor fuel pellets, including a mechanical handling system and a computer controlled data processing system, is described. Having investigated the diameter, length, surface flaws and weights of the pellets, they are sorted accordingly and the relevant data are stored. (U.K.)

Fabrication of Fast Reactor Fuel Pins for Test Irradiations

Energy Technology Data Exchange (ETDEWEB)

Karsten, G. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Dippel, T. [Institute for Radiochemistry, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Laue, H. J. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

1967-09-15

An extended irradiation programme is being carried out for the fuel element development of the Karlsruhe fast breeder project. A very important task within the programme is the testing of plutonium-containing fuel pins in a fast-reactor environment. This paper deals with fabrication of such pins by our laboratories at Karlsruhe. For the fast reactor test positions at present envisaged a fuel with 15% plutonium and the uranium fully enriched is appropriate. Hie mixed oxide is both pelletized and vibro-compacted with smeared densities between 80 and 88% theoretical. The pin design is, for example, such that there are two gas plena at the top and bottom, and one blanket above the fuel with the fuel zone fitting to the test reactor core length. The specifications both for fuel and cladding have been adapted to the special purpose of a fast-breeder reactor - the outer dimensions, the choice of cladding and fuel types, the data used and the kind of tests outline the targets of the development. The fuel fabrication is described in detail, and also the powder line used for vibro-compaction. The source materials for the fuel are oxalate PuO{sub 2} and UO{sub 2} from the UF{sub 6} process. The special problems of mechanical mixing and of plutonium homogeneity have been studied. The development of the sintering technique and grain characteristics for vibratory compactive fuel had to overcome serious problems in order to reach 82-83% theoretical. The performance of the pin fabrication needed a major effort in welding, manufacturing of fits and decontamination of the pin surfaces. This was a stimulation for the development of some very subtle control techniques, for example taking clear X-ray photographs and the tube testing. In general the selection of tests was a special task of the production routine. In conclusion the fabrication of the pins resulted in valuable experiences for the further development of fast reactor fuel elements. (author)

Reactor fueling system

International Nuclear Information System (INIS)

Hattori, Noriaki; Hirano, Haruyoshi.

1983-01-01

Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

Pilot studies of an extraction process for reprocessing of spent fuel from fast reactors: Hardware and process details of extractor selection

International Nuclear Information System (INIS)

Anisimov, V.I.; Pavlovich, V.B.; Smetanin, E.Ya.; Glazunov, N.V.; Shklyar, L.I.; Dubrovskii, V.G.; Serov, A.V.; Zakharkin, B.S.; Konorchenko, V.D.; Korotkov, I.A.; Neumoev, N.V.; Renard, E.V.

1992-01-01

While acknowledging the bold and persistent efforts of U.S. and Russian specialists to develop the concept of pyrochemical reprocessing of spent nuclear fuel from fast reactors on remote-controlled equipment for removal of actinides from the fission products one should recognize that the tasks of reprocessing such fuel can be handled only by using water-extraction technology, especially since the known Purex process continues to be improved to the point that a single-cycle scheme may be developed. This article presents results of pilot studies conducted in hot cells using multistage extractors in continuous counterflow operation; data on various extractor types used in reprocessing spent mixed oxide nuclear fuel; advantages and disadvantages of centrifugal and pulsed column extractor; comparison of column-type and centrifugal extractors; and extraction process

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1981-01-01

An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

Fuel processing for PEM fuel cells: transport and kinetic issues of system design

Science.gov (United States)

Zalc, J. M.; Löffler, D. G.

In light of the distribution and storage issues associated with hydrogen, efficient on-board fuel processing will be a significant factor in the implementation of PEM fuel cells for automotive applications. Here, we apply basic chemical engineering principles to gain insight into the factors that limit performance in each component of a fuel processor. A system consisting of a plate reactor steam reformer, water-gas shift unit, and preferential oxidation reactor is used as a case study. It is found that for a steam reformer based on catalyst-coated foils, mass transfer from the bulk gas to the catalyst surface is the limiting process. The water-gas shift reactor is expected to be the largest component of the fuel processor and is limited by intrinsic catalyst activity, while a successful preferential oxidation unit depends on strict temperature control in order to minimize parasitic hydrogen oxidation. This stepwise approach of sequentially eliminating rate-limiting processes can be used to identify possible means of performance enhancement in a broad range of applications.

Fuels for Canadian research reactors

International Nuclear Information System (INIS)

Feraday, M.A.

1993-01-01

This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

1982-01-01

A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

Behaviour of molten reactor fuels under accident conditions

International Nuclear Information System (INIS)

Xavier Swamikannu, A.; Mathews, C.K.

1980-01-01

The behaviour of molten reactor fuels under accident conditions has received considerable importance in recent times. The chemical processes that occur in the molten state among the fuel, the clad components and the concrete of the containment building under the conditions of a core melt down accident in oxide fuelled reactors have been reviewed with the purpose of identifying areas of developmental work required to be performed to assess and minimize the consequences of such an accident. This includes the computation and estimation of vapour pressure of various gaseous species over the fuel, the clad and the coolant, providing of sacrificial materials in the concrete in order to protect the containment building in order to prevent release of radioactive gases into the atmosphere and understanding the distribution and chemical state of fission products in the molten fuel in order to provide for the effective removal of their decay heats. (auth.)

Application of the pulsed magnetic welding process to nuclear breeder reactor fuel pin end closures

International Nuclear Information System (INIS)

Brown, W.F.

1984-01-01

The pulsed magnetic welding process is a solid state welding process in which metallurgical bonding is effected by impacting metal or alloy parts against each other at high velocity by use of controlled high frequency, high intensity pulsed magnetic fields. This process is similar to the explosive welding process except that magnetic energy is used for impacting the parts together instead of using explosive energy. The pulsed magnetic welding (PMW) process is readily applied to the welding of cylindrical plugs to small diameter tubes. Although breeder reactor fuel pin design may vary in size, the application described here consisted of cladding tubes approximately 6.4 mm in diameter by 244 cm long with a wall thickness of 0.38 mm. After the cladding tubes are filled with fuel pellets and associated metal hardware, tapered end plugs are inserted into the end of the tubes and welded. A typical setup for PMW is described

Process to produce homogenized reactor fuels

International Nuclear Information System (INIS)

Hart, P.E.; Daniel, J.L.; Brite, D.W.

1980-01-01

The fuels consist of a mixture of PuO 2 and UO 2 . In order to increase the homogeneity of mechanically mixed fuels the pellets are sintered in a hydrogen atmosphere with a sufficiently low oxygen potential. This results in a reduction of Pu +4 to Pu +3 . By the reduction process water vapor is obtained increasing the pressure within the PuO 2 particles and causing PuO 2 to be pressed into the uranium oxide structure. (DG) [de

New Concept of Designing Composite Fuel for Fast Reactors with Closing Fuel Cycle

International Nuclear Information System (INIS)

Savchenko, A.; Vatulin, A.; Uferov, O.; Kulakov, G.; Sorokin, V.

2013-01-01

For fast reactors a novel type of promising composite U-PuO2 fuel is proposed which is based on dispersion fuel elements. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. Novel fuel features higher characteristics in comparison to metallic or MOX fuel its fabrication technology is readily accomplished and is environmentally clean. A possibility is demonstrated of fabricating coated steel claddings to protect from interaction with fuel and fission products when use standard rod type MOX or metallic U-Pu-Zr fuel. Novel approach to reprocessing of composite fuel is demonstrated, which allows to separate uranium from burnt plutonium as well as the newly generated fissile plutonium from burnt one without chemical processes, which simplifies the closing of the nuclear fuel cycle. Novel composite fuel combines the advantages of metallic and ceramic types of fuel and has high uranium density that allows also to implicate it in BREST types reactor with conversion ratio more than 1. Peculiarities of closing nuclear cycle with composite fuel are demonstrated that allows more effective re-usage of generated Pu as well as, minimizing r/a wastes by incineration of MA in specially developed IMF design

Research reactor spent fuel in Ukraine

International Nuclear Information System (INIS)

Trofimenko, A.P.

1996-01-01

This paper describes the research reactors in Ukraine, their spent fuel facilities and spent fuel management problems. Nuclear sciences, technology and industry are highly developed in Ukraine. There are 5 NPPs in the country with 14 operating reactors which have total power capacity of 12,800 MW

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

International Nuclear Information System (INIS)

Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

2015-01-01

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF 2 -ThF 4 - 233 UF 4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155

Reactor-specific spent fuel discharge projections: 1985 to 2020

International Nuclear Information System (INIS)

Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

1986-09-01

The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

International Nuclear Information System (INIS)

Ikegami, Tetsuo

2008-01-01

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO 2 spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle. (author)

Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ

International Nuclear Information System (INIS)

Hernandez G, J.

2012-10-01

Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

Micro-Reactor Physics of MOX-Fueled Core

International Nuclear Information System (INIS)

Takeda, T.

2001-01-01

Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design

Static fuel molten salt reactors - simpler, cheaper and safer

International Nuclear Information System (INIS)

Scott, Ian

2015-01-01

The many conceptual designs for Molten Salt Reactors (MSR's) today are all evolutions from the prototype MSR that went critical at Oak Ridge 50 years ago. Critically, they are based on pumping the molten fuel salt from a reaction chamber where the fuel achieves critical mass through a heat exchanger where the resulting heat is transferred to another working fluid. This basic concept was not the first idea that the Oak Ridge scientists considered. Their initial preference was to put the molten salt fuel into tubes, just like solid fuel pellets in their cladding, and circulate a coolant past the tubes. They concluded however that the low thermal conductivity of the salt meant that the tubes could be no wider than 2mm which would be entirely impractical. In this analysis they ignored the contribution of convection to heat transfer in fluids, probably because they were designing an aircraft engine where varying g forces would make convection unreliable. Moltex Energy has re-examined this decision using the modern tools of computational fluid dynamics to simulate convective flow in the molten salt and discovered that in fact tubes of similar diameter to those used for solid fuels are entirely practical. Power densities of 250kW/litre of fuel salt are readily attainable providing a higher overall power density than a PWR reactor. This discovery permits MSR's to be built without any of the complex pumping, passively safe drain systems, on line degassing, filtration and chemical processing needed in pumped MSR's. Their design is very simple and they have many intrinsic safety factors including low pressure operation, chemically unreactive fluids and strongly negative fuel thermal and coolant voiding reactivity coefficients. Most importantly, the highly radioactive fission products are retained in non-volatile form within the fuel tubes in the reactor core. Radioactive fuel salt never leaves the reactor vessel except in an immobile frozen form during

Fuel handling grapple for nuclear reactor plants

International Nuclear Information System (INIS)

Rousar, D.L.

1992-01-01

This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

Standardization of specifications and inspection procedures for LEU plate-type research reactor fuels

International Nuclear Information System (INIS)

1988-06-01

With the transition to high density uranium LEU fuel, fabrication costs of research reactor fuel elements have a tendency to increase because of two reasons. First, the amount of the powder of the uranium compound required increases by more than a factor of five. Second, fabrication requirements are in many cases nearer the fabrication limits. Therefore, it is important that measures be undertaken to eliminate or reduce unnecessary requirements in the specification or inspection procedures of research reactor fuel elements utilizing LEU. An additional stimulus for standardizing specifications and inspection procedures at this time is provided by the fact that most LEU conversions will occur within a short time span, and that nearly all of them will require preparation of new specifications and inspection procedures. In this sense, the LEU conversions offer an opportunity for improving the rationality and efficiency of the fuel fabrication and inspection processes. This report focuses on the standardization of specifications and inspection processes of high uranium density LEU fuels for research reactors. However, in many cases the results can also be extended directly to other research reactor fuels. 15 refs, 1 fig., 3 tabs

Reactor-specific spent fuel discharge projections: 1986 to 2020

International Nuclear Information System (INIS)

Heeb, C.M.; Walling, R.C.; Purcell, W.L.

1987-03-01

The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs

The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1980-01-01

The law intends under the principles of the atomic energy act to regulate the refining, processing and reprocessing businesses of nuclear raw and fuel metarials and the installation and operation of reactors for the peaceful and systematic utilization of such materials and reactors and for securing public safety by preventing disasters, as well as to control internationally regulated things for effecting the international agreements on the research, development and utilization of atomic energy. Basic terms are defined, such as atomic energy; nuclear fuel material; nuclear raw material; nuclear reactor; refining; processing; reprocessing; internationally regulated thing. Any person who is going to engage in refining businesses other than the Power Reactor and Nuclear Fuel Development Corporation shall get the special designation by the Prime Minister and the Minister of International Trade Industry. Any person who is going to engage in processing businesses shall get the particular admission of the Prime Minister. Any person who is going to establish reactors shall get the particular admission of the Prime Minister, The Minister of International Trade and Industry or the Minister of Transportation according to the kinds of specified reactors, respectively. Any person who is going to engage in reprocessing businesses other than the Power Reactor and Nuclear Fuel Development Corporation and the Japan Atomic Energy Research Institute shall get the special designation by the Prime Minister. The employment of nuclear fuel materials and internationally regulated things is defined in detail. (Okada, K.)

Research reactors for power reactor fuel and materials testing - Studsvik's experience

International Nuclear Information System (INIS)

Grounes, M.

1998-01-01

Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

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

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.

Hydriding failure in water reactor fuel elements

International Nuclear Information System (INIS)

Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

1980-01-01

Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

Nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Sasaki, Y.; Tashima, J.

1975-01-01

A description is given of nuclear reactor fuel assemblies arranged in the form of a lattice wherein there is attached to the interface of one of two adjacent fuel assemblies a plate spring having a concave portion curved toward said interface and to the interface of the other fuel assembly a plate spring having a convex portion curved away from said interface

Fuel management approach in IRIS Reactor

International Nuclear Information System (INIS)

Petrovic, B.; Franceschini, F.

2004-01-01

This paper provides an overview of fuel management approach employed in IRIS (International Reactor Innovative and Secure). It introduces the initial, rather ambitious, fuel management goals and discusses their evolution that reflected the fast pace of progress of the overall project. The updated objectives rely on using currently licensed fuel technology, thus enabling near-term deployment of IRIS, while still providing improved fuel utilization. The paper focuses on the reference core design and fuel management strategy that is considered in pre-application licensing, which enables extended cycle of three to four years. The extended cycle reduces maintenance outage time and increases capacity factor, thus reducing the cost of electricity. Approaches to achieving this goal are discussed, including use of different reloading strategies. Additional fuel management options, which are not part of the licensing process, but are pursued as long-term research for possible future implementation, are presented as well. (Author)

Power generation costs for alternate reactor fuel cycles

International Nuclear Information System (INIS)

Smolen, G.R.; Delene, J.G.

1980-09-01

The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

International Nuclear Information System (INIS)

Klein, Andrew; Lenhof, Renae; Deason, Wesley; Harter, Jackson

2015-01-01

Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

Energy Technology Data Exchange (ETDEWEB)

Klein, Andrew [Oregon State Univ., Corvallis, OR (United States). Nuclear Engineering and Radiation Health Physics; Matthews, Topher [Oregon State Univ., Corvallis, OR (United States); Lenhof, Renae [Oregon State Univ., Corvallis, OR (United States); Deason, Wesley [Oregon State Univ., Corvallis, OR (United States); Harter, Jackson [Oregon State Univ., Corvallis, OR (United States)

2015-01-16

Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

2007-07-01

Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

2007-01-01

Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-02-01

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

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Gjertsen, R.K.

1982-01-01

A fuel assembly in a nuclear reactor comprises a locking mechanism that is capable of locking the fuel assembly to the core plate of a nuclear reactor to prevent inadvertent movement of the fuel assembly. The locking mechanism comprises a ratchet mechanism 108 that allows the fuel assembly to be easily locked to the core plate but prevents unlocking except when the ratchet is disengaged. The ratchet mechanism is coupled to the locking mechanism by a rotatable guide tube for a control rod or water displacer rod. (author)

Technical feasibility of an Integral Fast Reactor (IFR) as a future option for fast reactor cycles. Integrate a small metal-fueled fast reactor and pyroprocessing facilities

International Nuclear Information System (INIS)

Tanaka, Nobuo

2017-01-01

Integral Fast Reactor that integrated fast reactor and pyrorocessing facilities developed by Argonne National Laboratory in the U.S. is an excellent nuclear fuel cycle system for passive safety, nuclear non-proliferation, and reduction in radioactive waste. In addition, this system can be considered as a technology applicable to the treatment of the fuel debris caused by the Fukushima Daiichi Nuclear Power Station accident. This study assessed the time required for debris processing, safety of the facilities, and construction cost when using this technology, and examined technological possibility including future technological issues. In a small metal-fueled reactor, it is important to design the core that achieves both of reduction in combustion reactivity and reduction in coolant reactivity. In system design, calorimetric analysis, structure soundness assessment, seismic feasibility establishment study, etc. are important. Regarding safety, research and testing are necessary on the capabilities of passive reactor shutdown and reactor core cooling as well as measures for avoiding re-criticality, even when emergency stop has failed. In dry reprocessing system, studies on electrolytic reduction and electrolytic refining process for treating the debris with compositions different from those of normal fuel are necessary. (A.O.)

Simulation of the Fuel Reactor of a Coal-Fired Chemical Looping Combustor

Science.gov (United States)

Mahalatkar, Kartikeya; O'Brien, Thomas; Huckaby, E. David; Kuhlman, John

2009-06-01

Responsible carbon management (CM) will be required for the future utilization of coal for power generation. CO2 separation is the more costly component of CM, not sequestration. Most methods of capture require a costly process of gas separation to obtain a CO2-rich gas stream. However, recently a process termed Chemical Looping Combustion (CLC) has been proposed, in which an oxygen-carrier is used to provide the oxygen for combustion. This process quite naturally generates a separate exhaust gas stream containing mainly H2O and CO2 but requires two reaction vessels, an Air Reactor (AR) and a Fuel Reactor (FR). The carrier (M for metal, the usual carrier) is oxidized in the AR. This highly exothermic process provides heat for power generation. The oxidized carrier (MO) is separated from this hot, vitiated air stream and transported to the FR where it oxidizes the hydrocarbon fuel, yielding an exhaust gas stream of mainly H2O and CO2. This process is usually slightly endothermic so that the carrier must also transport the necessary heat of reaction. The reduced carrier (M) is then returned to the air reactor for regeneration, hence the term "looping." The net chemical reaction and energy release is identical to that of conventional combustion of the fuel. However, CO2 separation is easily achieved, the only operational penalty being the slight pressure losses required to circulate the carrier. CLC requires many unit operations involving gas-solid or granular flow. To utilize coal in the fuel reactor, in either a moving bed or bubbling fluidized bed, the granular flow is especially critical. The solid coal fuel must be heated by the recycled metal oxide, driving off moisture and volatile material. The remaining char must be gasified by H2O (or CO2), which is recycled from the product stream. The gaseous product of these reactions must then contact the MO before leaving the bed to obtain complete conversion to H2O and CO2. Further, the reduced M particles must be

Growing dimensions. Spent fuel management at research reactors

International Nuclear Information System (INIS)

Ritchie, I.G.

1998-01-01

More than 550 nuclear research reactors are operating or shout down around the world. At many of these reactors, spent fuel from their operations is stored, pending decisions on its final disposition. In recent years, problems associated with this spent fuel storage have loomed larger in the international nuclear community. In efforts to determine the overall scope of problems and to develop a database on the subject, the IAEA has surveyed research reactor operators in its Member States. Information for the Research Reactor Spent Fuel Database (RRSFDB) so far has been obtained from a limited but representative number of research reactors. It supplements data already on hand in the Agency's more established Research Reactor Database (RRDB). Drawing upon these database resources, this article presents an overall picture of spent fuel management and storage at the world's research reactors, in the context of associated national and international programmes in the field

Treatment of spent fuels from research reactors and reactor development programs in Germany

International Nuclear Information System (INIS)

Closs, K.D.

1999-01-01

Quite a great number of different types of spent fuel from research reactors and development programs exists in Germany. The general policy is to send back to the USA as long as possible fuel from MTRs and TRIGAs of USA origin. An option is reprocessing in Great Britain or France. This option is pursued as long as reprocessing and reuse of the recovered material is economically justifiable. For those fuels which cannot be returned to the USA or which will not be reprocessed, a domestic back-up solution of spent fuel management has been developed in Germany, compatible with the management of spent fuel from power reactors. It consists in dry storage in special casks and, later on, direct disposal. Preliminary results from experimental R and D investigations with research reactor fuel and experience from LWR fuel lead to the conclusion that the direct disposal option even for research reactor fuel or exotic fuel does not impose major technical difficulties for the German waste management and disposal concept. (author)

Reactor TRIGA PUSPATI (RTP) spent fuel pool conceptual design

International Nuclear Information System (INIS)

Mohd Fazli Zakaria; Tonny Lanyau; Ahmad Nabil Ab Rahim

2010-01-01

Reactor TRIGA PUSPATI (RTP) is the one and only research reactor in Malaysia that has been safely operated and maintained since 1982. In order to enhance technical capabilities and competencies especially in nuclear reactor engineering a feasibility study on RTP power upgrading was proposed to serve future needs for advance nuclear science and technology in the country with the capability of designing and develop reactor system. The need of a Spent Fuel Pool begins with the discharge of spent fuel elements from RTP for temporary storage that includes all activities related to the storage of fuel until it is either sent for reprocessed or sent for final disposal. To support RTP power upgrading there will be major RTP systems replacement such as reactor components and a new temporary storage pool for fuel elements. The spent fuel pool is needed for temporarily store the irradiated fuel elements to accommodate a new reactor core structure. Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and considered among the most common problems to all countries with nuclear reactors. The output of this paper will provide sufficient information to show the Spent Fuel Pool can be design and build with the adequate and reasonable safety assurance to support newly upgraded TRIGA PUSPATI TRIGA Research Reactor. (author)

Pyrochemical head-end treatment for fast reactor fuel elements

International Nuclear Information System (INIS)

Avogadro, A.

1978-01-01

The paper presents the R and D work performed at Ispra and Mol during the period 1965-1975 in order to find a way to overcome technical and economical difficulties arising when the conventional reprocessing is applied to fast reactor fuel elements. The work had been directed towards 3 specific topics: a) liquid-metal decladding of spent stainless steel - clad fuels (solinox process). b) oxidative pulverisation by fused salts and extraction of volatile fission products (satex process). c) Pyrochemical separation of plutonium from the bulk of the fuel

Plasma-gun fueling for tokamak reactors

International Nuclear Information System (INIS)

Ehst, D.A.

1980-11-01

In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

Transport of volatile fission products in the fuel-to-sheath gap of defective fuel elements during normal and reactor accident conditions

International Nuclear Information System (INIS)

Lewis, B.J.; Bonin, H.W.

1995-01-01

An analytical treatment has been used to model the vapour transport of radioactive fission products released into the fuel-to-sheath gap of defective nuclear fuel elements. The model accounts for both diffusive and bulk-convective transport. Convective transport becomes important as a result of a significant release of gaseous fission products into the gap during a high-temperature reactor accident. However, during normal reactor operation, diffusion is shown to be the dominant process of transport. The model is based on an analysis of several in-reactor tests with operating defective fuel elements, and high-temperature annealing experiments with irradiated fuel specimens. ((orig.))

Fuel elements of research reactors in China

International Nuclear Information System (INIS)

Zhou Yongmao; Chen Dianshan; Tan Jiaqiu

1987-01-01

This paper describes the current status of design, fabrication of fuel elements for research reactors in China, emphasis is placed on the technology of fuel elements for the High Flux Engineering Test Reactor (HFETR). (author)

Corrosion of aluminium-clad spent fuel at RA research reactor

International Nuclear Information System (INIS)

Pesic, M.; Maksin, T.; Dobrijevic, R.; Idjakovic, Z.

2003-01-01

Almost 95% of all spent fuel elements of the RA research reactor in the Vinca Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, are stored in 30 aluminium barrels and about 300 stainless steel channel-holders in the temporary spent fuel storage water pool. The first activities of sludge and water samples, taken from the pool, were measured in 1996-1997 and were followed by analysis of chemical composition of samples. Visual inspections of fuel elements in some stainless steel tubes and of the fuel channels stored in the reactor core have shown that some deposits cover aluminium cladding. Stains and surface discoloration are noted on many of the spent fuel elements that were examined visually during the core unloading and inspections carried out in 1979 - 1984. Some of water samples, taken from pool, about a 150 stainless steel tubes and 16 barrels have shown very high 137-Cs activity compared to low activity measured in pool water. It was concluded that aluminium cladding of the fuel elements was penetrated due to corrosion process. Study on influence of water corrosion processes in the RA reactor storage pool was started within the framework of the IAEA CRP 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water' in 2002. The first test rack with various aluminium and stainless steel coupons, supplied by the IAEA, was immersed in the pool already in 1996. New racks were immersed in 2002 and 2003. The rack immersed in 1996 was taken out from the pool in 2002 and the rack immersed in 2002 was taken out in 2003. Results of the examination of these racks, carried out according to the strategy and the protocol, proposed by the IAEA, are described in this paper. (author)

Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

Science.gov (United States)

Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

2015-12-01

U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

Nuclear reactor fuel assembly

International Nuclear Information System (INIS)

1975-01-01

A description is given of a nuclear reactor fuel assembly comprising a cluster of fuel elements supported by transversal grids so that their axes are parallel to and at a distance from each other, in order to establish interstices for the axial flow of a coolant. At least one of the interstices is occupied by an axial duct reserved for an auxiliary cooling fluid and is fitted with side holes through which the auxiliary cooling fluid is sprayed into the cluster. Deflectors extend as from a transversal grid in a position opposite the holes to deflect the cooling fluid jet towards those parts of the fuel elements that are not accessible to the auxiliary coolant. This assembly is intended for reactors cooled by light or heavy water [fr

High Temperature Gas Cooled Reactor Fuels and Materials

International Nuclear Information System (INIS)

2010-03-01

At the third annual meeting of the technical working group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), held in Vienna, in 2004, it was suggested 'to develop manuals/handbooks and best practice documents for use in training and education in coated particle fuel technology' in the IAEA's Programme for the year 2006-2007. In the context of supporting interested Member States, the activity to develop a handbook for use in the 'education and training' of a new generation of scientists and engineers on coated particle fuel technology was undertaken. To make aware of the role of nuclear science education and training in all Member States to enhance their capacity to develop innovative technologies for sustainable nuclear energy is of paramount importance to the IAEA Significant efforts are underway in several Member States to develop high temperature gas cooled reactors (HTGR) based on either pebble bed or prismatic designs. All these reactors are primarily fuelled by TRISO (tri iso-structural) coated particles. The aim however is to build future nuclear fuel cycles in concert with the aim of the Generation IV International Forum and includes nuclear reactor applications for process heat, hydrogen production and electricity generation. Moreover, developmental work is ongoing and focuses on the burning of weapon-grade plutonium including civil plutonium and other transuranic elements using the 'deep-burn concept' or 'inert matrix fuels', especially in HTGR systems in the form of coated particle fuels. The document will serve as the primary resource materials for 'education and training' in the area of advanced fuels forming the building blocks for future development in the interested Member States. This document broadly covers several aspects of coated particle fuel technology, namely: manufacture of coated particles, compacts and elements; design-basis; quality assurance/quality control and characterization techniques; fuel irradiations; fuel

Nuclear reactor, fuel assembly and neutron measuring system

International Nuclear Information System (INIS)

Chaki, Masao; Murase, Michio; Zukeran, Atsushi; Moriya, Kimiaki

1998-01-01

The present invention provides a BWR type reactor improved with the efficiency of used fuels and fuel economy by increasing a rated power and reducing exchange fuels. Namely, in a BWR type reactor at present, a thermal limit value is determined by conducting nuclear calculation of the reactor core based on data of reactor flow rate measurement and data of neutron flux measurement. However, since the neutron calculation of the reactor core is based on fuel assemblies while the points for the neutron measurement are present at the outside of the fuel assemblies, errors are caused. A margin including the errors has been used as a thermal limit value during operation. In the present invention, neutron fluxes in the fuel assembly as a base of the nuclear calculation can be measured by the same number of neutron detector tubes, but the number of the measuring points is increased to four times. With such procedures, errors caused by the difference of the neutron calculation and values at neutron measuring points can be reduced. As a result, a margin of the thermal limit value is reduced to increase the degree of freedom of reactor operation. Then, the economical property of the reactor operation can be improved. (N.H.)

REACTOR FUEL ELEMENTS TESTING CONTAINER

Science.gov (United States)

Whitham, G.K.; Smith, R.R.

1963-01-15

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Analysis of the ATR fuel element swaging process

International Nuclear Information System (INIS)

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

1995-12-01

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

Breeder reactor fuel fabrication system development

International Nuclear Information System (INIS)

Bennett, D.W.; Fritz, R.L.; McLemore, D.R.; Yatabe, J.M.

1981-01-01

Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

Energy Technology Data Exchange (ETDEWEB)

Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

2003-07-01

Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

Fuel assemblies for use in nuclear reactors

International Nuclear Information System (INIS)

Schluderberg, D.C.

1981-01-01

A fuel assembly for use in pressurized water cooled nuclear fast breeder reactors is described in which moderator to fuel ratios, conducive to a high Pu-U-D 2 O reactor breeding ratio, are obtained whilst at the same time ensuring accurate spacing of fuel pins without the parasitic losses associated with the use of spacer grids. (U.K.)

Catalyzed deuterium fueled reversed-field pinch reactor assessment

International Nuclear Information System (INIS)

Dobrott, D.

1985-01-01

This study is part of a Department of Energy supported alternate fusion fuels program at Science Applications International Corporation. The purpose of this portion of the study is to perform an assessment of a conceptual compact reversed-field pinch reactor (CRFPR) that is fueled by the catalyzed-deuterium (Cat-d) fuel cycle with respect to physics, technology, safety, and cost. The Cat-d CRFPR is compared to a d-t fueled fusion reactor with respect to several issues in this study. The comparison includes cost, reactor performance, and technology requirements for a Cat-d fueled CRFPR and a comparable cost-optimized d-t fueled conceptual design developed by LANL

IAEA activities related to research reactor fuel conversion and spent fuel return programs

International Nuclear Information System (INIS)

Goldman, Ira N.; Adelfang, Pablo; Ritchie, Iain G.

2005-01-01

The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programs have supported research reactor fuel conversion from HEU to low enriched uranium (LEU), and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. (author)

Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

International Nuclear Information System (INIS)

Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

1981-01-01

Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

Fuel assemblies for BWR type reactors

International Nuclear Information System (INIS)

Ishizuka, Takao.

1981-01-01

Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

International Nuclear Information System (INIS)

Delafosse, Jacques.

1976-01-01

French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

Nuclear fuels for material test reactors

International Nuclear Information System (INIS)

Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

1982-01-01

Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

Energy Technology Data Exchange (ETDEWEB)

Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)

2015-09-30

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

Concept of the plant for the BN-800 fast reactor fuel recycling with application of pyro-process and vibro-packing technology

International Nuclear Information System (INIS)

Bychkov, A.V.; Skiba, O.V.; Mayorshin, A.A.; Demidova, L.S.; Kormilitzyna, L.A.; Ishunin, V.S.

2000-01-01

The conception of Plant was developed for MOX-fuel recycle at two BN-800 type fast reactors by pyrochemical reprocessing of irradiated nuclear fuel (INF) and production of vibro-pac fuel pins and SA. INF production process and stages of pyrochemical reprocessing were analyzed. Starting materials were chosen. Characteristics of irradiated SA and requirements for finished products were defined. Volumes of production were estimated. Procedure of waste management was defined. The following description was made: (1) general flow sheet of fuel recycling and partial schemes of single reprocessing; (2) composition of production process equipment; (3) arrangement of production process equipment; (4) lay out of Plant building and engineering communications. Principle economical assessments were made for production under design. (authors)

Impacts of reactor. Induced cladding defects on spent fuel storage

International Nuclear Information System (INIS)

Johnson, A.B.

1978-01-01

Defects arise in the fuel cladding on a small fraction of fuel rods during irradiation in water-cooled power reactors. Defects from mechanical damage in fuel handling and shipping have been almost negligible. No commercial water reactor fuel has yet been observed to develop defects while stored in spent fuel pools. In some pools, defective fuel is placed in closed canisters as it is removed from the reactor. However, hundreds of defective fuel bundles are stored in numerous pools on the same basis as intact fuel. Radioactive species carried into the pool from the reactor coolant must be dealt with by the pool purification system. However, additional radiation releases from the defective fuel during storage appear tu be minimal, with the possible exception of fuel discharged while the reactor is operating (CANDU fuel). Over approximately two decades, defective commercial fuel has been handled, stored, shipped and reprocessed. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1989-01-01

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

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Reprocessing technology for present water reactor fuels

International Nuclear Information System (INIS)

McMurray, P.R.

1977-01-01

The basic Purex solvent extraction technology developed and applied in the U.S. in the 1950's provides a well-demonstrated and efficient process for recovering uranium and plutonium for fuel recycle and separating the wastes for further treatment and packaging. The technologies for confinement of radioactive effluents have been developed but have had limited utilization in the processing of commercial light water reactor fuels. Technologies for solidification and packaging of radioactive wastes have not yet been demonstrated but significant experience has been gained in laboratory and engineering scale experiments with simulated commercial reprocessing wastes and intermediate level wastes. Commercial scale experience with combined operations of all the required processes and equipment are needed to demonstrate reliable reprocessing centers

Fuel management codes for fast reactors

International Nuclear Information System (INIS)

Sicard, B.; Coulon, P.; Mougniot, J.C.; Gouriou, A.; Pontier, M.; Skok, J.; Carnoy, M.; Martin, J.

The CAPHE code is used for managing and following up fuel subassemblies in the Phenix fast neutron reactor; the principal experimental results obtained since this reactor was commissioned are analyzed with this code. They are mainly concerned with following up fuel subassembly powers and core reactivity variations observed up to the beginning of the fifth Phenix working cycle (3/75). Characteristics of Phenix irradiated fuel subassemblies calculated by the CAPHE code are detailed as at April 1, 1975 (burn-up steel damage)

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

LEU fuel fabrication program for the RECH-1 reactor. Status report

International Nuclear Information System (INIS)

Chavez, J.C.; Barrera, M.; Jimenez, O.; Lisboa, J.; Marin, J.

2000-01-01

In 1995 a 50 LEU U 3 Si 2 fuel elements fabrication program for the RECH-1 research reactor was established at the Comision Chilena de Energia Nuclear, CCHEN. After a fabrication process qualification stage, in 1998, four elements were early delivered to the reactor in order to start an irradiation qualification stage. The irradiation has reached an estimated 10% burn-up and no fabrication problems have been detected up to this burn-up level. During 1999 and up to the first quarter of 2000, 19 fuel elements were produced and 7 fuel elements are expected for the end of 2000. This report presents an updated summary of the main results obtained in this fuel fabrication program. A summary of other activities generated by this program, such as in core follow-up of the four leader fuel elements, ISO 9001 implementation for the fabrication process and a fabrication and qualification optimization planning, is also presented here. (author)

Options for treating high-temperature gas-cooled reactor fuel for repository disposal

Energy Technology Data Exchange (ETDEWEB)

Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

1992-02-01

This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

Options for treating high-temperature gas-cooled reactor fuel for repository disposal

International Nuclear Information System (INIS)

Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

1992-02-01

This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched 235 U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched 235 U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing

International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

IAEA activities related to research reactor fuel conversion and spent fuel return programmes

International Nuclear Information System (INIS)

Ritchie, I.G.; Adelfang, P.; Goldman, I.N.

2004-01-01

Full text: The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country of origin where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programmes have supported research reactor fuel conversion from HEU to low enriched uranium, and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. It is hoped that an announcement of the extension of the U.S. Acceptance Programme, which is expected in the very near future, will facilitate the life extensions of many productive TRIGA reactors around the world. (author)

A status report on the integral fast reactor fuels and safety program

International Nuclear Information System (INIS)

Pedersen, D.R.; Seidel, B.R.

1990-01-01

The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor (ALMR) concept being developed at Argonne National Laboratory. The IFR program is specifically responsible for the irradiation performance, advanced core design, safety analysis, and development of the fuel cycle for the US Department of Energy's ALMR program. The basic elements of the IFR concept are (a) metallic fuel, (b) liquid-sodium cooling, (c) modular, pool-type reactor configuration, (d) an integral fuel cycle based upon pyrometallurgical processing. The most significant safety aspects of the IFR program result from its unique fuel design, a ternary alloy of uranium, plutonium, and zirconium. This fuel is based on experience gained through > 25 yr operation of the Experimental Breeder Reactor II (EBR-II) with a uranium alloy metallic fuel. The ultimate criteria for fuel pin design is the overall integrity at the target burnup. The probability of core meltdown is remote; however, a theoretical possibility of core meltdown remains. The next major step in the IFR development program will be a full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. The IFR fuel cycle closure based on pyroprocessing will also have a dramatic impact on waste management options and on actinide recycling

The chemistry of water reactor fuel

International Nuclear Information System (INIS)

Potter, P.E.

1990-01-01

In this paper, the authors discuss features of the changes in chemical constitution which occur in fuel and fuel rods for water reactors during operation and in fault conditions. The fuel for water reactors consists of pellets of urania (UO 2 ) clad in Zircaloy. An essential step in the prediction of the fate of all the radionuclides in a fault or accident is to possess a detailed knowledge of their chemical behavior at all stages of the development of such incidents. In this paper, the authors consider: the chemical constitution of fuel during operation at temperatures corresponding to rather low ratings, together with a quite detailed discussion of the chemistry within the fuel-clad gap; the behavior of fuel subjected to higher temperatures and ratings than those of contemporary fuel; and the changes in constitution on failure of fuel rods in fault or accident conditions

Fuel designs for VVER reactors

International Nuclear Information System (INIS)

Simonov, K.V.; Carbon, P.; Silberstein, A.

1995-01-01

That progresses in efficiency and safety through progresses in technology and better prediction with fully benchmarked upgraded computer codes is a common goal for on the one hand the original designer of the VVER reactors and their respective fuels and on the other hand for EVF a western company resulting from a combined force with highly diversified and complementary talents in reactor and fuel design and manufacturing. It can be expected that this new challenge and dialogue between the two Russian and European industrial ventures will be mutually beneficial and yield innovative and high quality products and as a consequence strong return will be produced for the best interest of utilities operating VVER reactors. (orig./HP)

A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

International Nuclear Information System (INIS)

Guigon, B.; Vacelet, H.; Dornbusch, D.

2000-01-01

Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U 3 Si 2 fuel are discussed. (author)

Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

International Nuclear Information System (INIS)

Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

1985-01-01

The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

Overview of remote technologies applied to research reactor fuel

International Nuclear Information System (INIS)

Oerdoegh, M.; Takats, F.

1999-01-01

This paper gives a brief overview of the remote technologies applied to research reactor fuels. Due to many reasons, the remote technology utilization to research reactor fuel is not so widespread as it is for power reactor fuels, however, the advantages of the application of such techniques are obvious. (author)

Mixed Uranium/Refractory Metal Carbide Fuels for High Performance Nuclear Reactors

International Nuclear Information System (INIS)

Knight, Travis; Anghaie, Samim

2002-01-01

Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for advanced, high-performance reactors. Earlier studies of mixed carbides focused on uranium and either thorium or plutonium as a fuel for fast breeder reactors enabling shorter doubling owing to the greater fissile atom density. However, the mixed uranium/refractory carbides such as (U, Zr, Nb)C have a lower uranium densities but hold significant promise because of their ultra-high melting points (typically greater than 3700 K), improved material compatibility, and high thermal conductivity approaching that of the metal. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders, while hypo-stoichiometric samples with carbon-to-metal (C/M) ratios of 0.92 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold uniaxial pressing, dynamic magnetic compaction, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce high density (low porosity), single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for high performance, ultra-safe nuclear reactor applications. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

An evaluation of the deployment of AIROX-recycled fuel in pressurized water reactors

International Nuclear Information System (INIS)

Jahshan, S.N.; McGeehan, T.J.

1994-01-01

An analytical evaluation is made of the pressurized water reactor (PWR) in-core performance of recycled light water reactor fuel that has been Atomics International reduction oxidation (AIROX) reprocessed and reenriched with fissile materials. The neutronics performance is shown to lie within the neutronics performance of existing high-performance and high-burnup fuels. Three AIROX-recycled fuels are compared with a high-burnup virgin fuel and an equivalent mixed-oxide (MOX) fuel. The AIROX-recycled fuel neutronics performance lies consistently between the virgin and the MOX fuel for both the pin power peaking and the reactivity response characteristics in PWRs. Among the attractive features of AIROX-recycled fuel is that it can optimize fissile and fertile fuel use, minimize final fuel disposal impact on the environment, and provide energy in the process of denaturing weapons-grade fissile materials. The fuel material performance may be anticipated from high-burnup virgin fuel and from MOX fuel performance. Recommendations for lead rod testing and for optimization of the AIROX-processing and resintering techniques are made

In-reactor fuel cladding external corrosion measurement process and results

International Nuclear Information System (INIS)

Thomazet, J.; Musante, Y.; Pigelet, J.

1999-01-01

Analysis of the zirconium alloy cladding behaviour calls for an on-site corrosion measurement device. In the 80's, a FISCHER probe was used and allowed oxide layer measurements to be taken along the outer generating lines of the peripheral fuel rods. In order to allow measurements on inner rods, a thin Eddy current probe called SABRE was developed by FRAMATOME. The SABRE is a blade equipped with two E.C coils is moved through the assembly rows. A spring allows the measurement coil to be clamped on each of the generating lines of the scanned rods. By inserting this blade on all four assembly faces, measurements can also be performed along several generating lines of the same rod. Standard rings are fitted on the device and allow on-line calibration for each measured row. Signal acquisition and processing are performed by LAGOS, a dedicated software program developed by FRAMATOME. The measurements are generally taken at the cycle outage, in the spent fuel pool. On average, data acquisition calls for one shift per assembly (eight hours): this corresponds to more than 2500 measurement points. These measurements are processed statistically by the utility program SAN REMO. All the results are collected in a database for subsequent behaviour analysis: examples of investigated parameters are the thermal/hydraulic conditions of the reactors, the irradiation history, the cladding material, the water chemistry This analysis can be made easier by comparing the behaviour measurement and prediction by means of the COROS-2 corrosion code. (author)

Main results and status of the development of LEU fuel for Russian research reactors

International Nuclear Information System (INIS)

Vatulin, A.; Morozov, A.; Suprun, V.; Dobrikova, I.

2005-01-01

VNIINM develops low enrichment uranium (LEU) fuel on base U-Mo alloys and a novel design of pin-type fuel elements. The development is carried out both for existing reactors, and for new advanced designs of reactors. The work is carried on the following main directions: - irradiate LEU U-Mo dispersion fuel (the uranium density up to 6,0 g/cm 3 ) in two Russian research reactors: MIR (RIAR, Dimitrovgrad) as pin type fuel mini-elements and in WWR-M (PINP, Gatchina) within full-scaled fuel assembly (FA) with pin type fuel elements; - finalize development of design and fabrication process of IRT type FA with pin type fuel elements; - develop methods of reducing of U-Mo fuel --Al matrix interaction under irradiation; - develop fabricating methods of fuel elements on base of monolithic U-Mo fuel. The paper generally reviews the results of calculation, design and technology investigations accomplished by now. (author)

Fuel processing for molten-salt reactors

International Nuclear Information System (INIS)

Hightower, J.R. Jr.

1975-01-01

Progress is reported on the development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors. The metal transfer experiment MTE-3 (for removing rare earths from MSRE fuel salt) was completed and the equipment used in that experiment was examined. The examination showed that no serious corrosion had occurred on the internal surfaces of the vessels, but that serious air oxidation occurred on the external surfaces of the vessels. Analyses of the bismuth phases indicated that the surfaces in contact with the salts were enriched in thorium and iron. Mass transfer coefficients in the mechanically agitated nondispersing contactors were measured in the Salt/Bismuth Flow-through Facility. The measured mass transfer coefficients are about 30 to 40 percent of those predicted by the preferred literature correlation, but were not as low as those seen in some of the runs in MTE-3. Additional studies using water--mercury systems to simulate molten salt-bismuth systems indicated that the model used to interpret results from previous measurements in the water--mercury system has significant deficiencies. Autoresistance heating studies were continued to develop a means of internal heat generation for frozen-wall fluorinators. Equipment was built to test a design of a side arm for the heating electrode. Results of experiments with this equipment indicate that for proper operation the wall temperature must be held much lower than that for which the equipment was designed. Studies with an electrical analog of the equipment indicate that no regions of abnormally high current density exist in the side arm. (JGB)

Sodium fast reactors with closed fuel cycle

CERN Document Server

Raj, Baldev; Vasudeva Rao, PR 0

2015-01-01

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

Implementation of a dry process fuel cycle model into the DYMOND code

International Nuclear Information System (INIS)

Park, Joo Hwan; Jeong, Chang Joon; Choi, Hang Bok

2004-01-01

For the analysis of a dry process fuel cycle, new modules were implemented into the fuel cycle analysis code DYMOND, which was developed by the Argonne National Laboratory. The modifications were made to the energy demand prediction model, a Canada Deuterium Uranium (CANDU) reactor, direct use of spent Pressurized Water Reactor (PWR) fuel in CANDU reactors (DUPIC) fuel cycle model, the fuel cycle calculation module, and the input/output modules. The performance of the modified DYMOND code was assessed for the postulated once-through fuel cycle models including both the PWR and CANDU reactor. This paper presents modifications of the DYMOND code and the results of sample calculations for the PWR once-through and DUPIC fuel cycles

Fissile fuel dynamics of breeder/converter reactors

International Nuclear Information System (INIS)

Harms, A.A.

1978-01-01

The long-term fissile fuel dynamics for a hierarchy of fission reactors covering the range from pure-burners to super-breeders is examined. It is found that the breeding gains of the core and blanket can be used to identify several distinct fissile fuel histories and elucidate the importance of fuel cycle characteristics such as the time dependence of the fissile fuel doubling time. On this basis, a self-sufficient fission reactor is introduced and its determining characteristics are identified. (author)

Paired replacement fuel assemblies for BWR-type reactor

International Nuclear Information System (INIS)

Oguchi, Kazushige.

1997-01-01

There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok

2006-06-15

Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted.

A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2006-06-01

Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted

Wastes and waste management in the uranium fuel cycle for light water reactors

International Nuclear Information System (INIS)

Costello, J.M.

1975-08-01

The manufacturing processes in the uranium fuel cycle for light water reactors have been described with particular reference to the chemical and radiological wastes produced and the waste management procedures employed. The problems and possible solutions of ultimate disposal of high activity fission products and transuranium elements from reprocessing of irradiated fuel have been reviewed. Quantities of wastes arising in each stage of the fuel cycle have been summarised. Wastes arising from reactor operation have been described briefly. (author)

Cermet-fueled reactors for advanced space applications

International Nuclear Information System (INIS)

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

1987-12-01

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

Reactor fuel depletion benchmark of TINDER

International Nuclear Information System (INIS)

Martin, W.J.; Oliveira, C.R.E. de; Hecht, A.A.

2014-01-01

Highlights: • A reactor burnup benchmark of TINDER, coupling MCNP6 to CINDER2008, was performed. • TINDER is a poor candidate for fuel depletion calculations using its current libraries. • Data library modification is necessary if fuel depletion is desired from TINDER. - Abstract: Accurate burnup calculations are key to proper nuclear reactor design, fuel cycle modeling, and disposal estimations. The TINDER code, originally designed for activation analyses, has been modified to handle full burnup calculations, including the widely used predictor–corrector feature. In order to properly characterize the performance of TINDER for this application, a benchmark calculation was performed. Although the results followed the trends of past benchmarked codes for a UO 2 PWR fuel sample from the Takahama-3 reactor, there were obvious deficiencies in the final result, likely in the nuclear data library that was used. Isotopic comparisons versus experiment and past code benchmarks are given, as well as hypothesized areas of deficiency and future work

Advanced ceramic cladding for water reactor fuel

International Nuclear Information System (INIS)

Feinroth, H.

2000-01-01

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

MTR fuel plate qualification in OSIRIS reactor

International Nuclear Information System (INIS)

Sacristan, P.; Boulcourt, P.; Naury, S.; Marchard, L.; Carcreff, H.; Noirot, J.

2005-01-01

Qualification of new MTR fuel needs the irradiation in research reactors under representative neutronic, heat flux and thermohydraulic conditions. The experiments are performed in France in the OSIRIS reactor by irradiating MTR full size fuel plates in the IRIS device located in the reactor core. The fuel plates are easily removed from the device during the shutdown of the reactor for performing thickness measurements along the plates by means of a swelling measurement device. Beside the calculation capabilities, the experimental platform includes: the ISIS neutron mock-up for the measurement of neutron flux distribution along the plates; the γ spectrometry for the purpose of measuring the activities of the radionuclides representative of the power and the burnup and to compare with the neutronic calculation. Owing to the experience feedback, a good agreement is observed between calculation and measurement; destructive post irradiation examinations in the LECA facility (Cadarache). New irradiations with the IRIS device and at higher heat flux are under preparation for qualification of MTR fuels. (author)

Characterization of the molten salt reactor experiment fuel and flush salts

International Nuclear Information System (INIS)

Williams, D.F.; Peretz, F.J.

1996-01-01

Wise decisions about the handling and disposition of spent fuel from the Molten Salt Reactor Experiment (MSRE) must be based upon an understanding of the physical, chemical, and radiological properties of the frozen fuel and flush salts. These open-quotes staticclose quotes properties can be inferred from the extensive documentation of process history maintained during reactor operation and the knowledge gained in laboratory development studies. Just as important as the description of the salt itself is an understanding of the dynamic processes which continue to transform the salt composition and govern its present and potential physicochemical behavior. A complete characterization must include a phenomenological characterization in addition to the typical summary of properties. This paper reports on the current state of characterization of the fuel and flush salts needed to support waste management decisions

Fabrication and testing of uranium nitride fuel for space power reactors

Science.gov (United States)

Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

1988-02-01

Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

International Nuclear Information System (INIS)

Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

1978-01-01

Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

Fuel assembly and nuclear reactor core

International Nuclear Information System (INIS)

Masumi, Ryoji; Aoyama, Motoo; Yamashita, Jun-ichi.

1995-01-01

The present invention concerns a fuel assembly and a nuclear reactor core capable of improving a transmutation rate of transuranium elements while improving a residual rate of fission products. In a reactor core of a BWR type reactor to which fuel rods with transuranium elements (TRU) enriched are loaded, the enrichment degree of transuranium elements occupying in fuel materials is determined not less than 2wt%, as well as a ratio of number of atoms between hydrogen and fuel heavy metals in an average reactor core under usual operation state (H/HM) is determined not more than 3 times. In addition, a ratio of the volumes between coolant regions and fuel material regions is determined not more than 2 times. A T ratio (TRU/Pu) is lowered as the TRU enrichment degree is higher and the H/HM ratio is lower. In order to reduce the T ratio not more than 1, the TRU enrichment degree is determined as not less than 2wt%, and the H/HM ratio is determined to not more than 3 times. Accordingly, since the H/HM ratio is reduced to not more than 1, and TRU is transmuted while recycling it with plutonium, the transmutation ratio of transuranium elements can be improved while improving the residual rate of fission products. (N.H.)

Good Practices for Water Quality Management in Research Reactors and Spent Fuel Storage Facilities

International Nuclear Information System (INIS)

2011-01-01

Water is the most common fluid used to remove the heat produced in a research reactor (RR). It is also the most common media used to store spent fuel elements after being removed from the reactor core. Spent fuel is stored either in the at-reactor pool or in away-from-reactor wet facilities, where the fuel elements are maintained until submission to final disposal, or until the decay heat is low enough to allow migration to a dry storage facility. Maintaining high quality water is the most important factor in preventing degradation of aluminium clad fuel elements, and other structural components in water cooled research reactors. Excellent water quality in spent fuel wet storage facilities is essential to achieve optimum storage performance. Experience shows the remarkable success of many research reactors where the water chemistry has been well controlled. In these cases, aluminium clad fuel elements and aluminium pool liners show few, if any, signs of either localized or general corrosion, even after more than 30 years of exposure to research reactor water. In contrast, when water quality was allowed to degrade, the fuel clad and the structural parts of the reactor have been seriously corroded. The driving force to prepare this publication was the recognition that, even though a great deal of information on research reactor water quality is available in the open literature, no comprehensive report addressing the rationale of water quality management in research reactors has been published to date. This report is designed to provide a comprehensive catalogue of good practices for the management of water quality in research reactors. It also presents a brief description of the corrosion process that affects the components of a research reactor. Further, the report provides a basic understanding of water chemistry and its influence on the corrosion process; specifies requirements and operational limits for water purification systems of RRs; describes good practices

Situation of test and research reactors' spent fuels

International Nuclear Information System (INIS)

Shimizu, Kenichi; Uchiyama, Junzo; Sato, Hiroshi

1996-01-01

The U.S. DOE decided a renewal Off-Site Fuel Policy for stopping to spread a highly enriched uranium which was originally enriched at the U.S., the policy declared that to receive all HEU spent fuels from Test and Research reactors in all the world. In Japan, under bilateral agreement of cooperation between the government of the United States and the government of Japan concerning peaceful uses of nuclear energy, the highly enriched uranium of Test and Research Reactors' fuels was purchased from the U.S. and the fuels had been manufactured in Japan, America, Germany and France. On the other hand, a former president of the U.S. J. Carter proposed that to convert the fuels from HEU to LEU concerning a nonproliferation of nuclear materials in 1978, and Japan absolutely supported this policy. Under this condition, the U.S. stopped to receive the spent fuels from the other countries concerning legal action to the Off-Site Fuels Policy. As a result, the spent fuels are increasing, and to cross to each reactor's storage capacity, and if this policy start, a faced crisis of Test and Research Reactors will be avoided. (author)

Specificity in the licensing process of reduced enrichment in the Bulgarian research reactor

International Nuclear Information System (INIS)

Vitkova, Marietta; Gorinov, Ivan

2005-01-01

The presented paper considers some specific questions of the licensing process regarding the reconstruction of the Bulgarian IRT-2000 research reactor, which includes conversion to the low enriched fuel. This specificity has risen as a result of two facts. The design of the reactor reconstruction was made on the basis of the existing fresh 36% highly enriched fuel. But after finishing of the design process, this fresh highly enriched fuel was shipped back to Russia in the framework of the RERTR program. These facts have involved some changes in both - in the licensing and the design processes. Re-analysis of the neutronic and thermal-hydraulic calculations is required to be made on the base of the technical specifications of the new LEU fuel. To facilitate the licensing process the NRA has adopted regulatory acceptance criteria for approval of the reactor core design with LEU fuel. (author)

Status of research reactor spent fuel world-wide: Database summary

International Nuclear Information System (INIS)

Ritchie, I.G.

1996-01-01

Results complied in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialized and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author). 4 refs, 17 figs, 4 tabs

Fast molten salt reactor-transmuter for closing nuclear fuel cycle on minor actinides

International Nuclear Information System (INIS)

Dudnikov, A. A.; Alekseev, P. N.; Subbotin, S. A.

2007-01-01

Creation fast critical molten salt reactor for burning-out minor actinides and separate long-living fission products in the closed nuclear fuel cycle is the most perspective and actual direction. The reactor on melts salts - molten salt homogeneous reactor with the circulating fuel, working as burner and transmuter long-living radioactive nuclides in closed nuclear fuel cycle, can serve as an effective ecological cordon from contamination of the nature long-living radiotoxic nuclides. High-flux fast critical molten-salt nuclear reactors in structure of the closed nuclear fuel cycle of the future nuclear power can effectively burning-out / transmute dangerous long-living radioactive nuclides, make radioisotopes, partially utilize plutonium and produce thermal and electric energy. Such reactor allows solving the problems constraining development of large-scale nuclear power, including fueling, minimization of radioactive waste and non-proliferation. Burning minor actinides in molten salt reactor is capable to facilitate work solid fuel power reactors in system NP with the closed nuclear fuel cycle and to reduce transient losses at processing and fabrications fuel pins. At substantiation MSR-transmuter/burner as solvents fuel nuclides for molten-salt reactors various salts were examined, for example: LiF - BeF2; NaF - LiF - BeF2; NaF-LiF ; NaF-ZrF4 ; LiF-NaF -KF; NaCl. RRC 'Kurchatov institute' together with other employees have developed the basic design reactor installations with molten salt reactor - burner long-living nuclides for fluoride fuel composition with the limited solubility minor actinides (MAF3 10 mol %) allows to develop in some times more effective molten salt reactor with fast neutron spectrum - burner/ transmuter of the long-living radioactive waste. In high-flux fast reactors on melts salts within a year it is possible to burn ∼300 kg minor actinides per 1 GW thermal power of reactor. The technical and economic estimation given power

Testing of research reactor fuel in the high flux reactor (Petten)

International Nuclear Information System (INIS)

Guidez, J.; Markgraf, J.W.; Sordon, G.; Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.

1999-01-01

The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium ( 3 . However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

Energy Technology Data Exchange (ETDEWEB)

Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

1998-08-01

An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

Assessment of the thorium fuel cycle in power reactors

International Nuclear Information System (INIS)

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

1977-01-01

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

The development of mobile melt-dilute technology for the treatment of former Soviet Union research reactor fuel

International Nuclear Information System (INIS)

Sell, D.A.; Howden, E.A.; Allen, K.J.; Marsden, K.; Westphal, B.R.; Peacock, H.B.; Iyer, N.C.; Fisher, D.L.; Adams, T.M.; Sindelar, R.L.

2004-01-01

United States Government funded national security nuclear non-proliferation projects have historically focused on power reactor spent fuel assemblies that contain weapons usable materials. More recently concern and emphasis have been focused on the spent fuel located at the many research reactor facilities spread throughout the Former Soviet Union. The need exists for a mobile system that can be deployed at these research reactors for the purpose of ensuring that the nuclear materials cannot be used for weapons development. On-site application of the Mobile Melt-Dilute (MMD) process offers an economical method for converting weapons usable Former Soviet Union high enriched uranium research reactor fuel to a safe and secure low enriched uranium ingot. The process will generate little waste and will be performed in a sealed canister that will contain all off-gas products generated during the melting process, eliminating the need for an off-gas treatment system. The process is modular, reusable, and readily portable to a desired reactor site or storage location. The storage canisters containing the melted ingot can be configured for compatibility with the fuel storage technologies currently available or returned to Russia for reprocessing under the Russian Research Reactor Fuel Return Program. The objective of the MMD Project is to develop the mobile melt and dilute technology in preparation for active deployment at Russian built and fueled research reactors. The project has just completed conceptual design and is beginning proof of principle experiments and integrated prototype design of the furnace and canister. (authors)

Gaseous-fuel nuclear reactor research for multimegawatt power in space

Science.gov (United States)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

Burnup performance of OTTO cycle pebble bed reactors with ROX fuel

International Nuclear Information System (INIS)

Ho, Hai Quan; Obara, Toru

2015-01-01

Highlights: • A 300 MW t Small Pebble Bed Reactor with Rock-like oxide fuel is proposed. • Using ROX fuel can achieve high discharged burnup of spent fuel. • High geological stability can be expected in direct disposal of the spent ROX fuel. • The Pebble Bed Reactor with ROX fuel can be critical at steady state operation. • All the reactor designs have a negative temperature coefficient. - Abstract: A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO 2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO 2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO 2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor

Structural analysis of reactor fuel elements

International Nuclear Information System (INIS)

Weeks, R.W.

1977-01-01

An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design

QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

Energy Technology Data Exchange (ETDEWEB)

Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

1968-06-01

Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.

Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

Directory of Open Access Journals (Sweden)

Chan Bock Lee

2016-10-01

Full Text Available Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

Deuterium-tritium fuel self-sufficiency in fusion reactors

International Nuclear Information System (INIS)

Abdou, M.A.; Vold, E.L.; Gung, C.Y.; Youssef, M.Z.; Shin, K.

1986-01-01

Conditions necessary to achieve deuterium-tritium fuel self-sufficiency in fusion reactors are derived through extensive modeling and calculations of the required and achievable tritium breeding ratios as functions of the many reactor parameters and candidate design concepts. It is found that the excess margin in the breeding potential is not sufficient to cover all present uncertainties. Thus, the goal of attaining fuel self-sufficiency significantly restricts the allowable parameter space and design concepts. For example, the required breeding ratio can be reduced by (A) attaining high tritium fractional burnup, >5%, in the plasma, (B) achieving very high reliability, >99%, and very short times, <1 day, to fix failures in the tritium processing system, and (C) ensuring that nonradioactive decay losses from all subsystems are extremely low, e.g., <0.1% for the plasma exhaust processing system. The uncertainties due to nuclear data and calculational methods are found to be significant, but they are substantially smaller than those due to uncertainties in system definition

Fuel-to-cladding heat transfer coefficient into reactor fuel element

International Nuclear Information System (INIS)

Lassmann, K.

1979-01-01

Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. (orig.) [de

Fact reactor fuel alloys: Retrospective and prospective views

International Nuclear Information System (INIS)