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Sample records for htgr fuel reprocessing

  1. HTGR fuel reprocessing technology

    International Nuclear Information System (INIS)

    Brooks, L.H.; Heath, C.A.; Shefcik, J.J.

    1976-01-01

    The following aspects of HTGR reprocessing technology are discussed: characteristics of HTGR fuels, criteria for a fuel reprocessing flowsheet; selection of a reference reprocessing flowsheet, and waste treatment

  2. Flowsheet development for HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Baxter, B.; Benedict, G.E.; Zimmerman, R.D.

    1976-01-01

    Development studies to date indicate that the HTGR fuel blocks can be effectively crushed with two stages of eccentric jaw crushing, followed by a double-roll crusher, a screener and an eccentrically mounted single-roll crusher for oversize particles. Burner development results indicate successful long-term operation of both the primary and secondary fluidized-bed combustion systems can be performed with the equipment developed in this program. Aqueous separation development activities have centered on adapting known Acid-Thorex processing technology to the HTGR reprocessing task. Significant progress has been made on dissolution of burner ash, solvent extraction feed preparation, slurry transfer, solids drying and solvent extraction equipment and flowsheet requirements

  3. Characteristics of radioactive waste streams generated in HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Lin, K.H.

    1976-01-01

    Results are presented of a study concerned with identification and characterization of radioactive waste streams from an HTGR fuel reprocessing plant. Approximate quantities of individual waste streams as well as pertinent characteristics of selected streams have been estimated. Most of the waste streams are unique to HTGR fuel reprocessing. However, waste streams from the solvent extraction system and from the plant facilities do not differ greatly from the corresponding LWR fuel reprocessing wastes

  4. Safety aspects of solvent nitration in HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Wilbourn, R.G.

    1977-06-01

    Reprocessing of HTGR fuels requires evaporative concentration of uranium and thorium nitrate solutions. The results of a bench-scale test program conducted to assess the safety aspects of planned concentrator operations are reported

  5. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  6. Status of reprocessing technology in the HTGR fuel cycle

    International Nuclear Information System (INIS)

    Kaiser, G.; Merz, E.; Zimmer, E.

    1977-01-01

    For more than ten years extensive R and D work has been carried out in the Federal Republic of Germany in order to develop the technology necessary for closing the fuel cycle of high-temperature gas-cooled reactors. The efforts are concentrated primarily on fuel elements having either highly enriched 235 U or recycled 233 U as the fissile and thorium as the fertile material embedded in a graphite matrix. They include the development of processes and equipment for reprocessing and remote preparation of coated microspheres from the recovered uranium. The paper reviews the issues and problems associated with the requirements to deal with high burn-up fuel from HTGR's of different design and composition. It is anticipated that a grind-burn-leach head-end treatment and a modified THOREX-type chemical processing are the optimum choice for the flowsheet. An overview of the present status achieved in construction of a small reprocessing facility, called JUPITER, is presented. It includes a discussion of problems which have already been solved and which have still to be solved like the treatment of feed/breed particle systems and for minimizing environmental impacts envisaged with a HTGR fuel cycle technology. Also discussed is the present status of remote fuel kernel fabrication and coating technology. Additional activities include the design of a mock-up prototype burning head-end facility, called VENUS, with a throughput equivalent to about 6000 MW installed electrical power, as well as a preliminary study for the utilisation of the Karlsruhe LWR prototype reprocessing plant (WAK) to handle HTGR fuel after remodelling of the installations. The paper concludes with an outlook of projects for the future

  7. Process control of an HTGR fuel reprocessing cold pilot plant

    International Nuclear Information System (INIS)

    Rode, J.S.

    1976-10-01

    Development of engineering-scale systems for a large-scale HTGR fuel reprocessing demonstration facility is currently underway in a cold pilot plant. These systems include two fluidized-bed burners, which remove the graphite (carbon) matrix from the crushed HTGR fuel by high temperature (900 0 C) oxidation. The burners are controlled by a digital process controller with an all analog input/output interface which has been in use since March, 1976. The advantages of such a control system to a pilot plant operation can be summarized as follows: (1) Control loop functions and configurations can be changed easily; (2) control constants, alarm limits, output limits, and scaling constants can be changed easily; (3) calculation of data and/or interface with a computerized information retrieval system during operation are available; (4) diagnosis of process control problems is facilitated; and (5) control panel/room space is saved

  8. Research on solvent extraction process for reprocessing of Th-U fuel from HTGR

    International Nuclear Information System (INIS)

    Bao Borong; Wang Gaodong; Qian Jun

    1992-05-01

    The unique properties of spent fuel from HTGR (high temperature gas cooled reactor) have been analysed. The single solvent extraction process using 30% TBP for separation and purification of Th-U fuel has been studied. In addition, the solvent extraction process for second uranium purification is also investigated to meet different needs of reprocessing and reproduction of Th-U spent fuel from HTGR

  9. HTGR fuel reprocessing pilot plant: results of the sequential equipment operation

    International Nuclear Information System (INIS)

    Strand, J.B.; Fields, D.E.; Kergis, C.A.

    1979-05-01

    The second sequential operation of the HTGR fuel reprocessing cold-dry head-end pilot plant equipment has been successfully completed. Twenty standard LHGTR fuel elements were crushed to a size suitable for combustion in a fluid bed burner. The graphite was combusted leaving a product of fissile and fertile fuel particles. These particles were separated in a pneumatic classifier. The fissile particles were fractured and reburned in a fluid bed to remove the inner carbon coatings. The remaining products are ready for dissolution and solvent extraction fuel recovery

  10. Study on reprocessing of uranium-thorium fuel with solvent extraction for HTGR

    International Nuclear Information System (INIS)

    Jiao Rongzhou; He Peijun; Liu Bingren; Zhu Yongjun

    1992-08-01

    A single cycle process by solvent extraction with acid feed solution is suggested. The purpose is to reprocess uranium-thorium fuel elements which are of high burn-up and rich of 232 U from HTGR (high temperature gas cooled reactor). The extraction cascade tests have been completed. The recovery of uranium and thorium is greater than 99.6%. By this method, the requirement, under remote control to re-fabricate fuel elements, of decontamination factors for Cs, Sr, Zr-Nb and Ru has been reached

  11. Process behavior and environmental assessment of 14C releases from an HTGR fuel reprocessing facility

    International Nuclear Information System (INIS)

    Snider, J.W.; Kaye, S.V.

    1976-01-01

    Large quantities of 14 CO 2 will be evolved when graphite fuel blocks are burned during reprocessing of spent fuel from HTGR reactors. The possible release of some or all of this 14 C to the environment is a matter of concern which is investigated in this paper. Various alternatives are considered in this study for decontaminating and releasing the process off-gas to the environment. Concomitant radiological analyses have been done for the waste process scenarios to supply the necessary feedbacks for process design

  12. Overview of HTGR fuel recycle

    International Nuclear Information System (INIS)

    Notz, K.J.

    1976-01-01

    An overview of HTGR fuel recycle is presented, with emphasis placed on reprocessing and fuel kernel refabrication. Overall recycle operations include (1) shipment and storage, (2) reprocessing, (3) refabrication, (4) waste handling, and (5) accountability and safeguards

  13. Fluidized combustion of beds of large, dense particles in reprocessing HTGR fuel

    International Nuclear Information System (INIS)

    Young, D.T.

    1977-03-01

    Fluidized bed combustion of graphite fuel elements and carbon external to fuel particles is required in reprocessing high-temperature gas-cooled reactor (HTGR) cores for recovery of uranium. This burning process requires combustion of beds containing both large particles and very dense particles as well as combustion of fine graphite particles which elutriate from the bed. Equipment must be designed for optimum simplicity and reliability as ultimate operation will occur in a limited access ''hot cell'' environment. Results reported in this paper indicate that successful long-term operation of fuel element burning with complete combustion of all graphite fines leading to a fuel particle product containing <1% external carbon can be performed on equipment developed in this program

  14. Reprocessing yields and material throughput: HTGR recycle demonstration facility

    International Nuclear Information System (INIS)

    Holder, N.; Abraham, L.

    1977-08-01

    Recovery and reuse of residual U-235 and bred U-233 from the HTGR thorium-uranium fuel cycle will contribute significantly to HTGR fuel cycle economics and to uranium resource conservation. The Thorium Utilization National Program Plan for HTGR Fuel Recycle Development includes the demonstration, on a production scale, of reprocessing and refabrication processes in an HTGR Recycle Demonstration Facility (HRDF). This report addresses process yields and material throughput that may be typically expected in the reprocessing of highly enriched uranium fuels in the HRDF. Material flows will serve as guidance in conceptual design of the reprocessing portion of the HRDF. In addition, uranium loss projections, particle breakage limits, and decontamination factor requirements are identified to serve as guidance to the HTGR fuel reprocessing development program

  15. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  16. Separation of the fission product noble gases krypton and xenon from dissolver off-gas in reprocessing HTGR-fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bohnenstingl, J.; Djoa, S. H.; Laser, M.; Mastera, S.; Merz, E.; Morschl, P.

    1976-04-15

    This paper describes a process developed for the retainment and separation of volatile (3H, 129 +131I) and gaseous (85Kr, Xe) fission products from the off-gas produced during dissolution of HTGR-fuel. To prevent unnecessary dilution of liberated noble gases by surrounding atmosphere, a helium purge-gas cycle is applied to enable a coarse fractionating of krypton and xenon by cold-trapping at about 80 deg K after precleaning the gas stream. The process consists of the following steps: deposition of droplets and solid aerosols; chemisorption of iodine on silver impregnated silica gel; catalytic removal of nitrogen oxides and oxygen; drying of the process gas stream; final filtering of abraded solids; deposition of xenon in solid form at 80 deg K and low subpressure; deposition of krypton in solid form at 80 deg K after compression to about 6 bar; decontamination of 85krypton-containing xenon by batch distillation for eventual industrial utilization; and removal of nitrogen and argon enrichment during continuous operation in the purge-gas stream by inleaking air with charcoal. A continuously operating dissolver vessel, closed to the surrounding atmosphere, yields a very high content of noble gases, e.g., 0.35 vol % krypton and 2.0 vol % xenon. The presented off-gas treatment unit is operated in cold runs with 1/3 of the full capacity and can treat about 1 m3 STP/h helium, corresponding to a quantity of about 10,000 MW(e) HTGR-fuel reprocessing plant.

  17. Separation of the fission product noble gases krypton and xenon from dissolver off-gas in reprocessing HTGR-fuel

    International Nuclear Information System (INIS)

    Bohnenstingl, J.; Djoa, S.H.; Laser, M.; Mastera, S.; Merz, E.; Morschl, P.

    1976-01-01

    This paper describes a process developed for the retainment and separation of volatile ( 3 H, 129+131 I) and gaseous ( 85 Kr, Xe) fission products from the off-gas produced during dissolution of HTGR-fuel. To prevent unnecessary dilution of liberated noble gases by surrounding atmosphere, a helium purge-gas cycle is applied to enable a coarse fractionating of krypton and xenon by cold-trapping at about 80 0 K after precleaning the gas stream. The process consists of the following steps: deposition of droplets and solid aerosols; chemisorption of iodine on silver impregnated silica gel; catalytic removal of nitrogen oxides and oxygen; drying of the process gas stream; final filtering of abraded solids; deposition of xenon in solid form at 80 0 K and low subpressure; deposition of krypton in solid form at 80 0 K after compression to about 6 bar; decontamination of 85 Kr-containing xenon by batch distillation for eventual industrial utilization; and removal of nitrogen and argon enrichment during continuous operation in the purge-gas stream by inleaking air with charcoal. A continuously operating dissolver vessel, closed to the surrounding atmosphere, yields a very high content of noble gases, i.e., 0.35 vol % krypton and 2.0 vol % xenon. The presented off-gas treatment unit is operated in cold runs with 1 / 3 of the full capacity and can treat about 1 m 3 STP/h helium, corresponding to a quantity of about 10,000 MW/sub e/ HTGR-fuel reprocessing plant

  18. HTGR fuel cycle

    International Nuclear Information System (INIS)

    1987-08-01

    In the spring of 1987, the HTGR fuel cycle project has been existing for ten years, and for this reason a status seminar has been held on May 12, 1987 in the Juelich Nuclear Research Center, that gathered the participants in this project for a discussion on the state of the art in HTGR fuel element development, graphite development, and waste management. The papers present an overview of work performed so far and an outlook on future tasks and goals, and on taking stock one can say that the project has been very successful so far: The HTGR fuel element now available meets highest requirements and forms the basis of today's HTGR safety philosophy; research work on graphite behaviour in a high-temperature reactor has led to complete knowledge of the temperature or neutron-induced effects, and with the concept of direct ultimate waste disposal, the waste management problem has found a feasible solution. (orig./GL) [de

  19. HTGR Fuel performance basis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents

  20. General Atomic HTGR fuel reprocessing pilot plant: results of initial sequential equipment operation

    International Nuclear Information System (INIS)

    1978-09-01

    In September 1977, the processing of 20 large high-temperature gas-cooled reactor (LHTGR) fuel elements was completed sequentially through the head-end cold pilot plant equipment. This report gives a brief description of the equipment and summarizes the results of the sequential operation of the pilot plant. 32 figures, 15 tables

  1. HTGR fuel performance basis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-01-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents. The slow release of fission products over hundreds of hours allows for decay of short-lived isotopes. The slow and limited release of fission products under HTGR accident conditions results in very low off-site doses. The slow nature of the accident provides more time for operator action to mitigate the accident and for local and state authorities to respond. These features can be used to take advantage of close-in siting for process applications, flexibility in site selection, and emergency planning

  2. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  3. HTGR fuel particle crusher design evaluation

    International Nuclear Information System (INIS)

    Johanson, N.W.

    1978-10-01

    This report describes an evaluation of the design of the existing engineering-scale fuel particle crushing system for the HTGR reprocessing cold pilot plant at General Atomic Company (GA). The purpose of this evaluation is to assess the suitability of the existing design as a prototype of the HTGR Recycle Reference Facility (HRRF) particle crushing system and to recommend alternatives where the existing design is thought to be unsuitable as a prototype. This evaluation has led to recommendations for an upgraded design incorporating improvements in bearing and seal arrangement, housing construction, and control of roll gap thermal expansion. 23 figures, 6 tables

  4. FIPS: a process for the solidification of fission product solutions using a drum drier. [HTGR fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Halaszovich, St.; Laser, M.; Merz, E.; Thiele, D.

    1976-08-01

    A new process consisting of the steps concentration of the fission product solution, denitration of the solution by addition of formaldehyde, addition of glass-forming additives, drying of the slurry using a drum drier, melting of the dry product in the crucible by rising level in-pot-melting, and off-gas treatment and recovery of nitric acid is under development. A small plant with a capacity of 1 kg glass per hour has been tested in hot cells with fission product solutions from LWR fuel element reprocessing since December 1974. The equipment is very simple to operate and to control. No serious problems arose during operation.

  5. Reprocessing flowsheet and material balance for MEU spent fuel

    International Nuclear Information System (INIS)

    Abraham, L.

    1978-10-01

    In response to nonproliferation concerns, the high-temperature gas-cooled reactor (HTGR) Fuel Recycle Development Program is investigating the processing requirements for a denatured medium-enriched uranium--thorium (MEU/Th) fuel cycle. Prior work emphasized the processing requirements for a high-enriched uranium--thorium (HEU/Th) fuel cycle. This report presents reprocessing flowsheets for an HTGR/MEU fuel recycle base case. Material balance data have been calculated for reprocessing of spent MEU and recycle fuels in the HTGR Recycle Reference Facility (HRRF). Flowsheet and mass flow effects in MEU-cycle reprocessing are discussed in comparison with prior HEU-cycle flowsheets

  6. Selected studies in HTGR reprocessing development

    International Nuclear Information System (INIS)

    Notz, K.J.

    1976-03-01

    Recent work at ORNL on hot cell studies, off-gas cleanup, and waste handling is reviewed. The work includes small-scale burning tests with irradiated fuels to study fission product release, development of the KALC process for the removal of 85 Kr from a CO 2 stream, preliminary work on a nonfluidized bed burner, solvent extraction studies including computer modeling, characterization of reprocessing wastes, and initiation of a development program for the fixation of 14 C as CaCO 3

  7. Spent fuel reprocessing system availability definition by process simulation

    International Nuclear Information System (INIS)

    Holder, N.; Haldy, B.B.; Jonzen, M.

    1978-05-01

    To examine nuclear fuel reprocessing plant operating parameters such as maintainability, reliability, availability, equipment redundancy, and surge storage requirements and their effect on plant throughput, a computer simulation model of integrated HTGR fuel reprocessing plant operations is being developed at General Atomic Company (GA). The simulation methodology and the status of the computer programming completed on reprocessing head end systems is reported

  8. Consolidated fuel reprocessing program. Progress report, January 1-March 31, 1981

    International Nuclear Information System (INIS)

    1981-06-01

    Progress and activities are reported on process development, laboratory R and D, engineering research, engineering systems, Integrated Equipment Test (IET) facility operations, and HTGR fuel reprocessing

  9. Nuclear fuel reprocessing expansion strategies

    International Nuclear Information System (INIS)

    Gallagher, J.M.

    1975-01-01

    A description is given of an effort to apply the techniques of operations research and energy system modeling to the problem of determination of cost-effective strategies for capacity expansion of the domestic nuclear fuel reprocessing industry for the 1975 to 2000 time period. The research also determines cost disadvantages associated with alternative strategies that may be attractive for political, social, or ecological reasons. The sensitivity of results to changes in cost assumptions was investigated at some length. Reactor fuel types covered by the analysis include the Light Water Reactor (LWR), High-Temperature Gas-Cooled Reactor (HTGR), and the Fast Breeder Reactor (FBR)

  10. Interim development report: engineering-scale HTGR fuel particle crusher

    International Nuclear Information System (INIS)

    Baer, J.W.; Strand, J.B.

    1978-09-01

    During the reprocessing of HTGR fuel, a double-roll crusher is used to fracture the silicon carbide coatings on the fuel particles. This report describes the development of the roll crusher used for crushing Fort-St.Vrain type fissile and fertile fuel particles, and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. Recommendations are made for design improvements and further testing

  11. HTGR spent fuel storage study

    International Nuclear Information System (INIS)

    Burgoyne, R.M.; Holder, N.D.

    1979-04-01

    This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification

  12. Nuclear Fuel Reprocessing

    International Nuclear Information System (INIS)

    Simpson, Michael F.; Law, Jack D.

    2010-01-01

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

  13. Consolidated fuel reprocessing program

    Science.gov (United States)

    1985-04-01

    A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

  14. Advances in HTGR fuel performance models

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  15. Wastes from fuel reprocessing

    International Nuclear Information System (INIS)

    Eschrich, H.

    1976-01-01

    Handling, treatment, and interim storage of radioactive waste, problems confronted with during the reprocessing of spent fuel elements from LWR's according to the Purex-type process, are dealt with in detail. (HR/LN) [de

  16. Spent fuel reprocessing options

    International Nuclear Information System (INIS)

    2008-08-01

    The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the

  17. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  18. Consolidated fuel reprocessing program. Progress report, July 1-September 30, 1981

    International Nuclear Information System (INIS)

    1981-12-01

    Technical progress is reported in overview fashion in the following areas: process development, laboratory R and D, engineering research, engineering systems, integrated equipment test facility (IET) operations, and HTGR fuel reprocessing

  19. Consolidated fuel reprocessing. Program progress report, April 1-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-09-01

    This progress report is compiled from major contributions from three programs: (1) the Advanced Fuel Recycle Program at ORNL; (2) the Converter Fuel Reprocessing Program at Savannah River Laboratory; and (3) the reprocessing components of the HTGR Fuel Recycle Program, primarily at General Atomic and ORNL. The coverage is generally overview in nature; experimental details and data are limited.

  20. Consolidated fuel reprocessing program

    International Nuclear Information System (INIS)

    Kuban, D.P.; Noakes, M.W.; Bradley, E.C.

    1987-01-01

    The Advanced Servomanipulator (ASM) System consists of three major components: the ASM slave, the dual arm master controller or master, and the control system. The ASM is a remotely maintainable force-reflecting servomanipulator developed at the Oak Ridge National Laboratory (ORNL) as part of the Consolidated Fuel Reprocessing Program of (CFRP). This new manipulator addresses requirements of advanced nuclear fuel reprocessing with emphasis on force reflection, remote maintainability, and reliability. It uses an all-gear force transmission system. The master arms were designed as a kinematic replica of ASM and use cable force transmission. Special digital control algorithms were developed to improve the system performance. The system is presently operational and undergoing evaluation. Preliminary testing has been completed and is reported. The system is now undergoing commercialization by transferring the technology to the private sector

  1. Technical aspects of fuel reprocessing

    International Nuclear Information System (INIS)

    Groenier, W.S.

    1982-02-01

    The purpose of this paper is to present a brief description of fuel reprocessing and some present developments which show the reliability of nuclear energy as a long-term supply. The following topics are discussed: technical reasons for reprocessing; economic reasons for reprocessing; past experience; justification for advanced reprocessing R and D; technical aspects of current reprocessing development. The present developments are mainly directed at the reprocessing of breeder reactor fuels but there are also many applications to light-water reactor fuel reprocessing. These new developments involve totally remote operation, and maintenance. To demonstrate this advanced reprocessing concept, pilot-scale demonstration facilities are planned with commercial application occurring sometime after the year 2000

  2. The investigation of HTGR fuel regeneration process

    Energy Technology Data Exchange (ETDEWEB)

    Lazarev, L N; Bertina, L E; Popik, V P; Isakov, V P; Alkhimov, N B; Pokhitonov, Yu A

    1985-07-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning.

  3. The investigation of HTGR fuel regeneration process

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Bertina, L.E.; Popik, V.P.; Isakov, V.P.; Alkhimov, N.B.; Pokhitonov, Yu.A.

    1985-01-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning

  4. HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

    Strand, J.B.; Cramer, G.T.

    1978-06-01

    Reprocessing of high-temperature gas-cooled reactor fuel requires development of a fuel element size reduction system. This report describes pilot plant testing of crushing equipment designed for this purpose. The test program, the test results, the compatibility of the components, and the requirements for hot reprocessing are discussed

  5. MOX fuel reprocessing and recycling

    International Nuclear Information System (INIS)

    Guillet, J.L.

    1990-01-01

    This paper is devoted to the reprocessing of MOX fuel in UP2-800 plant at La Hague, and to the MOX successive reprocessing and recycling. 1. MOX fuel reprocessing. In a first step, the necessary modifications in UP2-800 to reprocess MOX fuel are set out. Early in the UP2-800 project, actions have been taken to reprocess MOX fuel without penalty. They consist in measures regarding: Dissolution; Radiological shieldings; Nuclear instrumentation; Criticality. 2. Mox successive reprocessing and recycling. The plutonium recycling in the LWR is now a reality and, as said before, the MOX fuel reprocessing is possible in UP2-800 plant at La Hague. The following actions in this field consist in verifying the MOX successive reprocessing and recycling possibilities. After irradiation, the fissile plutonium content of irradiated MOX fuel is decreased and, in this case, the re-use of plutonium in the LWR need an important increase of initial Pu enrichment inconsistent with the Safety reactor constraints. Cogema opted for reprocessing irradiated MOX fuel in dilution with the standard UO2 fuel in appropriate proportions (1 MOX for 4 UO2 fuel for instance) in order to save a fissile plutonium content compatible with MOX successive recycling (at least 3 recyclings) in LWR. (author). 2 figs

  6. Reprocessing of nuclear fuels

    International Nuclear Information System (INIS)

    Hatfield, G.W.

    1960-11-01

    One of the persistent ideas concerning nuclear power is that the fuel costs are negligible. This, of course, is incorrect and, in fact, one of the major problems in the development of economic nuclear power is to get the cost of the fuel cycles down to an acceptable level. The irradiated fuel removed from the nuclear power reactors must be returned as fresh fuel into the system. Aside from the problems of handling and shipping involved in the reprocessing cycles, the two major steps are the chemical separation and the refabrication. The chemical separation covers the processing of the spent fuel to separate and recover the unburned fuel as well as the new fuel produced in the reactor. This includes the decontamination of these materials from other radioactive fission products formed in the reactor. Refabrication involves the working and sheathing of recycled fuel into the shapes and forms required by reactor design and the economics of the fabrication problem determines to a large extent the quality of the material required from the chemical treatment. At present there appear to be enough separating facilities in the United States and the United Kingdom to handle the recycling of fuel from power reactors for the next few years. However, we understand the costs of recycling fuel in these facilities will be high or low depend ing on whether or not the capital costs of the plant are included in the processing cost. Also, the present plants may not be well adapted to carry out the chemical processing of the very wide variety of power reactor fuel elements which are being considered and will continue to be considered over the years to come. (author)

  7. HTGR fuel element structural design consideration

    International Nuclear Information System (INIS)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1987-01-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabilistic stress analysis techniques coupled with probabilistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistant with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the U.S.A. is discussed in the context of stress analysis uncertainty and structural criteria development. (author)

  8. HTGR fuel element structural design considerations

    International Nuclear Information System (INIS)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1986-09-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabalistic stress analysis techniques coupled with probabalistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistent with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the USA is discussed in the context of stress analysis uncertainty and structural criteria development

  9. Thorium utilization program progress report for January 1, 1974--June 30, 1975. [Reprocessing; refabrication; recycle fuel irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Kasten, P.R.

    1976-05-01

    Work was carried out on the following: HTGR reprocessing development and pilot plant, refabrication development and pilot plant, recycle fuel irradiations, engineering and economic studies, and conceptual design of a commercial recycle plant. (DLC)

  10. Fission-product retention in HTGR fuels

    International Nuclear Information System (INIS)

    Homan, F.J.; Kania, M.J.; Tiegs, T.N.

    1982-01-01

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed

  11. Fuel reprocessing/fabrication interface

    International Nuclear Information System (INIS)

    Benistan, G.; Blanchon, T.; Galimberti, M.; Mignot, E.

    1987-01-01

    EDF has conducted a major research, development and experimental programme concerning the recycling of plutonium and reprocessed uranium in pressurized water reactors, in collaboration with its major partners in the nuclear fuel cycle industry. Studies already conducted have demonstrated the technical and economic advantages of this recycling, as also its feasibility with due observance of the safety and reliability criteria constantly applied throughout the industrial development of the nuclear power sector in France. Data feedback from actual experience will make it possible to control the specific technical characteristics of MOX and reprocessed uranium fuels to a higher degree, as also management, viewed from the economic standpoint, of irradiated fuels and materials recovered from reprocessing. The next step will be to examine the reprocessing of MOX for reprocessed uranium fuels, either for secondary recycling in the PWR units, or, looking further ahead, in the fast breeders or later generation PWR units, after a storage period of a few years

  12. Thermal stress analysis of HTGR fuel and control rod fuel blocks in the HTGR in-block carbonization and annealing furnace

    International Nuclear Information System (INIS)

    Gwaltney, R.C.; McAfee, W.J.

    1977-01-01

    A new approach that utilizes the equivalent solid plate method has been applied to the thermal stress analysis of HTGR fuel and control rod fuel blocks. Cases were considered where these blocks, loaded with reprocessed HTGR fuel pellets, were being cured at temperatures up to 1800 0 C. A two-dimensional segment of a fuel block cross section including fuel, coolant holes, and graphite matrix was analyzed using the ORNL HEATING3 heat transfer code to determine the temperature-dependent effective thermal conductivity for the perforated region of the block. Using this equivalent conductivity to calculate the temperature distributions through different cross sections of the blocks, two-dimensional thermal-stress analyses were performed through application of the equivalent solid plate method. In this approach, the perforated material is replaced by solid homogeneous material of the same external dimensions but whose material properties have been modified to account for the perforations

  13. Irradiation performance of HTGR recycle fissile fuel

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.

    1976-08-01

    The irradiation performance of candidate HTGR recycle fissile fuel under accelerated testing conditions is reviewed. Failure modes for coated-particle fuels are described, and the performance of candidate recycle fissile fuels is discussed in terms of these failure modes. The bases on which UO 2 and (Th,U)O 2 were rejected as candidate recycle fissile fuels are outlined, along with the bases on which the weak-acid resin (WAR)-derived fissile fuel was selected as the reference recycle kernel. Comparisons are made relative to the irradiation behavior of WAR-derived fuels of varying stoichiometry and conclusions are drawn about the optimum stoichiometry and the range of acceptable values. Plans for future testing in support of specification development, confirmation of the results of accelerated testing by real-time experiments, and improvement in fuel performance and reliability are described

  14. Reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

    Kidd, S.

    2008-01-01

    The closed fuel cycle is the most sustainable approach for nuclear energy, as it reduces recourse to natural uranium resources and optimises waste management. The advantages and disadvantages of used nuclear fuel reprocessing have been debated since the dawn of the nuclear era. There is a range of issues involved, notably the sound management of wastes, the conservation of resources, economics, hazards of radioactive materials and potential proliferation of nuclear weapons. In recent years, the reprocessing advocates win, demonstrated by the apparent change in position of the USA under the Global Nuclear Energy Partnership (GNEP) program. A great deal of reprocessing has been going on since the fourties, originally for military purposes, to recover plutonium for weapons. So far, some 80000 tonnes of used fuel from commercial power reactors has been reprocessed. The article indicates the reprocessing activities and plants in the United Kigdom, France, India, Russia and USA. The aspect of plutonium that raises the ire of nuclear opponents is its alleged proliferation risk. Opponents of the use of MOX fuels state that such fuels represent a proliferation risk because the plutonium in the fuel is said to be 'weapon-use-able'. The reprocessing of used fuel should not give rise to any particular public concern and offers a number of potential benefits in terms of optimising both the use of natural resources and waste management.

  15. HTGR fuel particle crusher: Mark 2 design

    International Nuclear Information System (INIS)

    Baer, J.W.

    1979-06-01

    The double-roll crusher for fracturing the silicon carbide coatings of high-temperature gas-cooled reactor (HTGR) fuel particles has been redesigned to improve the equipment. The housing was simplified and reduced to a two-piece assembly; the bearings were changed to accommodate thermal effects; the bearing protection seals were improved with triple redundancy; the bearing preload arrangement was simplified and improved; and localized wear areas were reinforced with better materials or special treatment. In addition, the crusher drive was changed for impoved characteristics and an increase in power

  16. HTGR fuel particle crusher: Mark 2 design

    Energy Technology Data Exchange (ETDEWEB)

    Baer, J.W.

    1979-06-01

    The double-roll crusher for fracturing the silicon carbide coatings of high-temperature gas-cooled reactor (HTGR) fuel particles has been redesigned to improve the equipment. The housing was simplified and reduced to a two-piece assembly; the bearings were changed to accommodate thermal effects; the bearing protection seals were improved with triple redundancy; the bearing preload arrangement was simplified and improved; and localized wear areas were reinforced with better materials or special treatment. In addition, the crusher drive was changed for impoved characteristics and an increase in power.

  17. Nondestructive assay of HTGR fuel rods

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1974-01-01

    Performance characteristics of three different radioactive source NDA systems are compared for the assay of HTGR fuel rods and stacks of rods. These systems include the fast neutron Sb-Be assay system, the 252 Cf ''Shuffler,'' and the thermal neutron PAPAS assay system. Studies have been made to determinethe perturbation on the measurements from particle size, kernel Th/U ratio, thorium content, and hydrogen content. In addition to the total 235 U determination, the pellet-to-pellet or rod-to-rod uniformity of HTGR fuel rod stacks has been measured by counting the delayed gamma rays with a NaI through-hole in the PAPAS system. These measurements showed that rod substitutions can be detected easily in a fuel stack, and that detailed information is available on the loading variations in a uniform stack. Using a 1.0 mg 252 Cf source, assay rates of 2 to 4 rods/s are possible, thus facilitating measurement of 100 percent of a plant's throughput. (U.S.)

  18. Fuel reprocessing and waste management

    International Nuclear Information System (INIS)

    Philippone, R.L.; Kaiser, R.A.

    1989-01-01

    Because of different economic, social and political factors, there has been a tendency to compartmentalize the commercial nuclear power industry into separate power and fuel cycle operations to a greater degree in some countries compared to other countries. The purpose of this paper is to describe how actions in one part of the industry can affect the other parts and recommend an overall systems engineering approach which incorporates more cooperation and coordination between individual parts of the fuel cycle. Descriptions are given of the fuel cycle segments and examples are presented of how a systems engineering approach has benefitted the fuel cycle. Descriptions of fuel reprocessing methods and the waste forms generated are given. Illustrations are presented describing how reprocessing options affect waste management operations and how waste management decisions affect reprocessing

  19. Calorimetric assay of HTGR fuel samples

    International Nuclear Information System (INIS)

    Allen, E.J.; McNeany, S.R.; Jenkins, J.D.

    1979-04-01

    A calorimeter using a neutron source was designed and fabricated by Mound Laboratory, according to ORNL specifications. A calibration curve of the device for HTGR standard fuel rods was experimentally determined. The precision of a single measurement at the 95% confidence level was estimated to be +-0.8 μW. For a fuel sample containing 0.3 g 235 U and a neutron source containing 691 μg 252 Cf, this represents a relative standard deviation of 0.5%. Measurement time was approximately 5.5 h per sample. Use of the calorimeter is limited by its relatively poor precision, long measurement time, manual sample changing, sensitivity to room environment, and possibility of accumulated dust blocking water flow through the calorimeter. The calorimeter could be redesigned to resolve most of these difficulties, but not without significant development work

  20. Indian experience in fuel reprocessing

    International Nuclear Information System (INIS)

    Prasad, A.N.; Kumar, S.V.

    1977-01-01

    Plant scale experience in fuel reprocessing in India was started with the successful design, execution and commissioning of the Trombay plant in 1964 to reprocess aluminium clad metallic uranium fuel from the 40 MWt research reactor. The plant has helped in generating expertise and trained manpower for future reprocessing plants. With the Trombay experience, a larger plant of capacity 100 tonnes U/year to reprocess spent oxide fuels from the Tarapur (BWR) and Rajasthan (PHWR) power reactors has been built at Tarapur which is undergoing precommissioning trial runs. Some of the details of this plant are dealt with in this paper. In view of the highly corrosive chemical attack the equipment and piping are subjected to in a fuel reprocessing plant, some of them require replacement during their service if the plant life has to be extended. This calls for extensive decontamination for bringing the radiation levels low enough to establish direct accesss to such equipment. For making modifications in the plant to extend its life and also to enable expansion of capacity, the Trombay plant has been successfully decontaminated and partially decommissioned. Some aspects of thi decontamination campaign are presented in this paper

  1. Reprocessing in breeder fuel cycles

    International Nuclear Information System (INIS)

    Burch, W.D.; Groenier, W.S.

    1982-01-01

    Over the past decade, the United States has developed plans and carried out programs directed toward the demonstration of breeder fuel reprocessing in connection with the first breeder demonstration reactor. A renewed commitment to moving forward with the construction of the Clinch River Breeder Reactor (CRBR) has been made, with startup anticipated near the end of this decade. While plans for the CRBR and its associated fuel cycle are still being firmed up, the basic research and development programs required to carry out the demonstrations have continued. This paper updates the status of the reprocessing plans and programs. Policies call for breeder recycle to begin in the early to mid-1990's. Contents of this paper are: (1) evolving plans for breeder reprocessing (demonstration reprocessing plant, reprocessing head-end colocated at an existing facility); (2) relationship to LWR reprocessing; (3) integrated equipment test (IET) facility and related hardware development activities (mechanical considerations in shearing and dissolving, remote operations and maintenance demonstration phase of IET, integrated process demonstration phase of IET, separate component development activities); and (4) supporting process R and D

  2. Reprocessing and fuel fabrication systems

    International Nuclear Information System (INIS)

    Field, F.R.; Tooper, F.E.

    1978-01-01

    The study of alternative fuel cycles was initiated to identify a fuel cycle with inherent technical resistance to proliferation; however, other key features such as resource use, cost, and development status are major elements in a sound fuel cycle strategy if there is no significant difference in proliferation resistance. Special fuel reprocessing techniques such as coprocessing or spiking provide limited resistance to diversion. The nuclear fuel cycle system that will be most effective may be more dependent on the institutional agreements that can be implemented to supplement the technical controls of fuel cycle materials

  3. Reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

    Schmitt, D.

    1985-01-01

    How should the decision in favour of reprocessing and against alternative waste management concepts be judged from an economic standpoint. Reprocessing is not imperative neither for resource-economic reasons nor for nuclear energy strategy reasons. On the contrary, the development of an ultimate storage concept representing a real alternative promising to close, within a short period of time, the nuclear fuel cycle at low cost. At least, this is the result of an extensive economic efficiency study recently submitted by the Energy Economics Institute which investigated all waste management concepts relevant for the Federal Republic of Germany in the long run, i.e. direct ultimate storage of spent fuel elements (''Other waste disposal technologies'' - AE) as well as reprocessing of spent fuel elements where re-usable plutonium and uranium are recovered and radioactive waste goes to ultimate storage (''Integrated disposal'' - IE). Despite such fairly evident results, the government of the Federal Republic of Germany has favoured the construction of a reprocessing plant. From an economic point of view there is no final answer to the question whether or not the argumentation is sufficient to justify the decision to construct a reprocessing plant. This is true for both the question of technical feasibility and issues of overriding significance of a political nature. (orig./HSCH) [de

  4. Reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

    Gal, I.

    1964-12-01

    This volume contains the following reports: Experimental facility for testing and development of pulsed columns and auxiliary devices; Chemical-technology study of the modified 'Purex' process; Chemical and radiometric control analyses; Chromatographic separation of rare earth elements on paper treated by di-n butylphosphate; Preliminary study of some organic nitrogen extracts significant in fuel reprocessing

  5. The radiological risks associated with the thorium fuelled HTGR fuel cycle. A comparative risk evaluation

    International Nuclear Information System (INIS)

    Dodd, D.H.; Hienen, J.F.A. van.

    1995-10-01

    This report presents the results of task B.3 of the 'Technology Assessment of the High Temperature Reactor' project. The objective of task B.3 was to evaluate the radiological risks to the general public associated with the sustainable HTGR cycle. Since the technologies to be used at several stages of this fuel cycle are still in the design phase and since a detailed specification of this fuel cycle has not yet been developed, the emphasis was on obtaining a global impression of the risk associated with a generic thorium-based HTGR fuel cycle. This impression was obtained by performing a comparative risk analysis on the basis of data given in the literature. As reference for the comparison a generic uranium fuelled LWR cycle was used. The major benefit with respect to the radiological rsiks of basing the fuel cycle around modular HTGR technology instead of the LWR technology is the increase in reactor safety. The design of the modular HTGR is expected to prevent the release of a significant amount of radioactive material to the environment, and hence early deaths in the surrounding population, during accident conditions. This implies that there is no group risk as defined in the Dutch risk management policy. The major benefit of thorium based fuel cycles over uranium based fuel cycles is the reduction in the radiological risks from unraium mining and milling. The other stages of the nuclear fuel cycle which make a significant contribution to the radiological risks are electricity generation, reprocessing and final disposal. The risks associated with the electricity generation stage are dominated by the risks from fission products, activated corrosion products and the activation products tritium and carbon-14. The risks associated with the reprocessing stage are determined by fission and activation products (including actinides). (orig./WL)

  6. The radiological risks associated with the thorium fuelled HTGR fuel cycle. A comparative risk evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Dodd, D.H.; Hienen, J.F.A. van

    1995-10-01

    This report presents the results of task B.3 of the `Technology Assessment of the High Temperature Reactor` project. The objective of task B.3 was to evaluate the radiological risks to the general public associated with the sustainable HTGR cycle. Since the technologies to be used at several stages of this fuel cycle are still in the design phase and since a detailed specification of this fuel cycle has not yet been developed, the emphasis was on obtaining a global impression of the risk associated with a generic thorium-based HTGR fuel cycle. This impression was obtained by performing a comparative risk analysis on the basis of data given in the literature. As reference for the comparison a generic uranium fuelled LWR cycle was used. The major benefit with respect to the radiological rsiks of basing the fuel cycle around modular HTGR technology instead of the LWR technology is the increase in reactor safety. The design of the modular HTGR is expected to prevent the release of a significant amount of radioactive material to the environment, and hence early deaths in the surrounding population, during accident conditions. This implies that there is no group risk as defined in the Dutch risk management policy. The major benefit of thorium based fuel cycles over uranium based fuel cycles is the reduction in the radiological risks from unraium mining and milling. The other stages of the nuclear fuel cycle which make a significant contribution to the radiological risks are electricity generation, reprocessing and final disposal. The risks associated with the electricity generation stage are dominated by the risks from fission products, activated corrosion products and the activation products tritium and carbon-14. The risks associated with the reprocessing stage are determined by fission and activation products (including actinides). (orig./WL).

  7. Nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    White, D.

    1981-01-01

    A simple friction device for cutting nuclear fuel wrappers comprising a thin metal disc clamped between two large diameter clamping plates. A stream of gas ejected from a nozzle is used as coolant. The device may be maintained remotely. (author)

  8. Trends in fuel reprocessing safety research

    International Nuclear Information System (INIS)

    Tsujino, Takeshi

    1981-01-01

    With the operation of a fuel reprocessing plant in the Power Reactor and Nuclear Fuel Development Corporation (PNC) and the plan for a second fuel reprocessing plant, the research on fuel reprocessing safety, along with the reprocessing technology itself, has become increasingly important. As compared with the case of LWR power plants, the safety research in this field still lags behind. In the safety of fuel reprocessing, there are the aspects of keeping radiation exposure as low as possible in both personnel and local people, the high reliability of the plant operation and the securing of public safety in accidents. Safety research is then required to establish the safety standards and to raise the rate of plant operation associated with safety. The following matters are described: basic ideas for the safety design, safety features in fuel reprocessing, safety guideline and standards, and safety research for fuel reprocessing. (J.P.N.)

  9. Spent fuel reprocessing method

    International Nuclear Information System (INIS)

    Shoji, Hirokazu; Mizuguchi, Koji; Kobayashi, Tsuguyuki.

    1996-01-01

    Spent oxide fuels containing oxides of uranium and transuranium elements are dismantled and sheared, then oxide fuels are reduced into metals of uranium and transuranium elements in a molten salt with or without mechanical removal of coatings. The reduced metals of uranium and transuranium elements and the molten salts are subjected to phase separation. From the metals of uranium and transuranium elements subjected to phase separation, uranium is separated to a solid cathode and transuranium elements are separated to a cadmium cathode by an electrolytic method. Molten salts deposited together with uranium to the solid cathode, and uranium and transuranium elements deposited to the cadmium cathode are distilled to remove deposited molten salts and cadmium. As a result, TRU oxides (solid) such as UO 2 , Pu 2 in spent fuels can be reduced to U and TRU by a high temperature metallurgical method not using an aqueous solution to separate them in the form of metal from other ingredients, and further, metal fuels can be obtained through an injection molding step depending on the purpose. (N.H.)

  10. Feasibility study of the Dragon reactor for HTGR fuel testing

    International Nuclear Information System (INIS)

    Wallroth, C.F.

    1975-01-01

    The Organization of European Community Development (OECD) Dragon high-temperature reactor project has performed HTGR fuel and fuel element testing for about 10 years. To date, a total of about 250 fuel elements have been irradiated and the test program continues. The feasibility of using this test facility for HTGR fuel testing, giving special consideration to U. S. needs, is evaluated. A detailed description for design, preparation, and data acquisition of a test experiment is given together with all possible options on supporting work, which could be carried out by the experienced Dragon project staff. 11 references. (U.S.)

  11. Fuel reprocessing and environmental problem

    International Nuclear Information System (INIS)

    Ichikawa, Ryushi

    1977-01-01

    The radioactive nuclides which are released from the reprocessing plants of nuclear fuel are 137 Cs, 106 Ru, 95 Zr and 3 H in waste water and 85 Kr in the atmosphere. This release affects the environment for example, the reprocessing plant of the Nuclear Fuel Service Co in the USA releases about 2 x 10 5 Ci/y of 85 Kr, which is evaluated as about 0.025 mr/y as external exposure dose. The radioactivity in milk around this plant was measured as less than 10 pCi/lit of 129sup(I. The radioactive concentration in the sea, especially in fish and shellfish, was measured near the reprocessing plant of Windscale in UK. The radioactive release rate from this plants more than 10)5sup( Ci/y as the total amount of )137sup(Cs, )3sup(H, )106sup(Ru, )95sup(Zr, )95sup(Nb, )90sup(Sr, )144sup(Ce, etc., and the radioactivity in seaweeds near Windscale is about 400 pCi/g as the maximum value, and the mayonnaise which was made of this seaweeds contained about 1 pCi/g of )106sup(Ru, which is estimated as about 7 mr/y for the digestive organ if 100 g is eaten every day. On the other hand, the experimental result is presented for the reprocessing plant of La Hague in France, in which the radioactive release rate from this plant is about 10)4sup( Ci/y, and the radioactivity in sea water and shellfish is about 4 pCi/l of )106sup(Ru and about 400 pCi/kg of )137 Cs, respectively, near this plant. The philosophy of ALAP (as low as practicable) is also applied to reprocessing plants. (Nakai, Y.)

  12. Study on the Efficient Disintegration of HTGR Fuel Elements by Electrochemical Method

    International Nuclear Information System (INIS)

    Piao Nan; Chen Ji; Xiao Cuiping; We Mingfen; Che Jing

    2014-01-01

    The spent fuel elements in High- temperature gas-cooled reactor (HTGR) have a special structure, so the head-end process of the spent fuel reprocessing is different from the process of water reactor spent fuel. The first step of head-end process of the HTGR spent fuel reprocessing process is disintegration of the graphite matrix and separation of the coated fuel particles. Electrochemical method with nitrate solution as an electrolyte for fuel element disintegration has been conducted by the Institute of Nuclear and New Energy Technology in Tsinghua University. This method allows a total disintegration of graphite matrix, while still preserving the integrity of TRISO particles. The influences of the pretreatment methods such as heating oxidation of graphite, hydrothermal and oxidants oxidation were investigated in the present work. The experimental results showed that there were no significant effects on increasing the disintegration rate when pretreatment methods were used ahead of electrochemical disintegration. This phenomenon indicated that the fuel elements which were calcined at 1073 K and pressed under 300 MPa are too compact to be broken by these pretreatment methods. And the electrochemical disintegration is an effective but slow method in breaking the graphite matrix. (author)

  13. Radioactive characteristics of spent fuels and reprocessing products in thorium fueled alternative cycles

    International Nuclear Information System (INIS)

    Maeda, Mitsuru

    1978-09-01

    In order to provide one fundamental material for the evaluation of Th cycle, compositions of the spent fuels were calculated with the ORIGEN code on following fuel cycles: (1) PWR fueled with Th- enriched U, (2) PWR fueled with Th-denatured U, (3) CANDU fueled with Th-enriched U and (4) HTGR fueled with Th-enriched U. Using these data, product specifications on radioactivity for their reprocessing were calculated, based on a criterion that radioactivities due to foreign elements do not exceed those inherent in nuclear fuel elements, due to 232 U in bred U or 228 Th in recovered Th, respectively. Conclusions are as the following: (1) Because of very high contents of 232 U and 228 Th in the Th cycle fuels from water moderated reactors, especially from PWR, required decontamination factors for their reprocessing will be smaller by a factor of 10 3 to 10 4 , compared with those from U-Pu fueled LWR cycle. (2) These less stringent product specifications on the radioactivity of bred U and recovered Th will justify introduction of some low decontaminating process, with additional advantage of increased proliferation resistance. (3) Decontamination factors required for HTGR fuel will be 10 to 30 times higher than for the other fuels, because of less 232 U and 228 Th generation, and higher burn-up in the fuel. (author)

  14. Transport and reprocessing of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Lenail, B.

    1981-01-01

    This contribution deals with transport and packaging of oxide fuel from and to the Cogema reprocessing plant at La Hague (France). After a general discussion of nuclear fuel and the fuel cycle, the main aspects of transport and reprocessing of oxide fuel are analysed. (Auth.)

  15. Reprocessing of LEU silicide fuel at Dounreay

    International Nuclear Information System (INIS)

    Cartwright, P.

    1996-01-01

    UKAEA have recently reprocessed two LEU silicide fuel elements in their MTR fuel reprocessing plant at Dounreay. The reprocessing was undertaken to demonstrate UKAEA's commitment to the world-wide research reactor communities future needs. Reprocessing of LEU silicide fuel is seen as a waste treatment process, resulting in the production of a liquid feed suitable for conditioning in a stable form of disposal. The uranium product from the reprocessing can be used as a blending feed with the HEU to produce LEU for use in the MTR cycle. (author)

  16. Features of spherical uranium-graphite HTGR fuel elements control

    International Nuclear Information System (INIS)

    Kreindlin, I.I.; Oleynikov, P.P.; Shtan, A.S.

    1985-01-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described

  17. Features of spherical uranium-graphite HTGR fuel elements control

    Energy Technology Data Exchange (ETDEWEB)

    Kreindlin, I I; Oleynikov, P P; Shtan, A S

    1985-07-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described.

  18. Electrochemical reprocessing of nuclear fuels

    International Nuclear Information System (INIS)

    Brambilla, G.; Sartorelli, A.

    1980-01-01

    A method is described for the reprocessing of irradiated nuclear fuel which is particularly suitable for use with fuel from fast reactors and has the advantage of being a dry process in which there is no danger of radiation damage to a solvent medium as in a wet process. It comprises the steps of dissolving the fuel in a salt melt under such conditions that uranium and plutonium therein are converted to sulphate form. The plutonium sulphate may then be thermally decomposed to PuO 2 and removed. The salt melt is then subjected to electrolysis conditions to achieve cathodic deposition of UO 2 (and possibly PuO 2 ). The salt melt can then be recycled or conditioned for final disposal. (author)

  19. 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)

  20. Nuclear fuel reprocessing in the UK

    International Nuclear Information System (INIS)

    Allardice, R.; Harris, D.; Mills, A.

    1983-01-01

    Nuclear fuel reprocessing has been carried out on an industrial scale in the United Kingdom since 1952. Two large reprocessing plants have been constructed and operated at Windscale, Cumbria and two smaller specialized plants have been constructed and operated at Dounreay, Northern Scotland. At the present time, the second of the two Windscale plants is operating, and Government permission has been given for a third reprocessing plant to be built on that site. At Dounreay, one of the plants is operating in its original form, whilst the second is now operating in a modified form, reprocessing fuel from the prototype fast reactor. This chapter describes the development of nuclear fuel reprocessing in the UK, commencing with the research carried out in Canada immediately after the Second World War. A general explanation of the techniques of nuclear fuel reprocessing and of the equipment used is given. This is followed by a detailed description of the plants and processes installed and operated in the UK

  1. Reprocessing of MTR fuel at Dounreay

    International Nuclear Information System (INIS)

    Hough, N.

    1997-01-01

    UKAEA at Dounreay has been reprocessing MTR fuel for over 30 years. During that time considerable experience has been gained in the reprocessing of traditional HEU alloy fuel and more recently with dispersed fuel. Latterly a reprocessing route for silicide fuel has been demonstrated. Reprocessing of the fuel results in a recycled uranium product of either high or low enrichment and a liquid waste stream which is suitable for conditioning in a stable form for disposal. A plant to provide this conditioning, the Dounreay Cementation Plant is currently undergoing active commissioning. This paper details the plant at Dounreay involved in the reprocessing of MTR fuel and the treatment and conditioning of the liquid stream. (author)

  2. Design evaluation of the HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

    Strand, J.B.

    1978-06-01

    A fuel element size reduction system for the ''cold'' pilot plant of the General Atomic HTGR Reference Recycle Facility has been designed and tested. This report is both an evaluation of the design based on results of initial tests and a description of those designs which require completion or modification for hot cell use. 11 figures

  3. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, J.M.

    1980-01-01

    A control algorithm has been derived for an HTGR Fuel Rod Fabrication Process utilizing the method of G.E.P. Box and G.M. Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented. 1 ref

  4. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, M.J.

    1980-01-01

    A control algorithm has been derived for a HTGR Fuel Rod Fabrication Process utilizing the method of Box and Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented

  5. Development of the krypton absorption in liquid carbon dioxide (KALC) process for HTGR off-gas reprocessing

    International Nuclear Information System (INIS)

    Glass, R.W.; Beaujean, H.W.R.; Cochran, H.D. Jr.; Haas, P.A.; Levins, D.M.; Woods, W.M.

    1975-01-01

    Reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel involves burning of the graphite-matrix elements to release the fuel for recovery purposes. The resulting off-gas is primarily CO 2 with residual amounts of N 2 , O 2 , and CO, together with fission products. Trace quantities of krypton-85 must be recovered in a concentrated form from the gas stream, but processes commonly employed for rare gas removal and concentration are not suitable for use with off-gas from graphite burning. The KALC (Krypton Absorption in Liquid CO 2 ) process employs liquid CO 2 as a volatile solvent for the krypton and is, therefore, uniquely suited to the task. Engineering development of the KALC process is currently under way at the Oak Ridge National Laboratory (ORNL) and the Oak Ridge Gaseous Diffusion Plant (ORGDP). The ORNL system is designed for close study of the individual separation operations involved in the KALC process, while the ORGDP system provides a complete pilot facility for demonstrating combined operations on a somewhat larger scale. Packed column performance and process control procedures have been of prime importance in the initial studies. Computer programs have been prepared to analyze and model operational performance of the KALC studies, and special sampling and in-line monitoring systems have been developed for use in the experimental facilities. (U.S.)

  6. Statement on the Consolidated Fuel Reprocessing Program

    International Nuclear Information System (INIS)

    Trauger, D.B.

    1984-01-01

    Oak Ridge National Laboratory has chosen the following objectives for future reprocessing plant design: reduced radiation exposure to workers; minimal environmental impact; improved plant operation and maintenance; improved accountability; no plutonium diversion; and reduced overall capital and operating cost. These objectives lead to a plant with totally remote operation. The Breeder Reactor Engineering Test (BRET) has been designed to perform a key role in demonstrating advanced reprocessing technology. It has been scheduled to be available to reprocess spent fuel from the Fast Flux Test Facility. The principal features of the Consolidated Fuel Reprocessing Program and of the BRET facility are appropriate for all reactor types

  7. The calculation - experimental investigations of the HTGR fuel element construction

    International Nuclear Information System (INIS)

    Eremeev, V.S.; Kolesov, V.S.; Chernikov, A.S.

    1985-01-01

    One of the most important problems in the HTGR development is the creation of the fuel element gas-tight for the fission products. This problem is being solved by using fuel elements of dispersion type representing an ensemble of coated fuel particles dispersed in the graphite matrix. Gas-tightness of such fuel elements is reached at the expense of deposing a protective coating on the fuel particles. It is composed of some layers serving as diffusion barriers for fission products. It is apparent that the rate of fission products diffusion from coated fuel particles is determined by the strength and temperature of the protective coating

  8. Importance of nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Allday, C.

    1977-01-01

    The following topics are discussed: world energy requirements; energy conservation and the economics of recycle environmental considerations and the timescale of reprocessing; and problems associated with reprocessing. The conclusion is reached that reprocessing is essential to the conservation of the world's energy resources and is an environmentally, and probably an economically, more acceptable option to the ''throw away'' alternative

  9. Corrosion control in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Steele, D.F.

    1986-01-01

    This article looks in detail at tribology-related hazards of corrosion in irradiated fuel reprocessing plants and tries to identify and minimize problems which could contribute to disaster. First, the corrosion process is explained. Then the corrosion aspects at each of four stages in reprocessing are examined, with particular reference to oxide fuel reprocessing. The four stages are fuel receipt and storage, fuel breakdown and dissolution, solvent extraction and product concentration and waste management. Results from laboratory and plant corrosion trails are used at the plant design stage to prevent corrosion problems arising. Operational procedures which minimize corrosion if it cannot be prevented at the design stage, are used. (UK)

  10. Use of non-proliferation fuel cycles in the HTGR

    International Nuclear Information System (INIS)

    Baxter, A.M.; Merrill, M.H.; Dahlberg, R.C.

    1978-10-01

    All high-temperature gas-cooled reactors (HTGRs) built or designed to date utilize a uranium-thorium fuel cycle (HEU/Th) in which fully-enriched uranium (93% U-235) is the initial fuel and thorium is the fertile material. The U-233 produced from the thorium is recycled in subsequent loadings to reduce U-235 makeup requirements. However, the recent interest in proliferation-proof fuel cycles for fission reactors has prompted a review and evaluation of possible alternate cycles in the HTGR. This report discusses these alternate fuel cycles, defines those considered usable in an HTGR core, summarizes their advantages and disadvantages, and briefly describes the effect on core design of the most important cycles. Examples from design studies are also given. These studies show that the flexibility afforded by the HTGR coated-particle fuel design allows a variety of alternative cycles, each having special advantages and attractions under different circumstances. Moreover, these alternate cycles can all use the same fuel block, core layout, control scheme, and basic fuel zoning concept

  11. 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

  12. 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

  13. Storage and Reprocessing of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-02

    Addressing the problem of waste, especially high-level waste (HLW), is a requirement of the nuclear fuel cycle that cannot be ignored. We explore the two options employed currently, long-term storage and reprocessing.

  14. ORR irradiation experiment OF-1: accelerated testing of HTGR fuel

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Long, E.L. Jr.; Kania, M.J.; Thoms, K.R.; Allen, E.J.

    1977-08-01

    The OF-1 capsule, the first in a series of High-Temperature Gas-Cooled Reactor fuel irradiations in the Oak Ridge Research Reactor, was irradiated for more than 9300 hr at full reactor power (30 MW). Peak fluences of 1.08 x 10 22 neutrons/cm 2 (> 0.18 MeV) were achieved. General Atomic Company's magazine P13Q occupied the upper two-thirds of the test space and the ORNL magazine OF-1 the lower one-third. The ORNL portion tested various HTGR recycle particles and fuel bonding matrices at accelerated flux levels under reference HTGR irradiation conditions of temperature, temperature gradient, and fast fluence exposure

  15. Remote maintenance in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Herndon, J.N.

    1985-01-01

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

  16. Reprocessing of spent fuel and public acceptance

    International Nuclear Information System (INIS)

    Imai, Ryukichi

    1977-01-01

    The public acceptance has to be considered regarding whole atomic power rather than the reprocessing of nuclear fuel separately, and the problems concerned are as follows; the release of radioactive materials in the normal and abnormal operations of reprocessing plants, the disposal of wastes with high level radioactivity, the transportation of high level radioactive material, the relation to the economic activity near nuclear plants, the environmental effect of 85 Kr. and 3 H, etc., and the physical protection for reprocessing facility itself, the special handling of the materials of very high radioactivity level such as fission products and plutonium, the radiation exposure of operators, and the demonstration of reprocessing techniques of commercial base, etc., as a part of the nuclear fuel cycle, and the relation between atomic power and other technologies in energy supply, the evalution of atomic power as the symbol of huge scale science, and the energy problem within the confrontation of economic development and the preservation of environment and resources regarding whole nuclear energy. The situations of fuel reprocessing in USA, UK, France, Germany and Japan are explained from the viewpoint of the history. The general background for the needs of nuclear energy in Japan, the image of nuclear energy and fuel reprocessing entertained by the general public, and the special feature of reprocessing techniques are described. (Nakai, Y.)

  17. Optimization of MOX fuel cycles in pebble bed HTGR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

    Compared with light water reactor (LWR), the pebble bed high temperature gas-cooled reactor (HTGR) is able to operate in a full mixed oxide (MOX) fuelled core without significant change to core structure design. Based on a reference design of 250 MW pebble bed HTGR, four MOX fuel cycles were designed and evaluated by VSOP program package, including the mixed Pu-U fuel pebbles and mixed loading of separate Pu-pebbles and U-pebbles. Some important physics features were investigated and compared for these four cycles, such as the effective multiplication factor of initial core, the pebble residence time, discharge burnup, and temperature coefficients. Preliminary results show that the overall performance of one case is superior to other equivalent MOX fuel cycles on condition that uranium fuel elements and plutonium fuel elements are separated as the different fuel pebbles and that the uranium fuel elements are irradiated longer in the core than the plutonium fuel elements, and the average discharge burnup of this case is also higher than others. (authors)

  18. Study on the inspection item and inspection method of HTGR fuel

    International Nuclear Information System (INIS)

    Na, Sang Ho; Kim, Y. K.; Jeong, K. C.; Oh, S. C.; Cho, M. S.; Kim, Y. M.; Lee, Y. W.

    2006-01-01

    The type of HTGR(High Temperature Gas-cooled Reactor) fuel is different according to the reactor type. Generally the HTGR fuel has two types. One is a block type, which is manufactured in Japan or America. And the other is a pebble type, which is manufactured in China. Regardless of the fuel type, the fuel manufacturing process started from the coated particle, which is consisted of fuel kernel and the 4 coating layers. Korea has a plan to fabricate a HTGR fuel in near future. The appropriate quality inspection standards are requested to produce a sound and reliable coated particle for HTGR fuel. Therefore, the inspection items and the inspection methods of HTGR fuel between Japan and China, which countries have the manufacturing process, are investigated to establish a proper inspection standards of our product characteristics

  19. Nuclear fuel reprocessing in the UK

    International Nuclear Information System (INIS)

    Allardice, R.H.; Harris, D.W.; Mills, A.

    1983-01-01

    Nuclear fuel reprocessing has been carried out on an industrial scale in the United Kingdom since 1952. Two large reprocessing plants have been constructed and operated at Windscale, Cumbria and two smaller specialized plants have been constructed and operated at Dounreay, Northern Scotland. At the present time, the second of the two Windscale plants is operating, and Government permission has been given for a third reprocessing plant to be built on that site. At Dounreay, one of the plants is operating in its original form, whilst the second is now operating in a modified form, reprocessing fuel from the prototype fast reactor. This chapter describes the development of nuclear fuel reprocessing in the UK, commencing with the research carried out in Canada immediately after the Second World War. A general explanation of the techniques of nuclear fuel reprocessing and of the equipment used is given. This is followed by a detailed description of the plants and processes installed and operated in the UK. (author)

  20. Power Reactor Fuel Reprocessing Plant-2, Tarapur: a benchmark in Indian PHWR spent fuel reprocessing

    International Nuclear Information System (INIS)

    Pradhan, Sanjay; Dubey, K.; Qureshi, F.T.; Lokeswar, S.P.

    2017-01-01

    Power Reactor Fuel Reprocessing Plant-2 (PREFRE-2) is latest operating spent nuclear fuel reprocessing plant in India. This plant has improved design based on latest technology and feedback provided by the earlier plants. The design of PREFRE-2 plant is in five cycles of solvent extraction using TBP as extractant. The plant is commissioned in year 2011 after regulatory clearances

  1. Spent fuel management: reprocessing or storage

    International Nuclear Information System (INIS)

    Lima Soares, M.L. de; Oliveira Lopes, M.J. de

    1986-01-01

    A review of the spent fuel management concepts generally adopted in several countries is presented, including an analysis of the brazilian situation. The alternatives are the reprocessing, the interim storage and the final disposal in a repository after appropriate conditioning. The commercial operating reprocessing facilities in the Western World are located in France and in the United Kingdom. In the USA the anti-reprocessing policy from 1977 changed in 1981, when the government supported the resumption of commercial reprocessing and designated the private sector as responsible for providing these services. Small scale facilities are operating in India, Italy, Japan and West Germany. Pilot plants for LWR fuel are being planned by Spain, Pakistan and Argentina. (Author) [pt

  2. Spent fuel management: reprocessing or storage

    International Nuclear Information System (INIS)

    Lima Soares, M.L. de; Oliveira Lopes, M.J. de.

    1986-01-01

    A review of the spent fuel management concepts generally adopted in several countries is presented, including an analysis of the brazilian situation. The alternatives are the reprocessing, the interim storage and the final disposal in a repository after appropriate conditioning. The commercial operating reprocessing facilities in the Western World are located in France and in the United Kingdom. In the USA the anti-reprocessing policy from 1977 changed in 1981, when the Government supported the resumption of commercial reprocessing and designated the private sector as responsible for providing these services. Small scale facilities are operating in India, Italy, Japan and West Germany. Pilot plant for LWR fuel are being planned by Spain, Pakistan and Argentina. (Author) [pt

  3. PYRO, a system for modeling fuel reprocessing

    International Nuclear Information System (INIS)

    Ackerman, J.P.

    1989-01-01

    Compact, on-site fuel reprocessing and waste management for the Integral Fast Reactor are based on the pyrochemical reprocessing of metal fuel. In that process, uranium and plutonium in spent fuel are separated from fission products in an electrorefiner using liquid cadmium and molten salt solvents. Quantitative estimates of the distribution of the chemical elements among the metal and salt phases are essential for development of both individual pyrochemical process steps and the complete process. This paper describes the PYRO system of programs used to generate reliable mass flows and compositions

  4. Reprocessing of ''fast'' fuel in France

    International Nuclear Information System (INIS)

    Sauteron, J.; Bourgeois, M.; Le Bouhellec, J.; Miquel, P.

    1976-05-01

    The results of laboratory studies as well as pilot testing (AT-I La Hague, Marcoule, Fontenay-aux-Roses) in reprocessing of fast breeder reactor fuels are described. The paper covers all steps: head end, aqueous and fluoride volatility processes, and waste treatment. In conclusion, it is demonstrated why it is still too early to define a strategy of industrial reprocessing for this reactor type

  5. Irradiation experience with HTGR fuels in the Peach Bottom Reactor

    International Nuclear Information System (INIS)

    Scheffel, W.J.; Scott, C.B.

    1974-01-01

    Fuel performance in the Peach Bottom High-Temperature Gas-Cooled Reactor (HTGR) is reviewed, including (1) the driver elements in the second core and (2) the test elements designed to test fuel for larger HTGR plants. Core 2 of this reactor, which is operated by the Philadelphia Electric Company, performed reliably with an average nuclear steam supply availability of 85 percent since its startup in July 1970. Core 2 had accumulated a total of 897.5 equivalent full power days (EFPD), almost exactly its design life-time of 900 EFPD, when the plant was shut down permanently on October 31, 1974. Gaseous fission product release and the activity of the main circulating loop remained significantly below the limits allowed by the technical specifications and the levels observed during operation of Core 1. The low circulating activity and postirradiation examination of driver fuel elements have demonstrated the improved irradiation stability of the coated fuel particles in Core 2. Irradiation data obtained from these tests substantiate the performance predictions based on accelerated tests and complement the fuel design effort by providing irradiation data in the low neutron fluence region

  6. Nuclear fuel reprocessing is challenged

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    This article is a brief discussion of litigation to determine if the Thermal Oxide Reprocessing Plant (THORP) in the United Kingdom will be allowed to operate. Litigants (including Greenpeace) contend that the government's December approval of discharge permits for the plant was unlawful without a public hearing. A description of the THORP process is also provided in this article

  7. Consolidated fuel reprocessing program. Developments for the future in reprocessing

    International Nuclear Information System (INIS)

    Burch, W.D.

    1982-01-01

    The future reprocessing developments focus on three major areas: (1) the retention of gaseous fission products to reduce off-site doses to very low values; (2) the initial steps of breakdown, shearing, and dissolution of breeder fuels; and (3) advanced facility and equipment concepts, which are expected to lead to a reliable, cost-effective, totally remotely operated and maintained plant. Work in the first area - removal of fission gases (the most important of which is 85 Kr) - is largely completed through tracer and bench-scale engineering equipment. Efforts are now mainly devoted to breeder fuels and advanced remote concepts. A facility, the Integrated Equipment Test Facility, which will be used to carry out much of this work, is nearing completion in Oak Ridge. In it a large, simulated, remote reprocessing cell will house a disassembly-shear machine for either breeder or LWR fuels, a rotary continuous dissolver, a solvent extraction cycle utilizing a new generation of centrifugal contactors, and related equipment

  8. Progress report on evaluation of potential impact of 14C releases from an HTGR reprocessing facility

    International Nuclear Information System (INIS)

    Killough, G.G.; Dixon, K.R.; Edwards, N.T.; Murphy, B.D.; Rohwer, P.S.; Harris, W.F.; Kaye, S.V.

    1976-07-01

    The potential radiological impacts of atmospheric releases of 14 CO 2 are assessed for a model HTGR reprocessing facility. Two off-gas systems were considered: (1) a 300-ft stack with no thermal output, and (2) a 1000-ft stack with a stack gas temperature of 80 0 C and heat output of 4.2 x 10 7 Btu/hr. Meteorological data for the Oak Ridge area were used with an assumed annual release rate of 5000 Ci as input to an atmospheric transport model, which in turn was used to predict air concentrations of 14 C at points of habitation and food production in the local area (within 50 miles) of the facility. The total-body dose rates estimated for the average resident living in the local area were 0.107 mrem/yr for the 300-ft stack and 0.063 mrem/yr for the 1000-ft stack. Population doses were computed for a population of 10 6 individuals uniformly distributed within the 50-mile local area of the facility; these were 110 man-rem for the 300-ft stack and 63 man-rem for the 1000-ft stack. The results of these dose calculations suggest that a 1000-ft stack would be very effective in reducing the estimated doses. Plant growth carbon assimilation model was derived in order to investigate the adequacy of the assumption of tissue equilibration with time-averaged ambient specific activity as a basis for dose estimates. Simulation runs with these models suggest that in the presence of frequent fluctuations of large amplitude in the ambient air 14 CO 2 concentrations, specific activity in plant tissue can exceed conventionally calculated time-averaged specific activity

  9. Light water reactor fuel reprocessing and recycling

    International Nuclear Information System (INIS)

    1977-07-01

    This document was originally intended to provide the basis for an environmental impact statement to assist ERDA in making decisions with respect to possible LWR fuel reprocessing and recycling programs. Since the Administration has recently made a decision to indefinitely defer reprocessing, this environmental impact statement is no longer needed. Nevertheless, this document is issued as a report to assist the public in its consideration of nuclear power issues. The statement compares the various alternatives for the LWR fuel cycle. Costs and environmental effects are compared. Safeguards for plutonium from sabotage and theft are analyzed

  10. Advances in HTGR spent fuel treatment technology

    International Nuclear Information System (INIS)

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

    1984-08-01

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

  11. Evaluation of a blender for HTGR fuel particles

    International Nuclear Information System (INIS)

    Johnson, D.R.

    1977-03-01

    An experimental blender for mixing HTGR fuel particles prior to molding the particles into fuel rods was evaluated. The blender consists of a conical chamber with an air inlet in the bottom. A pneumatically operated valve provides for discharge of the particles out of the bottom of the cone. The particles are mixed by periodically levitating with pulses of air. The blender has provision for regulating the air flow rate and the number and duration of the air flow pulses. The performance of the blender was governed by the particle blend being mixed, the air flow rate, and the pulse time. Adequately blended fuel rods can be made, if the air flow rate and pulse time are carefully controlled for each fuel rod composition

  12. Study on the HTGR axial fuel loading

    International Nuclear Information System (INIS)

    Tanaka, Ryokichi

    1981-01-01

    In the nuclear and thermal design of reactor cores, it is one of the important targets for reactor safety to flatten fuel temperature distribution as far as possible to prevent local peaking. As a macroscopic method to prevent temperature peaking, it is considered to give exponential type power output distribution in coolant flow direction, while flattening radial power output distribution. Assuming rod-shaped fuel, the distribution of fuel heat generation is given by an exponential function under constant maximum fuel temperature condition in the direction of channel. By applying this function to neutron source distribution, and in a premise that U-235 loading can be changed continuously, the preliminary investigation on no-reflector core by one-dimensional one-group consideration, and then the analytical solution of the diffusion equation for a core with reflectors by two group one-dimensional approximation were carried out. The results of these investigations revealed that the U-235 concentration required for achieving exponential type power output distribution is necessary to have large concentration gradient up to the distance equivalent to the length of a few fuel elements from the core inlet, but it is sufficient to have constant concentration in downstream fuel elements, which is 0.8 to 0.9 times as much as the average value along the channel, except for large flow rate channel. (Wakatsuki, Y.)

  13. Automatic particle-size analysis of HTGR recycle fuel

    International Nuclear Information System (INIS)

    Mack, J.E.; Pechin, W.H.

    1977-09-01

    An automatic particle-size analyzer was designed, fabricated, tested, and put into operation measuring and counting HTGR recycle fuel particles. The particle-size analyzer can be used for particles in all stages of fabrication, from the loaded, uncarbonized weak acid resin up to fully-coated Biso or Triso particles. The device handles microspheres in the range of 300 to 1000 μm at rates up to 2000 per minute, measuring the diameter of each particle to determine the size distribution of the sample, and simultaneously determining the total number of particles. 10 figures

  14. Power Reactor Fuel Reprocessing Plant-1: a stepping stone in Indian PHWR spent fuel reprocessing

    International Nuclear Information System (INIS)

    Pradhan, Sanjay; Dubey, K.; Qureshi, F.T.; Lokeswar, S.P.

    2017-01-01

    India has low reserves of uranium and high reserves of thorium. In order to optimize resource utilization India has adopted a closed fuel cycle to ensure long-term energy security. The optimum resource utilization is feasible only by adopting reprocessing, conditioning and recycle options. It is very much imperative to view spent fuel as a vital resource material and not a waste to be disposed off. Thus, spent nuclear fuel reprocessing forms an integral part of the Indian Nuclear Energy Programme. Aqueous reprocessing based on PUREX technology is in use for more than 50 years and has reached a matured status

  15. Experimental design for HTGR fuel rods

    International Nuclear Information System (INIS)

    Bayne, C.K.

    1975-01-01

    Fuel rods for the high temperature gas cooled reactor are composed of pyrolytic carbon coated fuel particles bounded by a carbonaceous matrix. Because of differential shrinkage between coated particles and the carbonaceous matrix, breakage of the pyrolytic coating has been observed with certain combinations of coated particles and matrix compositions. The pyrolytic coating is intended to be the primary containment for fission products. Therefore, an experiment is desired to determine the breakage characteristics of different strength coated particles combined with different matrix compositions during irradiation

  16. Review of thorium fuel reprocessing experience

    International Nuclear Information System (INIS)

    Brooksbank, R.E.; McDuffee, W.T.; Rainey, R.H.

    1978-01-01

    The review reveals that experience in the reprocessing of irradiated thorium materials is limited. Plants that have processed thorium-based fuels were not optimized for the operations. Previous demonstrations of several viable flowsheets provide a sound technological base for the development of optimum reprocessing methods and facilities. In addition to the resource benefit by using thorium, recent nonproliferation thrusts have rejuvenated an interest in thorium reprocessing. Extensive radiation is generated as the result of 232 U-contamination produced in the 233 U, resulting in the remote operation and fabrication operations and increased fuel cycle costs. Development of the denatured thorium flowsheet, which is currently of interest because of nonproliferation concerns, represents a difficult technological challenge

  17. HTGR fuel behavior at very high temperature

    International Nuclear Information System (INIS)

    Kashimura, Satoru; Ogawa, Touru; Fukuda, Kousaku; Iwamoto, Kazumi

    1986-03-01

    Fuel behavior at very high temperature simulating abnormal transient of the reactor operation and accidents have been investigated on TRISO coating LEU oxide particle fuels at JAERI. The test simulating the abnormal transient was carried out by irradiation of loose coated particles above 1600 deg C. The irradiation test indicated that particle failure was principally caused by kernel migration. For simulation of the core heat-up accident, two experiments of out-of-pile heating were made. Survival temperature limits were measured and fuel performance at very high temperature were investigated by the heatings. Study on the fuel behavior under reactivity initiated accident was made by NSRR(Nuclear Safety Research Reactor) pulse irradiation, where maximum temperature was higher than 2800 deg C. It was found in the pulse irradiation experiments that the coated particles incorporated in the compacts did not so severely fail unlike the loose coated particles at ultra high temperature above 2800 deg C. In the former particles UO 2 material at the center of the kernel vaporized, leaving a spherical void. (author)

  18. Device for sampling HTGR recycle fuel particles

    International Nuclear Information System (INIS)

    Suchomel, R.R.; Lackey, W.J.

    1977-03-01

    Devices for sampling High-Temperature Gas-Cooled Reactor fuel microspheres were evaluated. Analysis of samples obtained with each of two specially designed passive samplers were compared with data generated by more common techniques. A ten-stage two-way sampler was found to produce a representative sample with a constant batch-to-sample ratio

  19. Summary of nuclear fuel reprocessing activities around the world

    International Nuclear Information System (INIS)

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

    1984-11-01

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

  20. Chemical thermodynamics of iodine species in the HTGR fuel particle

    International Nuclear Information System (INIS)

    Lindemer, T.B.

    1982-09-01

    The iodine-containing species in an intact fuel particle in the high-temperature gas-cooled reactor (HTGR) have been calculated. Assumptions include: (1) attainment of chemical thermodynamic equilibrium among all species in the open porosity of the particle, primarily in the buffer layer; and (2) fission-product concentrations in proportion to their yields. The primary gaseous species is calculated to be cesium iodide; in carbide-containing fuels, gaseous barium iodide may exhibit equivalent pressures. The condensed iodine-containing phase is usually cesium iodide, but in carbide-containing fuels, barium iodide may be stable instead. Absorption of elemental iodine on the carbon in the particle appears to be less than or equal to 10 -4 μg I/g C. The fission-product-spectra excess of cesium over iodine would generally be adsorbed on the carbon, but may form Cs 2 MoO 4 under some circumstances

  1. Fuel reprocessing: Citizens' questions and experts' answers

    International Nuclear Information System (INIS)

    1982-10-01

    In connection with the intention of DWK to erect a fuel reprocessing plant in the Oberpfalz, citizens have asked a great number of questions which are of interest to the general public. They have been collected, grouped into subject categories and answered by experts. (orig./HSCH) [de

  2. Fuel behavior and fission product release under HTGR accident conditions

    International Nuclear Information System (INIS)

    Fukuda, K.; Hayashi, K.; Shiba, K.

    1990-01-01

    In early 1989 a final decision was made over construction of a 30 MWth HTGR called the High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel is a pin-in-block type fuel element which is composed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod. It is necessary for the HTTR licensing to prove the fuel stability under predicted accidents related to the high temperature events. Therefore, the release of the fission products and the fuel failure have been investigated in the irradiation---and the heating experiments simulating these conditions at JAERI. This report describes the HTTR fuel behavior at extreme temperature, made clear in these experiments

  3. Fuel fabrication and reprocessing at UKAEA Dounreay

    International Nuclear Information System (INIS)

    Anderson, B.

    1994-01-01

    The Dounreay fuel plants, which are the most flexible anywhere in the world, will continue to carry out work for foreign commercial customers. A number of German companies are important customers of UKAEA and examples of the wide variety of the work currently being carried out for them in the Dounreay plants is given (reprocessing and fabrication of fuel elements from and for research reactors). (orig./HP) [de

  4. Reprocessing on the whole fuel cycle operations

    International Nuclear Information System (INIS)

    Megy, J.

    1983-11-01

    Spent fuel reprocessing, in France, is become an industrial reality which takes an importance place in several fields: place surely essential in the fuel cycle from the energetic material economy and waste management point of view; place priority in the CEA (Commissariat a l'Energie Atomique) research and development programs; place in the industry where it is an important activity sector with the realizations in progress [fr

  5. Ecological problems of fuel reprocessing

    International Nuclear Information System (INIS)

    Huebschmann, W.G.

    1981-01-01

    The problem of the effects of a reprocessing plant to its environment lies in the amount of the handled radioactivity and its longerity. According to the toxicity of the nuclides extensive measures for retainings and filtering are necessary, in order to keep the resulting radiation load in the surrounding within justified limits. The experiences with the WAK prove, that they managed to reduce that radiation load to values that are negligible compared with the natural one. The expected adaptation of the radiation protection legislation to the latest recommendations of the ICRP will in addition help to do more realistic estimations as to the radiotoxicity of certain nuclides (Kr-85, J-129), which means at lower levels than up to now. (orig./HP) [de

  6. Industrial experience of irradiated nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Delange, M.

    1981-01-01

    At the moment and during the next following years, France and La Hague plant particularly, own the greatest amount of industrial experience in the field of reprocessing, since this experience is referred to three types of reactors, either broadly spread all through the world (GCR and LWR) or ready to be greatly developed in the next future (FBR). Then, the description of processes and technologies used now in France, and the examination of the results obtained, on the production or on the security points of view, are a good approach of the actual industrial experience in the field of spent fuel reprocessing. (author)

  7. Device for reprocessing nuclear fuels

    International Nuclear Information System (INIS)

    Hatano, Mamoru.

    1981-01-01

    Purpose: To readily discharge a nuclear fuel by burning the nuclear fuel as it is without a pulverizing step and removing the graphite and other coated fuel particles. Constitution: An oxygen supply pipe is connected to the lower portion of a discharge chamber having an inlet for the fuel, and an exhaust pipe is connected to the upper portion of the chamber. The fuel mounted on a metallic gripping member made of metallic material is inserted from the inlet, the gripping member is connected through a conductor to a voltage supply unit, oxygen is then supplied through the oxygen supply tube to the discharge chamber, the voltage supply unit is subsequently operated, and discharge takes place among the fuels. Thus, high heat is generated by the discharge, the graphite carbon of the fuel is burnt, silicon carbide is destroyed and decomposed, the isolated nuclear fuel particles are discharged from the exhaust port, and the combustion gas and small embers are exhausted from the exhaust tube. Accordingly, radioactive dusts are not so much generated as when using a mechanical pulverizing means, and prescribed objective can be achieved. (Yoshino, Y.)

  8. Inventory estimation for nuclear fuel reprocessing systems

    International Nuclear Information System (INIS)

    Beyerlein, A.L.; Geldard, J.F.

    1987-01-01

    The accuracy of nuclear material accounting methods for nuclear fuel reprocessing facilities is limited by nuclear material inventory variations in the solvent extraction contactors, which affect the separation and purification of uranium and plutonium. Since in-line methods for measuring contactor inventory are not available, simple inventory estimation models are being developed for mixer-settler contactors operating at steady state with a view toward improving the accuracy of nuclear material accounting methods for reprocessing facilities. The authors investigated the following items: (1) improvements in the utility of the inventory estimation models, (2) extension of improvements to inventory estimation for transient nonsteady-state conditions during, for example, process upset or throughput variations, and (3) development of simple inventory estimation models for reprocessing systems using pulsed columns

  9. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  10. Nuclear fuel cycle: (5) reprocessing of irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Williams, J.A.

    1977-09-01

    The evolution of the reprocessing of irradiated fuel and the recovery of plutonium from it is traced out, starting by following the Manhatten project up to the present time. A brief description of the plant and processes used for reprocessing is given, while the Purex process, which is used in all plants today, is given special attention. Some of the important safety problems of reprocessing plants are considered, together with the solutions which have been adopted. Some examples of the more important safety aspects are the control of activity, criticality control, and the environmental impact. The related topic of irradiated fuel transport is briefly discussed.

  11. Crushing strength of HTGR fuel particles

    International Nuclear Information System (INIS)

    Lackey, W.J.; Stinton, D.P.; Davis, L.E.; Beatty, R.L.

    1976-01-01

    The whole-particle crushing strengths of High-Temperature Gas-Cooled Reactor fertile and fissile coated particles were measured and correlated with fabrication procedures. The crushing strength of Biso-coated fertile particles was increased by the following factors: (1) increasing the outer coating thickness by 10 μm increased strengths by 0.3 lb (1.3 N) for annealed particles and by 0.5 lb (2.2 N) for unannealed particles. (2) An 1800 0 C postcoating anneal increased strengths by 1 lb (4.4 N) for particles with thick outer coatings and by 2 lb (8.9 N) for particles having thin coatings. (3) Increasing the inner coating density by 0.1 g/cm 3 increased strength by 0.6 lb (2.7 N). The crushing strength of Triso-coated fissile particles was proportional to the thickness of the SiC coatings, and strength decreased on annealing by about 0.2 lb (0.9 N) when a porous plate was used to distribute the coating gas and by about 1.5 lb (6.7 N) when a conical gas distributor was used. The strengths of fertile and fissile coated particles as well as uncoated kernels appear adequate to allow fuel fabrication without excessive particle damage

  12. Problems of nuclear fuel reprocessing in Japan

    International Nuclear Information System (INIS)

    Tanaka, Naojiro

    1974-01-01

    The reprocessing capacity of the plant No. 1 of Power Reactor and Nuclear Fuel Development Corporation, which is scheduled to start operation in fiscal year 1975, will be insufficient after fiscal year 1978 for the estimated demand for reprocessing based on Japanese nuclear energy development program. Taking into consideration the results examined by JAIF's study team to Europe and the U.S., it is necessary that Japan builds 2nd reprocessing plant. But there will be a gap from 1978 to 1984 during which Japan must rely on overseas reprocessing services. The establishment of a reprocessing system is a task of national scale, and there are many problems to be solved before it can be done. These include the problems of site and environment, the problem of treatment and disposal of radioactive wastes, the raising of huge required funds and so on. Therefore, even if a private enterprise is allowed to undertake the task, it will be impossible to achieve the aim without the cooperation and assistance of the government. (Wakatsuki, Y.)

  13. Quality control procedures for HTGR fuel element components

    International Nuclear Information System (INIS)

    Delle, W.W.; Koizlik, K.; Luhleich, H.; Nickel, H.

    1976-08-01

    The growing use of nuclear reactors for the production of electric power throughout the world, and the consequent increase in the number of nuclear fuel manufacturers, is giving enhanced importance to the consideration of quality assurance in the production of nuclear fuels. The fuel is the place, where the radioactive fission products are produced in the reactor and, therefore, the integrity of the fuel is of utmost importance. The first and most fundamental means of insuring that integrity is through the exercise of properly designed quality assurance programmes during the manufacture of the fuel and other fuel element components. The International Atomic Energy Agency therefore conducted an International Seminar on Nuclear Fuel Quality Assurance in Oslo, Norway from 24 till 28 May, 1976. This KFA report contains a paper which was distributed preliminary during the seminar and - in the second part - the text of the oral presentation. The paper gives a summary of the procedures available in the present state for the production control of HTGR core materials and of the meaning of the particular properties for reactor operation. (orig./UA) [de

  14. Reprocessing of irradiated fuel: pros and cons

    International Nuclear Information System (INIS)

    Lebedev, O.G.; Novikov, V.M.

    1991-01-01

    The acceptable-safety nuclear reactors (APWR, LMFBR, MSBR, MSCR) can be provided by the enrichment industry and by plutonium reserves. But steady accumulation of spent fuel will inevitably make to return to the problems of fuel recycle. PUREX-processing increases a danger of radionuclides spreading due to the presence of large buffer tanks. Using of compact fluoride - volatility process will sharply reduce a nuclide leakage likewise permit to reprocess a fuel with a burnup as high as possible. Success of a powerful robots development give an opportunity to design a fluoride-volatility plant twice cheaper than PUREX. (author)

  15. Status and trends in spent fuel reprocessing

    International Nuclear Information System (INIS)

    2005-09-01

    The management of spent fuel arising from nuclear power production is a crucial issue for the sustainable development of nuclear energy. The IAEA has issued several publications in the past that provide technical information on the global status and trends in spent fuel reprocessing and associated topics, and one reason for this present publication is to provide an update of this information which has mostly focused on the conventional technology applied in the industry. However, the scope of this publication has been significantly expanded in an attempt to make it more comprehensive and by including a section on emerging technologies applicable to future innovative nuclear systems, as are being addressed in such international initiatives as INPRO, Gen IV and MICANET. In an effort to be informative, this publication attempts to provide a state-of-the-art review of these technologies, and to identify major issues associated with reprocessing as an option for spent fuel management. It does not, however, provide any detailed information on some of the related issues such as safety or safeguards, which are addressed in other relevant publications. This report provides an overview of the status of reprocessing technology and its future prospects in terms of various criteria in Section 2. Section 3 provides a review of emerging technologies which have been attracting the interest of Member States, especially in the international initiatives for future development of innovative nuclear systems. A historical review of IAEA activities associated with spent fuel reprocessing, traceable back to the mid-1970s, is provided in Section 4, and conclusions in Section 5. A list of references is provided at the end the main text for readers interested in further information on the related topics. Annex I summarizes the current status of reprocessing facilities around the world, including the civil operational statistics of Purex-based plants, progress with decommissioning and

  16. Consolidated Reprocessing Progam. Quarterly progress report ending February 28, 1979

    International Nuclear Information System (INIS)

    1979-03-01

    This publication continues the quarterly series presenting results of work performed under the Consolidated Reprocessing Program at General Atomic Company. Results of work on this program prior to June 1974 were included in a quarterly series on the HTGR Base Program. The work reported includes the development of unit processes and equipment for reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel, the design and development of an integrated pilot line to demonstrate the head end of HTGR reprocessing using unirradiated fuel materials, and design work in support of Hot Engineering Tests (HET). Work is also described on trade-off studies concerning the required design of facilities and equipment for the large-scale recycle of HTGR fuels in order to guide the development activities for HTGR fuel recycle

  17. 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

  18. Integrated international safeguards concepts for fuel reprocessing

    International Nuclear Information System (INIS)

    Hakkila, E.A.; Gutmacher, R.G.; Markin, J.T.; Shipley, J.P.; Whitty, W.J.; Camp, A.L.; Cameron, C.P.; Bleck, M.E.; Ellwein, L.B.

    1981-12-01

    This report is the fourth in a series of efforts by the Los Alamos National Laboratory and Sandia National Laboratories, Albuquerque, to identify problems and propose solutions for international safeguarding of light-water reactor spent-fuel reprocessing plants. Problem areas for international safeguards were identified in a previous Problem Statement (LA-7551-MS/SAND79-0108). Accounting concepts that could be verified internationally were presented in a subsequent study (LA-8042). Concepts for containment/surveillance were presented, conceptual designs were developed, and the effectiveness of these designs was evaluated in a companion study (SAND80-0160). The report discusses the coordination of nuclear materials accounting and containment/surveillance concepts in an effort to define an effective integrated safeguards system. The Allied-General Nuclear Services fuels reprocessing plant at Barnwell, South Carolina, was used as the reference facility

  19. Pyrochemical reprocessing of nitride fuel

    International Nuclear Information System (INIS)

    Nakazono, Yoshihisa; Iwai, Takashi; Arai, Yasuo

    2004-01-01

    Electrochemical behavior of actinide nitrides in LiCl-KCl eutectic melt was investigated in order to apply pyrochemical process to nitride fuel cycle. The electrode reaction of UN and (U, Nd)N was examined by cyclic voltammetry. The observed rest potential of (U, Nd)N depended on the equilibrium of U 3+ /UN and was not affected by the addition of NdN of 8 wt.%. (author)

  20. Nuclear safety in fuel-reprocessing plants

    International Nuclear Information System (INIS)

    Hennies, H.H.; Koerting, K.

    1976-01-01

    The danger potential of nuclear power and fuel reprocessing plants in normal operation is compared. It becomes obvious that there are no basic differences. The analysis of possible accidents - blow-up of an evaporator for highly active wastes, zircaloy burning, cooling failure in self-heating process solutions, burning of a charged solvent, criticality accidents - shows that they are kept under control by the plant layout. (HP) [de

  1. History and current status of nuclear fuel reprocessing technology

    International Nuclear Information System (INIS)

    Funasaka, Hideyuki; Nagai, Toshihisa; Washiya, Tadahiro

    2008-01-01

    History and present state of fast breeder reactor was reviewed in series. As a history and current status of nuclear fuel reprocessing technology, this ninth lecture presented the progress of the FBR fuel reprocessing technology and advanced reprocessing processes. FBR fuel reprocessing technology had been developed to construct the reprocessing equipment test facilities (RETF) based on PUREX process technologies. With economics, reduction of environmental burdens and proliferation resistance taken into consideration, advanced aqueous method for nuclear fuel cycle activities has been promoted as the government's basic policy. Innovative technologies on mechanical disassembly, continuous rotary dissolver, crystallizer, solvent extraction and actinides recovery have been mainly studied. (T. Tanaka)

  2. Reprocessing of nuclear fuels: economical, ecological and technical aspects

    International Nuclear Information System (INIS)

    Kueffer, K.

    1994-01-01

    The report deals with the questions on reprocessing and final storage of spent fuel elements from the point of view of the Swiss. The contractual obligations were discussed, of the present situation of reprocessing and their assessment. 1 fig

  3. Reprocessing method for spent fuel

    International Nuclear Information System (INIS)

    Fujie, Makoto; Shoji, Yuichi; Kobayashi, Tsuguyuki.

    1997-01-01

    After reducing oxides of uranium (U), plutonium (Pu) and miner actinides in spent fuels by magnesium (Mg) in a molten salt, rear earth element oxides and salts of alkali metals and alkaline earth metals contained in the molten salt phase are separated and removed. Further, the Mg phase containing the reduced metals is evaporated to separate and remove Mg, thereby recovering U, Pu and minor actinides. In a lithium (Li) process, Li 2 O also generated in the reduction step is regenerated to Li simultaneously, and the reduction is conducted while suppressing the Li 2 O concentration in the molten salt low. This can improve the reduction rate of oxides of U, Pu and minor actinides compared with conventional cases. Since Li 2 O is regenerated into Li in the reduction step of the Li process, deposited Li 2 O is not carried to an electrolysis purification step, and recovering rate of U, Pu and minor actinides is not lowered. (T.M.)

  4. Reprocessing in the thorium fuel cycle

    International Nuclear Information System (INIS)

    Merz, E.

    1984-01-01

    An overview of the authors personal view is presented on open questions in regard to still required research and development work for the thorium fuel cycle before its application in a technical-industrial scale may be tackled. For a better understanding, all stations of the back-end of the thorium fuel cycle are briefly illustrated and their special features discussed. They include storage and transportation measures, all steps of reprocessing, as well as the entire radioactive waste treatment. Knowledge gaps are, as far as they are obvious, identified and proposals put forward for additional worthwile investigations. (orig.) [de

  5. Fission-product SiC reaction in HTGR fuel

    International Nuclear Information System (INIS)

    Montgomery, F.

    1981-01-01

    The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels

  6. Legal problems of nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Rossnagel, A.

    1987-01-01

    The contributions in this book are intended to exemplify the legal situation in connection with the reprocessing of spent nuclear fuel from the point of view of constitutional law, administrative law, and international law. Outline solutions are presented with regard to ensuring health, personal freedom, democratic rights and other rights, and are discussed. The author Rossnagel investigates whether the principle of essential matter can guarantee a parliamentary prerogative concerning this field of large-scale technology. The author Schmidt shows that there is no legal obligation of commitment to a reprocessing technology that would exclude research for or application of a less hazardous technology. The contribution by Baumann explains the problems presented by a technology not yet developed to maturity with regard to the outline approval of the technological concept, which is a prerequisite of any partial licence to be issued. The final contribution by Guendling investigates the duties under international law, as for instance transfrontier information, consultation, and legal protection, and how these duties can be better put into practice in order to comply the seriousness of the hazards involved in nuclear fuel reprocessing. (orig./HP) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-08-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  9. Development of processes and equipment for the refabrication of HTGR fuels

    International Nuclear Information System (INIS)

    Sease, J.D.; Lotts, A.L.

    1976-06-01

    Refabrication is in the step in the HTGR thorium fuel cycle that begins with a nitrate solution containing 238 U and culminates in the assembly of this material into fuel elements for use in an HTGR. Refabrication of HTGR fuel is essentially a manufacturing operation and consists of preparation of fuel kernels, application of multiple layers of pyrolytic carbon and SiC, preparation of fuel rods, and assembly of fuel rods in fuel elements. All the equipment for refabrication of 238 U-containing fuel must be designed for completely remote operation and maintenance in hot cell facilities. This paper describes the status of processes and equipment development for the remote refabrication of HTGR fuels. The feasibility of HTGR refabrication processes has been proven by laboratory development. Engineering-scale development is now being performed on a unit basis on the majority of the major equipment items. Engineering-scale equipment described includes full-scale resin loading equipment, a 5-in.-dia (0.13-m) microsphere coating furnace, a fuel rod forming machine, and a cure-in-place furnace

  10. FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements

    International Nuclear Information System (INIS)

    Pierce, V.H.

    2005-01-01

    1 - Description of problem or function: The FREVAP type of code for estimating the release of longer-lived metallic fission products from HTGR fuel elements has been developed to take into account the combined effects of the retention of metallic fission products by fuel particles and the rather strong absorption of these fission products by the graphite of the fuel elements. Release calculations are made on the basis that the loss of fission product nuclides such as strontium, cesium, and barium is determined by their evaporation from the graphite surfaces and their transpiration induced by the flowing helium coolant. The code is devised so that changes of fission rate (fuel element power), fuel temperature, and graphite temperature may be incorporated into the calculation. Temperature is quite important in determining release because, in general, both release from fuel particles and loss by evaporation (transpiration) vary exponentially with the reciprocal of the absolute temperature. NESC0301/02: This version differs from the previous one in the following points: The source and output files were converted from BCD to ASCII coding. 2 - Method of solution: A problem is defined as having a one-dimensional segment made up of three parts - (1) the fission product source (fuel particles) in series with, (2) a non-source and absorption part (element graphite) and (3) a surface for evaporation to the coolant (graphite-helium interface). More than one segment may be connected (possibly segments stacked axially) by way of the coolant. At any given segment, a continuity equation is solved assuming equilibrium between the source term, absorption term, evaporation at coolant interface and the partial pressure of the fission product isotope in the coolant. 3 - Restrictions on the complexity of the problem - Maxima of: 5 isotopes; 10 time intervals for time-dependent variable; 49 segments (times number of isotopes); 5 different output print time-steps

  11. Fuel reprocessing at a loss to prove its justification

    International Nuclear Information System (INIS)

    Traube, K.

    1986-01-01

    Commercial utilization of nuclear energy is possible with or without fuel reprocessing of spent fuel elements. Demands on terminal storage are about equal in both cases. There is no reason - excluding the military one - to decide in favour of fuel reprocessing instead of direct terminal storage, for neither does fuel reprocessing offer advantages in regard of the safety of nuclear waste disposal, nor is it necessary to produce plutonium for the breeder reactor. Fuel reprocessing is analyzed considering those changed aspects with a view to scarcer uranium resources, juridical motives, and what is termed the development deficit. (DG) [de

  12. Irradiation performance of HTGR fuel in HFIR experiment HRB-13

    International Nuclear Information System (INIS)

    Tiegs, T.N.

    1982-03-01

    Irradiation capsule HRB-13 tested High-Temperature Gas-Cooled Reactor (HTGR) fuel under accelerated conditions in the High Flux Isotope Reactor (HFIR) at ORNL. The ORNL part of the capsule was designed to provide definitive results on how variously misshapen kernels affect the irradiation performance of weak-acid-resin (WAR)-derived fissile fuel particles. Two batches of WAR fissile fuel particles were Triso-coated and shape-separated into four different fractions according to their deviation from spericity, which ranged from 9.6 to 29.7%. The fissile particles were irradiated for 7721 h. Heavy-metal burnups ranged from 80 to 82.5% FIMA (fraction of initial heavy-metal atoms). Fast neutron fluences (>0.18 MeV) ranged from 4.9 x 10 25 neutrons/m 2 to 8.5 x 10 25 neutrons/m 2 . Postirradiation examination showed that the two batches of fissile particles contained chlorine, presumably introduced during deposition of the SiC coating

  13. Handbook on process and chemistry on nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Atsuyuki [Tokyo Univ., Tokyo (Japan); Asakura, Toshihide; Adachi, Takeo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO{sub 2} fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  14. Handbook on process and chemistry on nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Suzuki, Atsuyuki; Asakura, Toshihide; Adachi, Takeo

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO 2 fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  15. Handbook on process and chemistry on nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Atsuyuki (ed.) [Tokyo Univ., Tokyo (Japan); Asakura, Toshihide; Adachi, Takeo (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO{sub 2} fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  16. Feasibility study for adapting ITREC plant to reprocessing LMFBR fuels

    International Nuclear Information System (INIS)

    Moccia, A.; Rolandi, G.

    1976-05-01

    The report evaluates the feasibility of adapting ITREC plant to the reprocessing LMFBR fuels, with the double purpose of: 1) recovering valuable Pu contained in these fuels and recycling it to the fabrication plant; 2) trying, on a pilot scale, the chemical process technology to be applied in a future industrial plant for reprocessing the fuel elements discharged from fast breeder power reactors

  17. Simplified probabilistic risk assessment in fuel reprocessing

    International Nuclear Information System (INIS)

    Solbrig, C.W.

    1993-01-01

    An evaluation was made to determine if a backup mass tracking computer would significantly reduce the probability of criticality in the fuel reprocessing of the Integral Fast Reactor. Often tradeoff studies, such as this, must be made that would greatly benefit from a Probably Risk Assessment (PRA). The major benefits of a complete PRA can often be accrued with a Simplified Probabilistic Risk Assessment (SPRA). An SPRA was performed by selecting a representative fuel reprocessing operation (moving a piece of fuel) for analysis. It showed that the benefit of adding parallel computers was small compared to the benefit which could be obtained by adding parallelism to two computer input steps and two of the weighing operations. The probability of an incorrect material moves with the basic process is estimated to be 4 out of 100 moves. The actual values of the probability numbers are considered accurate to within an order of magnitude. The most useful result of developing the fault trees accrue from the ability to determine where significant improvements in the process can be made. By including the above mentioned parallelism, the error move rate can be reduced to 1 out of 1000

  18. Nuclear fuel cycle: reprocessing. A bibliography

    International Nuclear Information System (INIS)

    Smith, L.B.

    1982-12-01

    This bibliography contains information on the reprocessing portion of the nuclear fuel cycle included in the Department of Energy's Energy Data Base from January 1981 through November 1982. The abstracts are grouped by subject category. Entries in the subject index also facilitate access by subject. Within each category the arrangement is by report number for reports, followed by nonreports in reverse chronological order. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes, each preceded by a brief description, are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number

  19. The fuel reprocessing plant at Wackersdorf

    International Nuclear Information System (INIS)

    Held, M.

    1986-01-01

    For a more systematic discussion about the fuel reprocessing plant at Wackersdorf, the colloquium tried to cover the most important questions put forward in the controversies: economic efficiency and energy-political needs; safety and ecological repercussions; inner safety and consequences for basic rights and the regional economic structure; majority decisions and participation of the population of the region. Elements of evaluation are the conservation of resources, health, economic efficiency, and citizens' rights of liberty. The related basic ethical questions are considered. The 18 contributions are individually recorded in the data base. (DG) [de

  20. Nuclear fuel reprocessing: A time for decision

    International Nuclear Information System (INIS)

    O'Donnell, A.J.; Sandbery, R.O.

    1983-01-01

    Availability of adequate supplies of energy at an affordable cost is essential to continued growth of the world's economics. The tie between economic growth and electricity usage is particularly strong and the pervasive wordwide trend toward increasing electrification shows no signs of abating. Very few viable alternatives are available for supplying the projected increase in baseload electric generating capacity in the next several decades, and most industrialized nations have chosen nuclear power to play a major role. Sustained growth of nuclear power can only be achieved, however, by reprocessing spent fuel to recover and utilize the residual uranium and plutonium energy values

  1. Legal questions concerning the termination of spent fuel element reprocessing

    International Nuclear Information System (INIS)

    John, Michele

    2005-01-01

    The thesis on legal aspects of the terminated spent fuel reprocessing in Germany is based on the legislation, jurisdiction and literature until January 2004. The five chapters cover the following topics: description of the problem; reprocessing of spent fuel elements in foreign countries - practical and legal aspects; operators' responsibilities according to the atomic law with respect to the reprocessing of Geman spent fuel elements in foreign countries; compatibility of the prohibition of Geman spent fuel element reprocessing in foreign countries with international law, European law and German constitutional law; results of the evaluation

  2. In-pile tests of HTGR fuel particles and fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

    1985-01-01

    Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

  3. Reprocessing

    International Nuclear Information System (INIS)

    Couture, J.; Rougeau, J.-P.

    1987-01-01

    The course of development of a comprehensive nuclear power industry has its own pace which implies the timely progressive and consistent mastery of each industrial step. In the nuclear fuel it is not surprising that the back-end services have lastly reached the industrial stage. In France, we have now fully completed the industrial demonstration of the closed fuel cycle. Our experience covers all necessary steps : transportation of spent fuel, storage, reprocessing, waste conditioning, recovered uranium recycling, plutonium recycling in thermal MOX fuels, plutonium-based fuel for FBR. While FBR development is a long term target, recycling of fissile materials in present LWR reactors appears to be a source of noticable savings. In the meantime rational management of waste material is the key for increased safety and better environment protection. Reprocessing activity is certainly the major achievement of the back-end strategy. The proven efficiency of this technique as it is implemented at La Hague facility gives the full assurance of a smooth operation of the under completion UP3 unit. The base-load management system which applies during the first ten years of its operation will make possible a noticable reduction of the commercial price for reprocessing services by the end of the century. Industrial maturity being confirmed, economic maturity is now the outstanding merit of the reprocessing and recycling strategy. It is a permanent challenge, to which the response is definitely positive in the sense of reducing the nuclear KWh production cost. (author)

  4. HTGR Fuel Technology Program. Semiannual report for the period ending March 31, 1981

    International Nuclear Information System (INIS)

    1981-05-01

    This document reports the technical accomplishments on the HTGR Fuel Technology Program at General Atomic during the first half of FY-81. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, and core component verification testing tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR with a follow-on reformer (R) version. An important effort which was initiated during this period was the preparation of input data for a long-range technology program plan

  5. HTGR Fuel-Technology Program. Semiannual report for the period ending September 30, 1982

    International Nuclear Information System (INIS)

    1982-11-01

    This document reports the technical accomplishments on the HTGR Fuel Technology Program at GA Technologies Inc. during the second half of FY-1982. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, and core component verification testing tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR with a follow-on reformer (R) version. An important effort which was completed during this period was the preparation of input data for a long-range technology program plan

  6. Operations monitoring concept. Consolidated Fuel Reprocessing Program

    International Nuclear Information System (INIS)

    Kerr, H.T.

    1985-01-01

    Operations monitoring is a safeguards concept which could be applied in future fuel cycle facilities to significantly enhance the effectiveness of an integrated safeguards system. In general, a variety of operations monitoring techniques could be developed for both international and domestic safeguards application. The goal of this presentation is to describe specific examples of operations monitoring techniques as may be applied in a fuel reprocessing facility. The operations monitoring concept involves monitoring certain in-plant equipment, personnel, and materials to detect conditions indicative of the diversion of nuclear material. An operations monitoring subsystem should be designed to monitor operations only to the extent necessary to achieve specified safeguards objectives; there is no intent to monitor all operations in the facility. The objectives of the operations monitoring subsystem include: verification of reported data; detection of undeclared uses of equipment; and alerting the inspector to potential diversion activities. 1 fig

  7. Pyroelectrochemical process for reprocessing irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Brambilla, G.; Sartorelli, A.

    1982-01-01

    A pyroelectrochemical process for reprocessing irradiated fast reactor mixed oxide or carbide fuels is described. The fuel is dissolved in a bath of molten alkali metal sulfates. The Pu(SO 4 ) 2 formed in the bath is thermally decomposed, leaving crystalline PuO 2 on the bottom of the reaction vessel. Electrodes are then introduced into the bath, and UO 2 is deposited on the cathode. Alternatively, both UO 2 and PuO 2 may be electrodeposited. The molten salts, after decontamination by precipitating the fission products dissolved in the bath by introducing basic agents such as oxides, carbonates, or hydroxides, may be recycled. Since it is not possible to remove cesium from the molten salt bath, periodic disposal and partial renewal with fresh salts is necessary. The melted salts that contain the fission products are conditioned for disposal by embedding them in a metallic matrix

  8. Method of reprocessing spent nuclear fuels

    International Nuclear Information System (INIS)

    Kamiyama, Hiroaki; Inoue, Tadashi; Miyashiro, Hajime.

    1987-01-01

    Purpose: To facilitate the storage management for the wastes resulting from reprocessing by chemically separating transuranium elements such as actionoid elements together with uranium and plutonium. Method: Spent fuels from a nuclear reactor are separated into two groups, that is, a mixture of uranium, plutonium and transuranium elements and cesium, strontium and other nuclear fission products. Virgin uranium is mixed to adjust the mixture of uranium, plutonium and transuranium elements in the first group, which is used as the fuels for the nuclear reactor. After separating to recover useful metals such as cesium and strontium are separated from short half-decay nuclear fission products of the second group, other nuclear fission products are stored and managed. This enables to shorten the storage period and safety storage and management for the wastes. (Takahashi, M.)

  9. Selection of LEU/Th reference fuel for the HTGR-SC/C lead plant

    International Nuclear Information System (INIS)

    Turner, R.F.; Neylan, A.J.; Baxter, A.M.; McEachern, D.W.; Stansfield, O.M.

    1983-05-01

    This paper describes the reference fuel materials for the high-temperature gas-cooled reactor (HTGR) plant for steam cycle/cogeneration (SC/C). A development and testing program carried out in 1978 through 1982 led to the selection of coated fuel particles of uranium-oxycarbide (UCO) for fissile materials and thorium oxide (ThO 2 ) for fertiel materials. Low-enriched uranium (LEU) is the enrichment basis for the HTGR-SC/C application. While UC 2 and UO 2 would also meet the essential criteria for fissile fuel, the UCO, alternative was selected on the basis of improved performance, economics, and process conditions

  10. Nuclear fuel reprocessing and high level waste disposal: informational hearings. Volume V. Reprocessing. Part 2

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-03-08

    Testimony was presented by a four member panel on the commercial future of reprocessing. Testimony was given on the status of nuclear fuel reprocessing in the United States. The supplemental testimony and materials submitted for the record are included in this report. (LK)

  11. Safety problems in fuel reprocessing plants

    International Nuclear Information System (INIS)

    Amaury, P.; Jouannaud, C.; Niezborala, F.

    1979-01-01

    The document first situates the reprocessing in the fuel cycle as a whole. It shows that a large reprocessing plant serves a significant number of reactors (50 for a plant of 1500 tonnes per annum). It then assesses the potential risks with respect to the environment as well as with respect to the operating personnel. The amounts of radioactive matter handled are very significant and their easily dispersible physical form represents very important risks. But the low potential energy likely to bring about this dispersion and the very severe and plentiful confinement arrangements are such that the radioactive risks are very small, both with respect to the environment and the operating personnel. The problems of the interventions for maintenance or repairs are mentioned. The intervention techniques in a radioactive environment are perfected, but they represent the main causes of operating personnel irradiation. The design principle applied in the new plants take this fact into account, involving a very significant effort to improve the reliability of the equipment and ensuring the provision of devices enabling the failing components to be replaced without causing irradiation of the personnel [fr

  12. Nuclear fuel re-processing plant

    International Nuclear Information System (INIS)

    Sasaki, Yuko; Honda, Takashi; Shoji, Saburo; Kobayashi, Shiro; Furuya, Yasumasa

    1989-01-01

    In a nuclear fuel re-processing plant, high Si series stainless steels not always have sufficient corrosion resistance in a solution containing only nitric acid at medium or high concentration. Further, a method of blowing NOx gases may possibly promote the corrosion of equipment constituent materials remarkably. In view of the above, the corrosion promoting effect of nuclear fission products is suppressed without depositing corrosive metal ions as metals in the nitric acid solution. That is, a reducing atmosphere is formed by generating NOx by electrolytic reduction thereby preventing increase in the surface potential of stainless steels. Further, an anode is disposed in the nitric acid solution containing oxidative metal ions to establish an electrical conduction and separate them by way of partition membranes and a constant potential or constant current is applied while maintaining an ionic state so as not to deposit metals. Thus, equipments of re-processing facility can be protected from corrosion with no particular treatment for wastes as radioactive materials. (K.M.)

  13. Computer simulation of HTGR fuel microspheres using a Monte-Carlo statistical approach

    International Nuclear Information System (INIS)

    Hedrick, C.E.

    1976-01-01

    The concept and computational aspects of a Monte-Carlo statistical approach in relating structure of HTGR fuel microspheres to the uranium content of fuel samples have been verified. Results of the preliminary validation tests and the benefits to be derived from the program are summarized

  14. The regulations concerning the reprocessing business of spent fuels

    International Nuclear Information System (INIS)

    1981-01-01

    This rule is stipulated under the provisions of reprocessing business in the law concerning regulation of nuclear raw materials, nuclear fuel materials and nuclear reactors and to execute them. Basic terms are defined, such as exposure radiation dose, cumulative dose, control area, security area, surrounding monitoring area, worker, radioactive waste and facility for discharging into the sea. The application for the designation for reprocessing business under the law shall include the maximum reprocessing capacities per day and per year of each kind of spent fuel, to be reprocessed and the location, structure and equipment of reprocessing facilities as specified in the regulation. Records shall be made in each works or enterprise on the inspection, operation and maintenance of reprocessing facilities, radiation control, accidents and weather, and kept for particular periods respectively. Reprocessing enterprisers shall set up control area, security area and surrounding monitoring area to restrict entrance, etc. Specified measures shall be taken by these enterprisers concerning the exposure radiation doses of workers. Reprocessing facilities shall be inspected and examined more than once a day. The regular self-inspection and operation of reprocessing facilities, the transport and storage of nuclear fuel materials, the disposal of radioactive wastes in works or enterprises where reprocessing facilities are located, and security rules are defined in detail, respectively. (Okada, K.)

  15. On transient irradiation behavior of HTGR fuel particles

    International Nuclear Information System (INIS)

    Mortenson, S.C.; Okrent, D.

    1977-01-01

    An examination of HTGR TRISO coated fuel particles was made in which the particles' stress-strain histories were determined during both steady-state and transient operating conditions. The basis for the examination was a modified version of a computer code written by Kaae which assumed spherical symmetry, isotropic thermal expansion, isotropic elastic constants, time-temperature-irradiation invariant materials properties, and steady state operation during particle exposure. Additionally, the Kaae code modelled potential separation of layers at the SiC-inner PyC interface and considered that several entrapped fission products could exist in either the gaseous or solid state, dependent upon particle operating conditions. Using the modified code which modelled transient behavior in a quasi-static fashion, a series of both steady-state and transient operating condition computer simulations was made. For the former set of runs, a candidate set of particle dimensions and a nominal set of materials' properties was assumed. Layer thicknesses were assumed to be normally distributed about the nominal thickenesses and a probability distribution of SiC tensile stresses was generated; sensitivity of the stress distribution to assumed standard deviation of the layer thicknesses was acute. Further, this series of steady-state runs demonstrated that for certain combinations of the assumed PyC-SiC bond interface strength and irradiation-induced creep constant, anomalous predicted stresses may be obtained in the PyC layers. The steady-state runs also suggest that transient behavior would most likely not be significant at fast neutron exposures below about 10 21 NVT due to both low fission gas pressure and likely beneficial interface separation

  16. Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

    International Nuclear Information System (INIS)

    Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

    1993-01-01

    Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

  17. Formation of actinides in irradiated HTGR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    dos Santos, A. M.

    1976-03-15

    Actinide nuclide concentrations of 11 spent AVR fuel elements were determined experimentally. The burnup of the spheres varied in the range between 10% and 100% fifa, the Th : U ratio was 5 : 1. The separation procedures for an actinide isolation were tested with highly irradiated ThO/sub 2/. Separation and decontamination factors are presented. Build-up of /sup 232/U was discussed. The AVR breeding rate was ascertained to be 0.5. The hazard potential of high activity waste was calculated. Actinide recovery factors were proposed in order to reduce the hazard potential of the waste by an actinide removal under consideration of the reprocessing technology which is available presently.

  18. Calcination, Reduction and Sintering of ADU Spheres for HTGR Fuel

    International Nuclear Information System (INIS)

    Jeong, Kyung Chai; Eom, Sung Ho; Kim, Yeon Ku; Kim, Woong Ki; Kim, Young Min; Lee, Young Woo; Kim, Ju Hee; Cho, Hyo Jin; Cho, Moon Seoung

    2011-01-01

    The international oil market is again in turmoil in accordance with the increasing of human needs and energy consumption. Soaring oil prices, fears of supply security, and climate change are concerned becoming more concrete make for an uncertain energy future. In this view point, nuclear energy is an important, yet controversial option for energy supply. High Temperature Gas Reactor will play a dominant role in the worldwide fleet of nuclear reactors of the next decade for electricity production and high temperature heat. HTGR have two reactor types which use the basic fuel concept based on the dispersion of TRISO coated particles in graphite in shown Fig.1. The TRISO coated particle for these purposes is prepared with pyro-carbon and silicone carbide coatings on a spherical UO 2 kernel surface as fissile material. The TRISO fuel particle consists of a microsphere (i.e., UO 2 kernel) of nuclear material: encapsulated by multiple layers of pyro-carbon and a SiC layer. This multiple coating layers system has been engineered to retain the fission products generated by fission of the nuclear material in the kernel during normal operation and all licensing basis events over the design lifetime of the fuel. UO 2 kernels are produced by using the modified sol-gel process, a wet process, generally known as the GSP method. Wet chemical processes are flexible in producing kernels of different size and chemical composition with high throughout and yield, good spherical shape, and narrow size distribution. This chemical processing route is well-known to the potential kernel fabrication processes. The principle, as set out in Fig.2, involves first of all preparing a pseudo-sol(also known as a 'broth') from an initial uranyl nitrate solution . This broth solution is obtained through addition of a number of additives, as determined by process know-how, including a soluble organic polymer, that are subsequently gels into droplets and are dispersed for ADU precipitation. The

  19. Irradiation performance of HTGR fuel rods in HFIR experiments HRB-11 and -12

    International Nuclear Information System (INIS)

    Homan, F.J.; Tiegs, T.N.; Kania, M.J.; Long, E.L. Jr.; Thoms, K.R.; Robbins, J.M.; Wagner, P.

    1980-06-01

    Capsules HRB-11 and -12 were irradiated in support of development of weak-acid-resin-derived recycle fuel for the high-enriched uranium (HEU) fuel cycle for the HTGR. Fissil fuel particles with initial oxygen-to-metal ratios between 1.0 and 1.7 performed acceptably to full burnup for HEU fuel. Particles with ratios below 1.0 showed excessive chemical interaction between rare earth fission products and the SiC layer

  20. Benefit analysis of reprocessing and recycling light water reactor fuel

    International Nuclear Information System (INIS)

    1976-12-01

    The macro-economic impact of reprocessing and recycling fuel for nuclear power reactors is examined, and the impact of reprocessing on the conservation of natural uranium resources is assessed. The LWR fuel recycle is compared with a throwaway cycle, and it is concluded that fuel recycle is favorable on the basis of economics, as well as being highly desirable from the standpoint of utilization of uranium resources

  1. Equipment specifications for an electrochemical fuel reprocessing plant

    International Nuclear Information System (INIS)

    Hemphill, Kevin P.

    2010-01-01

    Electrochemical reprocessing is a technique used to chemically separate and dissolve the components of spent nuclear fuel, in order to produce new metal fuel. There are several different variations to electrochemical reprocessing. These variations are accounted for by both the production of different types of spent nuclear fuel, as well as different states and organizations doing research in the field. For this electrochemical reprocessing plant, the spent fuel will be in the metallurgical form, a product of fast breeder reactors, which are used in many nuclear power plants. The equipment line for this process is divided into two main categories, the fuel refining equipment and the fuel fabrication equipment. The fuel refining equipment is responsible for separating out the plutonium and uranium together, while getting rid of the minor transuranic elements and fission products. The fuel fabrication equipment will then convert this plutonium and uranium mixture into readily usable metal fuel.

  2. Radiation resistance of pyrocarbon-boned fuel and absorbing elements for HTGR

    International Nuclear Information System (INIS)

    Gurin, V.A.; Konotop, Yu.F.; Odejchuk, N.P.; Shirochenkov, S.D.; Yakovlev, V.K.; Aksenov, N.A.; Kuprienko, V.A.; Lebedev, I.G.; Samsonov, B.V.

    1990-01-01

    In choosing the reactor type, problems of nuclear and radiation safety are outstanding. The analysis of the design and experiments show that HTGR type reactors helium cooled satisfy all the safety requirements. It has been planned in the Soviet Union to construct two HTGR plants, VGR-50 and VG-400. Later it was decided to construct an experimental plant with a low power high temperature reactor (VGM). Spherical uranium-graphite fuel elements with coated fuel particles are supposed to be used in HTGR core. A unique technology for producing spherical pyrocarbon-bound fuel and absorbing elements of monolithic type has been developed. Extended tests were done to to investigate fuel elements behaviour: radiation resistance of coated fuel particles with different types of fuel; influence of the coated fuel particles design on gaseous fission products release; influence of non-sphericity on coated fuel particle performance; dependence of gaseous fission products release from fuel elements on the thickness of fuel-free cans; confining role of pyrocarbon as a factor capable of diminishing the rate of fission products release; radiation resistance of spherical fuel elements during burnup; radiation resistance of spherical absorbing elements to fast neutron fluence and boron burnup

  3. Used mixed oxide fuel reprocessing at RT-1 plant

    Energy Technology Data Exchange (ETDEWEB)

    Kolupaev, D.; Logunov, M.; Mashkin, A.; Bugrov, K.; Korchenkin, K. [FSUE PA ' Mayak' , 30, Lenins str, Ozersk, 460065 (Russian Federation); Shadrin, A.; Dvoeglazov, K. [ITCP ' PRORYV' , 2/8 Malaya Krasmoselskay str, Moscow, 107140 (Russian Federation)

    2016-07-01

    Reprocessing of the mixed uranium-plutonium spent nuclear fuel of the BN-600 reactor was performed at the RT-1 plant twice, in 2012 and 2014. In total, 8 fuel assemblies with a burn-up from 73 to 89 GW day/t and the cooling time from 17 to 21 years were reprocessed. The reprocessing included the stages of dissolution, clarification, extraction separation of U and Pu with purification from the fission products, refining of uranium and plutonium at the relevant refining cycles. Dissolution of the fuel composition of MOX used nuclear fuel (UNF) in nitric acid solutions in the presence of fluoride ion has occurred with the full transfer of actinides into solution. Due to the high content of Pu extraction separation of U and Pu was carried out on a nuclear-safe equipment designed for the reprocessing of highly enriched U spent nuclear fuel and Pu refining. Technological processes of extraction, separation and refining of actinides proceeded without deviations from the normal mode. The output flow of the extraction outlets in their compositions corresponded to the regulatory norms and remained at the level of the compositions of the streams resulting from the reprocessing of fuel types typical for the RT-1 plant. No increased losses of Pu into waste have been registered during the reprocessing of BN-600 MOX UNF an compare with VVER-440 uranium UNF reprocessing. (authors)

  4. Research and development of FBR fuel reprocessing in PNC

    International Nuclear Information System (INIS)

    Hoshino, T.

    1976-05-01

    The research program of the PNC for FBR fuel reprocessing in Japan is discussed. The general characteristics of FBR fuel reprocessing are pointed out and a comparison with LWR fuel is made. The R and D program is based on reprocessing using the aqueous Purex process. So far, some preliminary steps of the research program have been carried out, these include solvent extraction test, off-gas treatment test, voloxidation process study, solidification test of high-level liquid waste, and study of the dissolution behaviour of irradiated mixed oxide fuel. By the end of the 1980s, a pilot plant for FBR fuel reprocessing will be completed. For the design of the pilot plant, further research will be carried out in the following fields: head-end techniques; voloxidation process; dissolution and extraction techniques; waste treatment techniques. A time schedule for the different steps of the program is included

  5. Fuel reprocessing experience in India: Technological and economic considerations

    International Nuclear Information System (INIS)

    Prasad, A.N.; Kumar, S.V.

    1983-01-01

    The approach to the reprocessing of irradiated fuel from power reactors in India is conditioned by the non-availability of highly enriched uranium with the consequent need for plutonium for the fast-reactor programme. With this in view, the fuel reprocessing programme in India is developing in stages matching the nuclear power programme. The first plant was set up in Trombay to reprocess the metallic uranium fuel from the research reactor CIRUS. The experience gained in the construction and operation of this plant, and in its subsequent decommissioning and reconstruction, has not only provided the know-how for the design of subsequent plants but has indicated the fruitful areas of research and development for efficient utilization of limited resources. The Trombay plant also handled successfully, on a pilot scale, the reprocessing of irradiated thorium fuel to separate uranium-233. The second plant at Tarapur has been built for reprocessing spent fuels from the power reactors at Tarapur (BWR) and Rajasthan (PHWR). The third plant, at present under design, will reprocess the spent fuels from the power reactors (PHWR) and the Fast Breeder Test Reactor (FBTR) located at Kalpakkam. Through the above approach experience has been acquired which will be useful in the design and construction of even larger plants which will become necessary in the future as the nuclear power programme grows. The strategies considered for the sizing and siting of reprocessing plants extend from the idea of small plants, located at nuclear power station sites, to a large-size central plant, located at an independent site, serving many stations. The paper discusses briefly the experience in reprocessing uranium and thorium fuels and also in decommissioning. An attempt is made to outline the technological and economic aspects which are relevant under different circumstances and which influence the size and siting of the fuel reprocessing plants and the expected lead times for construction

  6. The importance of nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Allday, C.

    1977-01-01

    The subject is discussed under the following main headings: introduction; world energy requirement; energy conservation and the economics of recycle; environmental considerations and the timescale of reprocessing; and problems associated with reprocessing. It is concluded that reprocessing is essential to the conservation of the world's energy resources and is an environmentally and probably an economically more acceptable option to the 'throw away' alternative. The associated problems of proliferation and terrorism, although of the utmost importance, can and will be solved. (U.K.)

  7. Predicting the behaviour or neptunium during nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Drake, V.A.

    1988-01-01

    Behaviour of Np and its distribution over reprocessing flowsheet is studied due to the necessity of improvement of reprocessing methods of wastes formed during purex-process. Valency states of Np in solutions of reprocessing cycles, Np distribution in organic and acid phases, Np(5) oxidation by nitric acid at the stage of extraction, effect of U and Pu presence on Np behaviour, are considered. Calculation and experimental data are compared; the possibility of Np behaviour forecasting in the process of nuclear fuel reprocessing, provided initial data vay, is shown. 7 refs.; 4 figs.; 1 tab

  8. The regulations concerning the reprocessing business of spent fuels

    International Nuclear Information System (INIS)

    1980-01-01

    The office ordinance is established under the provisions related to reprocessing businesses of the law concerning regulation of nuclear raw materials, nuclear fuel materials and reactors, to enforce the provisions. The basic terms are defined, such as exposure radiation dose; accumulated dose; controlled area; maintenance area; surrounding watch area; employee; radioactive waste; the facilities for discharge to sea. An application for the designation of reprocessing businesses shall be filed, listing the following matters: the maximum daily and yearly reprocessing capacities for each kind of spent fuel; the location and general structure of reprocessing facilities; the structures of buildings; the structure and equipments of main reprocessing facilities, the storage facilities for products and the disposal facilities for radioactive wastes; the equipments of measuring and control system facilities and radiation control facilities, etc. Records shall be made on the inspection of reprocessing facilities, radiation control, operation, maintenance, the accidents of reprocessing facilities and weather, and kept for the period from one to ten years, respectively. Any person engaging in reprocessing businesses shall set up control, maintenance and surrounding watch areas, and take specified measures to restrict the entrance of persons. The measures to be taken against exposure radiation dose, the inspection, regular independent examination and operation of reprocessing facilities and other related matters are stipulated in detail. (Okada, K.)

  9. 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)

  10. Cost and availability of gadolinium for nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Klepper, O.H.

    1985-06-01

    Gadolinium is currently planned for use as a soluble neutron poison in nuclear fuel reprocessing plants to prevent criticality of solutions of spent fuel. Gadolinium is relatively rare and expensive. The present study was undertaken therefore to estimate whether this material is likely to be available in quantities sufficient for fuel reprocessing and at reasonable prices. It was found that gadolinium, one of 16 rare earth elements, appears in the marketplace as a by-product and that its present supply is a function of the production rate of other more prevalent rare earths. The potential demand for gadolinium in a fuel reprocessing facility serving a future fast reactor industry amounts to only a small fraction of the supply. At the present rate of consumption, domestic supplies of rare earths containing gadolinium are adequate to meet national needs (including fuel reprocessing) for over 100 years. With access to foreign sources, US demands can be met well beyond the 21st century. It is concluded therefore that the supply of gadolinium will quite likely be more than adequate for reprocessing spent fuel for the early generation of fast reactors. The current price of 99.99% pure gadolinium oxide lies in the range $50/lb to $65/lb (1984 dollars). By the year 2020, in time for reprocessing spent fuel from an early generation of large fast reactors, the corresponding values are expected to lie in the $60/lb to $75/lb (1984 dollars) price range. This increase is modest and its economic impact on nuclear fuel reprocessing would be minor. The economic potential for recovering gadolinium from the wastes of nuclear fuel reprocessing plants (which use gadolinium neutron poison) was also investigated. The cost of recycled gadolinium was estimated at over twelve times the cost of fresh gadolinium, and thus recycle using current recovery technology is not economical. 15 refs., 4 figs., 11 tabs

  11. Reprocessing of nuclear fuels - status report

    International Nuclear Information System (INIS)

    Schueller, W.

    1976-01-01

    The paper gives a survey on reprocessing plants at present under construction, in operation, and planned, as well as on the most important process steps such as receipt, storage, conversion, the extraction process, purification of the end products, gaseous waste treatment and waste treatment, and repair and maintenance of reprocessing plants. An outline on operational experience with WAK follows. (HR/LN) [de

  12. Open problems in reprocessing of a molten salt reactor fuel

    International Nuclear Information System (INIS)

    Lelek, Vladimir; Vocka, Radim

    2000-01-01

    The study of fuel cycle in a molten salt reactor (MSR) needs deeper understanding of chemical methods used for reprocessing of spent nuclear fuel and preparation of MSR fuel, as well as of the methods employed for reprocessing of MSR fuel itself. Assuming that all the reprocessing is done on the basis of electrorefining, we formulate some open questions that should be answered before a flow sheet diagram of the reactor is designed. Most of the questions concern phenomena taking place in the vicinity of an electrode, which influence the efficiency of the reprocessing and sensibility of element separation. Answer to these questions would be an important step forward in reactor set out. (Authors)

  13. Factors affecting defective fraction of biso-coated HTGR fuel particles during in-block carbonization

    International Nuclear Information System (INIS)

    Caputo, A.J.; Johnson, D.R.; Bayne, C.K.

    1977-01-01

    The performance of Biso-coated thoria fuel particles during the in-block processing step of HTGR fuel element refabrication was evaluated. The effect of various process variables (heating rate, particle crushing strength, horizontal and/or vertical position in the fuel element blocks, and fuel hole permeability) on pitch coke yield, defective fraction of fuel particles, matrix structure, and matrix porosity was evaluated. Of the variables tested, only heating rate had a significant effect on pitch coke yield while both heating rate and particle crushing strength had a significant effect on defective fraction of fuel particles

  14. Formation of actinides in irradiated HTGR fuel elements

    International Nuclear Information System (INIS)

    Santos, A.M. dos.

    1976-03-01

    Actinide nuclide concentrations of 11 spent AVR fuel elements were determined experimentally. The burnup of the spheres varied in the range between 10% and 100% fifa, the Th : U ratio was 5 : 1. The separation procedures for actinide isolation were tested with highly irradiated ThO 2 . Separation and decontamination factors are presented. Actinide nuclide formation can be described by exponential functions of the type ln msub(nuclide) = A + B x % fifa. The empirical factors A and B were calculated performing a least squares analysis. Build-up of 232 U was discussed. According to the experimental results, 232 U is mainly produced from 230 Th, a certain amount (e.g. about 20% at a 10 5 MWd/t burnup) originated from a (n,2n) reaction of 233 U; a formation from 233 Th by a (n,2n) followed by a (n,γ) reaction was not observed. The AVR breeding rate was ascertained to be 0.5. The hazard potential of high activity waste was calculated. After a 1,000 years' storage time, the elements Pa, Am and Cm will no longer influence the total hazard index. Actinide recovery factors were proposed in order to reduce the hazard potential of the waste by an actinide removal in consideration of the reprocessing technology which is available presently. (orig.) [de

  15. Reprocessing of research reactor fuel the Dounreay option

    Energy Technology Data Exchange (ETDEWEB)

    Cartwright, P.

    1997-08-01

    Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.

  16. Strategy to support HTGR fuel for the 10 MW Indonesia’s experimental power reactor (RDE)

    International Nuclear Information System (INIS)

    Taswanda Taryo; Geni Rina Sunaryo; Ridwan; Meniek Rachmawati

    2018-01-01

    The Indonesia’s 10 MW experimental power reactor (RDE) is developed based on high temperature gas-cooled reactor (HTGR) and the program of the RDE was firstly introduced to the Agency for National Development Planning (BAPPENAS) at the beginning of 2014. The RDE program is expected to have positive impacts on community prosperity, self-reliance and sovereignty of Indonesia. The availability of RDE will be able to accelerate advanced nuclear power technology development and hence elevate Indonesia to be the nuclear champion in the ASEAN region. The RDE is expected to be operable in 2022/2023. In terms of fuel supply for the reactor, the first batch of RDE fuel will be inclusive in the RDE engineering, procurement and construction (RDE-EPC) contract for the assurance of the RDE reactor operation from 2023 to 2027. Consideration of RDE fuel plant construction is important as RDE can be the basis for the development of reactors of similar type with small-medium power(25 MWe–200/300 MWe), which are preferable for eastern part of Indonesia. To study the feasibility of the construction of RDE fuel plant, current state of the art of the R&D on HTGR fuel in some advanced countries such as European countries, the United States, South Africa and Japan will be discussed and overviewed to draw a conclusion about the prospective countries for supporting the fuel for long-term RDE operation. The strategy and road map for the preparation of the RDE fuel plant construction with the involvement of national stake holders have been developed. The best possible vendor country to support HTGR fuel for long-term operation is finally accomplished. In the end, this paper can be assigned as a reference for the planning and construction of HTGR RDE fuel fabrication plant in Indonesia. (author)

  17. Capability of minor nuclide confinement in fuel reprocessing

    International Nuclear Information System (INIS)

    Fujine, Sachio; Uchiyama, Gunzo; Mineo, Hideaki; Kihara, Takehiro; Asakura, Toshihide

    1999-01-01

    Experiment with spent fuels has started with the small scale reprocessing facility in NUCEF-BECKY αγ cell. Primary purpose of the experiment is to study the capability of long-lived nuclide confinement both in the PUREX flow sheet applied to the large scale reprocessing plant and also in the PARC (Partitioning Conundrum key process) flow sheet which is our proposal as a simplified reprocessing of one cycle extraction system. Our interests in the experiment are the behaviors of minor long-lived nuclides and the behaviors of the heterogeneous substances, such as sedimentation in the dissolver, organic cruds in the extraction banks. The significance of those behaviors will be assessed from the standpoint of the process safety of reprocessing for high burn-up fuels and MOX fuels. (author)

  18. Thermal design and analysis of the HTGR fuel element vertical carbonizing and annealing furnace

    International Nuclear Information System (INIS)

    Llewellyn, G.H.

    1977-06-01

    Computer analyses of the thermal design for the proposed HTGR fuel element vertical carbonizing and annealing furnace were performed to verify its capability and to determine the required power input and distribution. Although the furnace is designed for continuous operation, steady-state temperature distributions were obtained by assuming internal heat generation in the fuel elements to simulate their mass movement. The furnace thermal design, the analysis methods, and the results are discussed herein

  19. Simulation of nuclear fuel reprocessing for safeguards

    International Nuclear Information System (INIS)

    Canty, M.J.; Dayem, H.A.; Kern, E.A.; Spannagel, G.

    1983-11-01

    For safeguarding the chemical process area of future reprocessing plants the near-real-time material accountancy (NRTMA) method might be applied. Experimental data are not yet available for testing the capability of the NRTMA method but can be simulated using a digital computer. This report describes the mathematical modeling of the Pu-bearing components of reprocessing plants and presents first results obtained by simulation models. (orig.) [de

  20. Review of monitoring instruments for transuranics in fuel fabrication and reprocessing plants. A progress report to the physical and technological programs, Division of Biomedical and Environmental Research, U.S. Energy Research and Development Administration

    International Nuclear Information System (INIS)

    Kordas, J.F.; Phelps, P.L.

    A comprehensive review of the monitoring instruments for transuranic elements released from nuclear fuel fabrication and reprocessing plants has been compiled. The extent of routine operational releases has been reviewed for the light water reactor (LWR) fuel cycle (including plutonium recycle), the breeder reactor fuel cycle, and the high-temperature gas cooled reactor (HTGR) fuel cycle. The stack monitoring instrumentation presently in use at the various fabrication and reprocessing plants around the country is discussed. Sampling difficulties and the effectiveness of the entire sampling system are reviewed, as are the measurement problems for alpha-emitting, long-lived, transuranic aerosols, 129 I, 106 Ru, and tritium oxide. The potential problems in the HTGR fuel cycle such as the measurement of releases of alpha-emitting aerosols and of gaseous releases of 220 Rn and 14 C are also considered

  1. Technology development of fast reactor fuel reprocessing technology in India

    International Nuclear Information System (INIS)

    Natarajan, R.; Raj, Baldev

    2009-01-01

    India is committed to the large scale induction of fast breeder reactors beginning with the construction of 500 MWe Prototype Fast Breeder Reactor, PFBR. Closed fuel cycle is a prerequisite for the success of the fast reactors to reduce the external dependence of the fuel. In the Indian context, spent fuel reprocessing, with as low as possible out of pile fissile inventory, is another important requirement for increasing the share in power generation through nuclear route as early as possible. The development of this complex technology is being carried out in four phases, the first phase being the developmental phase, in which major R and D issues are addressed, while the second phase is the design, construction and operation of a pilot plant, called CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell. The third phase is the construction and operation of Demonstration of Fast Reactor Fuel Reprocessing Plant (DFRP) which will provide experience in fast reactor fuel reprocessing with high availability factors and plant throughput. The design, construction and operation of the commercial plant (FRP) for reprocessing of PFBR fuel is the fourth phase, which will provide the requisite confidence for the large scale induction of fast reactors

  2. Study on erbium loading method to improve reactivity coefficients for low radiotoxic spent fuel HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Y., E-mail: fukaya.yuji@jaea.go.jp; Goto, M.; Nishihara, T.

    2015-11-15

    Highlights: • We attempted and optimized erbium loading methods to improve reactivity coefficients for LRSF-HTGR. • We elucidated the mechanism of the improvements for each erbium loading method by using the Bondarenko approach. • We concluded the erbium loading method by embedding into graphite shaft is preferable. - Abstract: Erbium loading methods are investigated to improve reactivity coefficients of Low Radiotoxic Spent Fuel High Temperature Gas-cooled Reactor (LRSF-HTGR). Highly enriched uranium is used for fuel to reduce the generation of toxicity from uranium-238. The power coefficients are positive without the use of any additive. Then, the erbium is loaded into the core to obtain negative reactivity coefficients owing to the large resonance the peak of neutron capture reaction of erbium-167. The loading methods are attempted to find the suitable method for LRSF-HTGR. The erbium is mixed in a CPF fuel kernel, loaded by binary packing with fuel particles and erbium particles, and embedded into the graphite shaft deployed in the center of the fuel compact. It is found that erbium loading causes negative reactivity as moderator temperature reactivity, and from the viewpoint of heat transfer, it should be loaded into fuel pin elements for pin-in-block type fuel. Moreover, the erbium should be incinerated slowly to obtain negative reactivity coefficients even at the End Of Cycle (EOC). A loading method that effectively causes self-shielding should be selected to avoid incineration with burn-up. The incineration mechanism is elucidated using the Bondarenko approach. As a result, it is concluded that erbium embedded into graphite shaft is preferable for LRSF-HTGR to ensure that the reactivity coefficients remain negative at EOC.

  3. Handbook on process and chemistry of nuclear fuel reprocessing. 3rd edition

    International Nuclear Information System (INIS)

    2015-03-01

    The fundamental data on spent nuclear fuel reprocessing and related chemistry was collected and summarized as a new edition of 'Handbook on Process and Chemistry of Nuclear Fuel Reprocessing'. The purpose of this handbook is contribution to development of the fuel reprocessing and fuel cycle technology for uranium fuel and mixed oxide fuel utilization. Contents in this book was discussed and reviewed by specialists of science and technology on fuel reprocessing in Japan. (author)

  4. Spent fuel reprocessing past experience and future prospects

    International Nuclear Information System (INIS)

    Megy, J.

    1983-09-01

    A large experience has been gathered from the early fifties till now in the field of spent fuel reprocessing. As the main efforts in the world have been made for developping the reactors and the fuel fabrication industry to feed them, the spent fuel reprocessing activities came later and have not yet reached the industrial maturity existing to day for plants such as PWRs. But in the principal nuclear countries spent fuel reprocessing is to day considered as a necessity with two simultaneous targets: 1. Recovering the valuable materials, uranium and plutonium. 2. Conditionning the radioactive wastes to ensure safe definitive storage. The paper reviews the main steps: 1. Reprocessing for thermal reactor fuels: large plants are already operating or in construction, but in parallel a large effort of R and D is still under way for improvements. 2. The development of fast breeder plants implies associated fuel reprocessing facilities: pilot plants have demonstrated the closing of the cycle. The main difficulties encountered will be examined and particularly the importance of taking into account the problems of effluents processing and wastes storage [fr

  5. HTGR fuel development: loading of uranium on carboxylic acid cation-exchange resins using solvent extraction of nitrate

    International Nuclear Information System (INIS)

    Haas, P.A.

    1975-09-01

    The reference fuel kernel for recycle of 233 U to HTGR's (High-Temperature Gas-Cooled Reactors) is prepared by loading carboxylic acid cation-exchange resins with uranium and carbonizing at controlled conditions. The purified 233 UO 2 (NO 3 ) 2 solution from a fuel reprocessing plant contains excess HNO 3 (NO 3 - /U ratio of approximately 2.2). The reference flowsheet for a 233 U recycle fuel facility at Oak Ridge uses solvent extraction of nitrate by a 0.3 M secondary amine in a hydrocarbon diluent to prepare acid-deficient uranyl nitrate. This nitrate extraction, along with resin loading and amine regeneration steps, was demonstrated in 14 runs. No significant operating difficulties were encountered. The process is controlled via in-line pH measurements for the acid-deficient uranyl nitrate solutions. Information was developed on pH values for uranyl nitrate solution vs NO 3 - /U mole ratios, resin loading kinetics, resin drying requirements, and other resin loading process parameters. Calculations made to estimate the capacities of equipment that is geometrically safe with respect to control of nuclear criticality indicate 100 kg/day or more of uranium for single nitrate extraction lines with one continuous resin loading contactor or four batch loading contactors. (auth)

  6. Application of electrochemical techniques in fuel reprocessing- an overview

    Energy Technology Data Exchange (ETDEWEB)

    Rao, M K; Bajpai, D D; Singh, R K [Power Reactor Fuel Reprocessing Plant, Tarapur (India)

    1994-06-01

    The operating experience and development work over the past several years have considerably improved the wet chemical fuel reprocessing PUREX process and have brought the reprocessing to a stage where it is ready to adopt the introduction of electrochemical technology. Electrochemical processes offer advantages like simplification of reprocessing operation, improved performance of the plant and reduction in waste volume. At Power Reactor Fuel Reprocessing plant, Tarapur, work on development and application of electrochemical processes has been carried out in stages. To achieve plant scale application of these developments, a new electrochemical cycle is being added to PUREX process at PREFRE. This paper describes the electrochemical and membrane cell development activities carried out at PREFRE and their current status. (author). 5 refs., 4 tabs.

  7. Evironmental assessment factors relating to reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

    1978-05-01

    This document is in two parts. Part I presents the criteria and evaluation factors, based primarily on US experience, which may be used to carry out an environmental assessment of spent fuel reprocessing. The concept of As Low as is Reasonably Achievable (ALARA) is introduced in limiting radiation exposure. The factors influencing both occupational and general public radiation exposure are reviewed. Part II provides information on occupational and general public radiation exposure in relation to reprocessing taken from various sources including UNSCEAR and GESMO. Some information is provided in relation to potential accidents at reprocessing or MOX fuel refabrication plants. The magnitude of the services, energy, land use and non-radiological effluents for the reference design of reprocessing plant are also presented

  8. The regulations concerning the reprocessing business of spent fuels

    International Nuclear Information System (INIS)

    1979-01-01

    The regulations are defined under provisions concerning the reprocessing business in the law for the regulations of nuclear source materials, nuclear fuel materials and reactors. The basic concepts and terms are explained, such as: exposure dose; accumulative dose; controlled area; safeguarded area; inspected surrounding area; employee; radioactive waste and marine discharging facilities. Any person who gets permission for design of reprocessing facilities and method of the construction shall file an application, listing name and address of the person and the works or the place of enterprise where reprocessing facilities are to be set up, design of such facilities and method of the construction, in and out-put chart of nuclear fuel materials in reprocessing course, etc. Records shall be made and kept for particularly periods in each works or enterprise on inspection of reprocessing facilities, control of dose, operation, maintenance, accident of reprocessing facilities and weather. Detailed prescriptions are settled on entrance limitation to controlled area, exposure dose, inspection and check, regular independent examination and operation of reprocessing facilities, transportation in the works or the enterprise, storage, disposal, safeguard and measures in dangerous situations, etc. Reports shall be filed on exposure dose of employees and other specified matters in the forms attached and in the case otherwise defined. (Okada, K.)

  9. Fuel temperature prediction during high burnup HTGR fuel irradiation test. US-JAERI irradiation test for HTGR fuel

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.

    1995-01-01

    This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)

  10. Roles of programmable logic controllers in fuel reprocessing plants

    International Nuclear Information System (INIS)

    Mishra, Hrishikesh; Balakrishnan, V.P.; Pandya, G.J.

    1999-01-01

    Fuel charging facility is another application of Programmable Logic Controllers (PLC) in fuel reprocessing plants, that involves automatic operation of fuel cask dolly, charging motor, pneumatic doors, clutches, clamps, stepper motors and rod pushers in a pre-determined sequence. Block diagram of ACF system is given for underlining the scope of control and interlocks requirements involved for automation of the fuel charging system has been provided for the purpose at KARP Plant, Kalpakkam

  11. Fuel reprocessing and waste management in the UK

    International Nuclear Information System (INIS)

    Heafield, W.; Griffin, N.L.

    1994-01-01

    The currently preferred route for the management of irradiated fuel in the UK is reprocessing. This paper, therefore, concentrates on outlining the policies, practices and achievement of British Nuclear Fuels plc (BNFL) associated with the management of its irradiated fuel facilities at Sellafield. The paper covers reprocessing and how the safe management of each of the major waste categories is achieved. BNFL's overall waste management policy is to develop, in close consultation with the regulatory authorities, a strategy to minimize effluent discharges and provide a safe, cost effective method of treating and preparing for disposal all wastes arising on the site

  12. Reprocessing RTR fuel in the La Hague plants

    International Nuclear Information System (INIS)

    Thomasson, J.; Drain, F.; David, A.

    2001-01-01

    Starting in 2006, research reactors operators will be fully responsible for the back-end management of their spent fuel. It appears that the only solution for this management is treatment-conditioning, which could be done at the La Hague reprocessing complex in France. The fissile material can be separated in the reprocessing plants and the final waste can be encapsulated in a matrix adapted to its potential hazards. RTR reprocessing at La Hague would require some modifications, since the plant had been primarily designed to reprocess fuel from light water reactors. Many provisions have been taken at the plant design stage, however, and the modifications would be feasible even during active operations, as was done from 1993 to 1995 when a new liquid waste management was implemented, and when one of the two vitrification facilities was improved. To achieve RTR back-end management, COGEMA and its partners are also conducting R and D to define a new generation of LEU fuel with performance characteristics approximating those of HEU fuel. This new-generation fuel would be easier to reprocess. (author)

  13. Reprocessing RTR fuel in the La Hague plants

    Energy Technology Data Exchange (ETDEWEB)

    Thomasson, J. [Cogema, F-78140 Velizy (France); Drain, F.; David, A. [SGN, F-78182 Saint Quentin en Yvelines (France)

    2001-07-01

    Starting in 2006, research reactors operators will be fully responsible for the back-end management of their spent fuel. It appears that the only solution for this management is treatment-conditioning, which could be done at the La Hague reprocessing complex in France. The fissile material can be separated in the reprocessing plants and the final waste can be encapsulated in a matrix adapted to its potential hazards. RTR reprocessing at La Hague would require some modifications, since the plant had been primarily designed to reprocess fuel from light water reactors. Many provisions have been taken at the plant design stage, however, and the modifications would be feasible even during active operations, as was done from 1993 to 1995 when a new liquid waste management was implemented, and when one of the two vitrification facilities was improved. To achieve RTR back-end management, COGEMA and its partners are also conducting R and D to define a new generation of LEU fuel with performance characteristics approximating those of HEU fuel. This new-generation fuel would be easier to reprocess. (author)

  14. Reprocessing RTR fuel in the La Hague plants

    Energy Technology Data Exchange (ETDEWEB)

    Thomasson, J. [Cogema, 78 - Velizy Villacoublay (France); Drain, F.; David, A. [SGN, 78 - Saint Quentin en Yveline (France)

    2001-07-01

    Starting in 2006, research reactors operators will be fully responsible for their research and testing reactors spent fuel back-end management. It appears that the only solution for this management is treatment-conditioning, which could be done at the La Hague reprocessing complex in France. The fissile material can be separated in the reprocessing plants and the final waste can be encapsulated in a matrix adapted to its potential hazards. RTR reprocessing at La Hague would require some modifications, since the plant had been primarily designed to reprocess fuel from light water reactors. Many provisions have been taken at the plant design stage, however, and the modifications would be feasible even during active operations, as was done from 1993 to 1995 when a new liquid waste management was implemented, and when one of the two vitrification facilities was improved. To achieve RTR back-end management, COGEMA and its partners are also conducting R and D to define a new generation of LEU fuel with performance characteristics approximating those of HEU fuel. This new-generation fuel would be easier to reprocess. (author)

  15. Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

    International Nuclear Information System (INIS)

    McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; Reif, Tyler J.; Morris, Robert N.; Hunn, John D.

    2017-01-01

    Highlights: • The minimum required uranium carbide content for HTGR UCO fuel kernels is calculated. • More nuclear and chemical factors have been included for more useful predictions. • The effect of transmutation products, like Pu and Np, on the oxygen distribution is included for the first time. - Abstract: Three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from O release when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. In the HTGR kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium apart from UO 2 in the form of a carbide, UC x and this fuel form is designated UCO. Here general oxygen balance formulas were developed for calculating the minimum UC x content to ensure negligible CO formation for 15.5% enriched UCO taken to 16.1% actinide burnup. Required input data were obtained from CALPHAD (CALculation of PHAse Diagrams) chemical thermodynamic models and the Serpent 2 reactor physics and depletion analysis tool. The results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmuted Pu and Np oxides on the oxygen distribution as the fuel kernel composition evolves with burnup.

  16. Study on the possibility of supercritical fluid extraction for reprocessing of spent nuclear fuel from high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Duan Wuhua; Zhu Liyang; Zhu Yongjun; Xu Jingming

    2011-01-01

    International interest in high temperature gas-cooled reactor (HTGR) has been increasing in recent years. It is important to study on reprocessing of spent nuclear fuel from HTGR for recovery of nuclear resource and reduction of nuclear waste. Treatment of UO 2 pellets for preparing fuel elements of the 10 MW high temperature gas-cooled reactor (HTR-10) using supercritical fluid extraction was investigated. UO 2 pellets are difficult to be directly dissolved and extracted with TBP-HNO 3 complex in supercritical CO 2 (SC-CO 2 ), and the extraction efficiency is only about 7% under experimental conditions. UO 2 pellets are also difficult to be converted completely into nitrate with N 2 O 4 . When UO 2 pellets break spontaneously into U 3 O 8 powders with particle size below 100 μm under O 2 flow and 600degc, the extraction efficiency of U 3 O 8 powders with TBP-HNO 3 complex in SC-CO 2 can reach more than 98%. U 3 O 8 powders are easy to be completely converted into nitrate with N 2 O 4 . The extraction efficiency of the nitrate product with TBP in SC-CO 2 can reach more than 99%. So it has a potential prospect that application of supercritical fluid extraction in reprocessing of spent nuclear fuel from HTGR. (author)

  17. Sustainable and safe energy supply with seawater uranium fueled HTGR and its economy

    International Nuclear Information System (INIS)

    Fukaya, Y.; Goto, M.

    2017-01-01

    Highlights: • We discussed uranium resources with an energy security perspective. • We concluded seawater uranium is preferable for sustainability and energy security. • We evaluated electricity generation cost of seawater uranium fueled HTGR. • We concluded electricity generation with seawater uranium is reasonable. - Abstract: Sustainable and safe energy supply with High Temperature Gas-cooled Reactor (HTGR) fueled by uranium from seawater have been investigated and discussed. From the view point of safety feature of self-regulation with thermal reactor of HTGR, the uranium resources should be inexhaustible. The seawater uranium is expected to be alternative resources to conventional resources because it exists so much in seawater as a solute. It is said that 4.5 billion tons of uranium is dissolved in the seawater, which corresponds to a consumption of approximately 72 thousand years. Moreover, a thousand times of the amount of 4.5 trillion tU of uranium, which corresponds to the consumption of 72 million years, also is included in the rock on the surface of the sea floor, and that is also recoverable as seawater uranium because uranium in seawater is in an equilibrium state with that. In other words, the uranium from seawater is almost inexhaustible natural resource. However, the recovery cost with current technology is still expensive compared with that of conventional uranium. Then, we assessed the effect of increase in uranium purchase cost on the entire electricity generation cost. In this study, the economy of electricity generation of cost of a commercial HTGR was evaluated with conventional uranium and seawater uranium. Compared with ordinary LWR using conventional uranium, HTGR can generate electricity cheaply because of small volume of simple direct gas turbine system compared with water and steam systems of LWR, rationalization by modularizing, and high thermal efficiency, even if fueled by seawater uranium. It is concluded that the HTGR

  18. Analytical chemistry needs for nuclear safeguards in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Hakkila, E.A.

    1977-01-01

    A fuel reprocessing plant designed to process 1500 tons of light water reactor fuel per year will recover 15 tons of Pu during that time, or approximately 40 to 50 kg of Pu per day. Conventional nuclear safeguards accountability has relied on batch accounting at the head and tail ends of the reprocessing plant with semi-annual plant cleanout to determine in-process holdup. An alternative proposed safeguards system relies on dynamic material accounting whereby in-line NDA and conventional analytical techniques provide indications on a daily basis of SNM transfers into the system and information of Pu holdup within the system. Some of the analytical requirements and problems for dynamic materials accounting in a nuclear fuel reprocessing plant are described. Some suggestions for further development will be proposed

  19. Irradiation Performance of HTGR Fuel in WWR-K Research Reactor

    International Nuclear Information System (INIS)

    Ueta, Shohei; Sakaba, Nariaki; Shaimerdenov, Asset; Gizatulin, Shamil; Chekushina, Lyudmila; Chakrov, Petr; Honda, Masaki; Takahashi, Masashi; Kitagawa, Kenichi

    2014-01-01

    A capsule irradiation test with the high temperature gas-cooled reactor (HTGR) fuel is being carried out using WWR-K research reactor in the Institute of Nuclear Physics of the Republic of Kazakhstan (INP) to attain 100 GWd/t-U of burnup under normal operating condition of a practical small-sized HTGR. This is the first HTGR fuel irradiation test for INP in Kazakhstan collaborated with Japan Atomic Energy Agency (JAEA) in frame of International Science and Technology Center (ISTC) project. In the test, TRISO coated fuel particle with low-enriched UO_2 (less than 10 % of "2"3"5U) is used, which was newly designed by JAEA to extend burnup up to 100 GWd/t-U comparing with that of the HTTR (33 GWd/t-U). Both TRISO and fuel compact as the irradiation test specimen were fabricated in basis of the HTTR fuel technology by Nuclear Fuel Industries, Ltd. in Japan. A helium-gas-swept capsule and a swept-gas sampling device installed in WWR-K were designed and constructed by INP. The irradiation test has been started in October 2012 and will be completed up to the end of February 2015. The irradiation test is in the progress up to 69 GWd/t of burnup, and integrity of new TRISO fuel has been confirmed. In addition, as predicted by the fuel design, fission gas release was observed due to additional failure of as-fabricated SiC-defective fuel. (author)

  20. French experience and prospects in the reprocessing of fast breeder reactor fuels

    International Nuclear Information System (INIS)

    Megy, J.

    1983-06-01

    Experience acquired in France in the field of reprocessing spent fuels from fast breeder reactors is recalled. Emphasis is put on characteristics and quantities of spent fuels reprocessed in La Hague and Marcoule facilities. Then reprocessing developments with the realisation of the new pilot plant TOR at Marcoule, new equipments and study of industrial reprocessing units are reviewed [fr

  1. Evaluation of subcritical hybrid systems loaded with reprocessed fuel

    International Nuclear Information System (INIS)

    Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L.

    2015-01-01

    Highlights: • Accelerator driven systems (ADS) and fusion–fission systems are investigated for transmutation and fuel regeneration. • The calculations were performed using Monteburns code. • The results indicate the most suitable system for achieve transmutation. - Abstract: Two subcritical hybrid systems containing spent fuel reprocessed by Ganex technique and spiked with thorium were submitted to neutron irradiation of two different sources: ADS (Accelerator-driven subcritical) and Fusion. The aim is to investigate the nuclear fuel evolution using reprocessed fuel and the neutronic parameters under neutron irradiation. The source multiplication factor and fuel depletion for both systems were analysed during 10 years. The simulations were performed using MONTEBURNS code (MCNP/ORIGEN). The results indicate the main differences when irradiating the fuel with different neutron sources as well as the most suitable system for achieving transmutation

  2. Safety aspects in fuel reprocessing and radioactive waste management

    International Nuclear Information System (INIS)

    Agarwal, K.

    2018-01-01

    Nuclear energy is used for generation of electricity and for production of a wide range of radionuclides for use in research and development, healthcare and industry. Nuclear industry uses nuclear fission as source of energy so a large amount of energy is available from very small amount of fuel. As India has adopted c losed fuel cycle , spent nuclear fuel from nuclear reactor is considered as a material of resource and reprocessed to recovery valuable fuel elements. Main incentive of reprocessing is to use the uranium resources effectively by recovering/recycling Pu and U present in the spent fuel. This finally leads to a very small percentage of residual material present in spent nuclear fuel requiring their management as radioactive waste. Another special feature of the Indian Atomic Energy Program is the attention paid from the very beginning to the safe management of radioactive waste

  3. The chemical stability of TRISO-coated HTGR fuel. Pt. 1. Status report

    International Nuclear Information System (INIS)

    Groot, P.; Cordfunke, E.H.P.; Konings, R.J.M.

    1994-12-01

    The US fuel seemed to be more difficult to produce than the German fuel. Also the chemical stability of this fuel must be investigated. The conditions are more severe in the US concept than in the German concept. Oxidation of the graphite seems to be no problem, according to US HTGR concept. A ZrC coating seems to have a number of advantages with regard to the SiC coating: (1) Better retention, (2) no reaction with Pd, (3) no thermal dissociation. Only the oxidation resistance is worse than SiC. Also the maximum stress must be determined that the ZrC coating can have. (orig./HP)

  4. Survey of economics of spent nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Valvoda, Z.

    1976-01-01

    Literature data are surveyed on the economic problems of reprocessing spent fuel from light-water reactors in the period 1970 to 1975 and on the capacity of some reprocessing plants, such as NFS, Windscale, Marcoule, etc. The sharp increase in capital and production costs is analyzed and the future trend is estimated. The question is discussed of the use of plutonium and the cost thereof. The economic advantageousness previously considered to be the primary factor is no longer decisive due to new circumstances. The main objective today is to safeguard uninterrupted operation of nuclear power plants and the separation of radioactive wastes from the fuel cycle and the safe disposal thereof. (Oy)

  5. Reprocessing free nuclear fuel production via fusion fission hybrids

    Energy Technology Data Exchange (ETDEWEB)

    Kotschenreuther, Mike, E-mail: mtk@mail.utexas.edu [Intitute for Fusion Studies, University of Texas at Austin (United States); Valanju, Prashant; Mahajan, Swadesh [Intitute for Fusion Studies, University of Texas at Austin (United States)

    2012-05-15

    Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively 'new' cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th{sup 232}-U{sup 233} conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO{sub 2} matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U{sup 235} fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

  6. Reprocessing free nuclear fuel production via fusion fission hybrids

    International Nuclear Information System (INIS)

    Kotschenreuther, Mike; Valanju, Prashant; Mahajan, Swadesh

    2012-01-01

    Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th 232 –U 233 conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO 2 matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U 235 fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

  7. Quantification of TRISO fuel heterogeneity effects in HTGR lattice physics calculations

    International Nuclear Information System (INIS)

    Perfetti, C. M.; Anghaie, S.; Dugan, E.; Marcille, T.

    2010-01-01

    A large number of LEU-MHR fuel compact models were generated with randomly distributed TRISO particle fuel and were simulated using MCNP5, and it was determined how several neutronic parameters, including k-infinite, the thermal and fast diffusion coefficients, and the four factors, varied across the randomly-generated cases. A sensitivity study was also performed to determine how the four factors depend on the definition of the thermal energy group. Values of k-infinite for the cases had a sample standard deviation of 248 pcm and were found to follow an approximately normal distribution about the mean value of k-infinite. Although all of the four factors were found to have similar sample standard deviations, the resonance escape probability was found to be the most variable parameter with a sample relative standard deviation between 0.07% and 0.08%. HTGR fuel compact homogenization methods typically examine only one reference fuel compact that contains a uniform distribution of TRISO particles, but in reality the TRISO particles are randomly distributed throughout the fuel compact. Thus, the neutronic parameters for actual fuel compacts differ randomly from those in the reference model. To license next-generation High-Temperature Gas Reactors engineers must quantify all uncertainties of the design and this random variation in neutron parameters is a previously unmeasured quantity; this study measures this uncertainty by examining the variation in k-infinite for HTGR fuel compact models with randomly distributed TRISO fuel. (authors)

  8. Cost analysis of the US spent nuclear fuel reprocessing facility

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, E.A.; Deinert, M.R. [Department of Mechanical Engineering, University of Texas, Austin TX (United States); Cady, K.B. [Department of Theoretical and Applied Mechanics, Cornell University, Ithaca NY (United States)

    2009-09-15

    The US Department of Energy is actively seeking ways in which to delay or obviate the need for additional nuclear waste repositories beyond Yucca Mountain. All of the realistic approaches require the reprocessing of spent nuclear fuel. However, the US currently lacks the infrastructure to do this and the costs of building and operating the required facilities are poorly established. Recent studies have also suggested that there is a financial advantage to delaying the deployment of such facilities. We consider a system of government owned reprocessing plants, each with a 40 year service life, that would reprocess spent nuclear fuel generated between 2010 and 2100. Using published data for the component costs, and a social discount rate appropriate for intergenerational analyses, we establish the unit cost for reprocessing and show that it increases slightly if deployment of infrastructure is delayed by a decade. The analysis indicates that achieving higher spent fuel discharge burnup is the most important pathway to reducing the overall cost of reprocessing. The analysis also suggests that a nuclear power production fee would be a way for the US government to recover the costs in a manner that is relatively insensitive to discount and nuclear power growth rates. (author)

  9. Design and operation of equipment used to develop remote coating capability for HTGR fuel particles

    International Nuclear Information System (INIS)

    Suchomel, R.R.; Stinton, D.P.; Preston, M.K.; Heck, J.L.; Bolfing, B.J.; Lackey, W.J.

    1978-12-01

    Refabrication of HTGR fuels is a manufacturing process that consists of preparation of fuel kernels, application of multiple layers of pyrolytic carbon and silicon carbide, preparation of fuel rods, and assembly of fuel rods into fuel elements. All the equipment for refabrication of 233 U-containing fuel must be designed for completely remote operation and maintenance in hot-cell facilities. Equipment to remotely coated HTGR fuel particles has been designed and operated. Although not all of the equipment development needed for a fully remote coating system has been completed, significant progress has been made. The most important component of the coating furnace is the gas distributor, which must be simple, reliable, and easily maintainable. Techniques for loading and unloading the coater and handling microspheres have been developed. An engineering-scale system, currently in operation, is being used to verify the workability of these concepts. Coating crucible handling components are used to remove the crucible from the furnace, remove coated particles, and exchange the crucible, if necessary. After the batch of particles has been unloaded, it is transferred, weighed, and sampled. The components used in these processes have been tested to ensure that no particle breakage or holdup occurs. Tests of the particle handling system have been very encouraging because no major problems have been encountered. Instrumentation that controls the equipment performed very smoothly and reliably and can be operated remotely

  10. Uranium loss from BISO-coated weak-acid-resin HTGR fuel

    International Nuclear Information System (INIS)

    Pearson, R.L.; Lindemer, T.B.

    1977-02-01

    Recycle fuel for the High-Temperature Gas-Cooled Reactor (HTGR) contains a weak-acid-resin (WAR) kernel, which consists of a mixture of UC 2 , UO 2 , and free carbon. At 1900 0 C, BISO-coated WAR UC 2 or UC 2 -UO 2 kernels lose a significant portion of their uranium in several hundred hours. The UC 2 decomposes and uranium diffuses through the pyrolytic coating. The rate of escape of the uranium is dependent on the temperature and the surface area of the UC 2 , but not on a temperature gradient. The apparent activation energy for uranium loss, ΔH, is approximately 90 kcal/mole. Calculations indicate that uranium loss from the kernel would be insignificant under conditions to be expected in an HTGR

  11. Fission product release from HTGR fuel under core heatup accident conditions - HTR2008-58160

    International Nuclear Information System (INIS)

    Verfondern, K.; Nabielek, H.

    2008-01-01

    Various countries engaged in the development and fabrication of modern fuel for the High Temperature Gas-Cooled Reactor (HTGR) have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under operating and accidental conditions of future HTGRs. Within the IAEA directed Coordinated Research Project CRP6 on 'Advances in HTGR Fuel Technology Development' active since 2002, the 13 participating Member States have agreed upon benchmark studies on fuel performance during normal operation and under accident conditions. While the former has been completed in the meantime, the focus is now on the extension of the national code developments to become applicable to core heatup accident conditions. These activities are supported by the fact that core heatup simulation experiments have been resumed recently providing new, highly valuable data. Work on accident performance will be - similar to the normal operation benchmark - consisting of three essential parts comprising both code verification that establishes the correspondence of code work with the underlying physical, chemical and mathematical laws, and code validation that establishes reasonable agreement with the existing experimental data base, but including also predictive calculations for future heating tests and/or reactor concepts. The paper will describe the cases to be studied and the calculational results obtained with the German computer model FRESCO. Among the benchmark cases in consideration are tests which were most recently conducted in the new heating facility KUEFA. Therefore this study will also re-open the discussion and analysis of both the validity of diffusion models and the transport data of the principal fission product species in the HTGR fuel materials as essential input data for the codes. (authors)

  12. Deactivating a major nuclear fuels reprocessing facility

    International Nuclear Information System (INIS)

    LeBaron, G.J.

    1997-01-01

    This paper describes three key processes used in deactivating the Plutonium Uranium Extraction (PUREX) Facility, a large, complex nuclear reprocessing facility, 15 months ahead of schedule and $77 million under budget. The organization was reengineered to refine its business processes and more effectively organize around the deactivation work scope. Multi-disciplined work teams were formed to be self-sufficient and empowered to make decisions and perform work. A number of benefits were realized by reengineering. A comprehensive process to develop end points which clearly identified specific results and the post-project facility configuration was developed so all areas of a facility were addressed. Clear and specific end points allowed teams to focus on completing deactivation activities and helped ensure there were no unfulfilled end-of-project expectations. The RCRA regulations require closure of permitted facilities within 180 days after cessation of operations which may essentially necessitate decommissioning. A more cost effective approach was adopted which significantly reduced risk to human health and the environment by taking the facility to a passive, safe, inexpensive-to-maintain surveillance and maintenance condition (deactivation) prior to disposition. PUREX thus became the first large reprocessing facility with active TSD [treatment, storage, and disposal] units to be deactivated under the RCRA regulations

  13. Standard model for safety analysis report of fuel reprocessing plants

    International Nuclear Information System (INIS)

    1979-12-01

    A standard model for a safety analysis report of fuel reprocessing plants is established. This model shows the presentation format, the origin, and the details of the minimal information required by CNEN (Comissao Nacional de Energia Nuclear) aiming to evaluate the requests of construction permits and operation licenses made according to the legislation in force. (E.G.) [pt

  14. Reprocessing of spent nuclear fuel; Prerada isluzenog nuklearnog goriva

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorija za visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    This report covers: chemical-technology investigation of modified purex process for reprocessing of spent fuel; implementation of the procedure for obtaining plutonium peroxide and oxalate; research in the field of uranium, plutonium, and fission products separation by inorganic ion exchangers and extraction by organic solutions; study of the fission products in the heavy water RA reactor.

  15. Reprocessing of spent nuclear fuels. Status and trends

    International Nuclear Information System (INIS)

    Hultgren, Aa.

    1993-01-01

    The report gives a short review of the status for industrial reprocessing and recycling of Uranium/Plutonium. The following countries are covered: Belgium, France, Germany, Great Britain, India, Japan, Russia, USA. Different fuel cycle strategies are accounted for, and new developments outlined. 116 refs, 27 figs, 12 tabs

  16. Implications of ICPR 60 for nuclear fuel reprocessing in france

    International Nuclear Information System (INIS)

    Mathieu, P.

    1992-01-01

    The ICRP 60 publication intends to guide the regulatory agencies on the main rules and principle of protection. The text contains recommendations for practices and for emergencies. The following report intends to develop the possible consequences of the publication for the reprocessing of spent fuel as managed by COGEMA in the plants of La Hague and Marcoule. (author)

  17. Reasons for and against reprocessing of spent fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Gries, W

    1983-06-01

    In the following the reasons for and against the main methods of waste disposal are compared. The author examines the advantages and disadvantages of waste disposal by reprocessing of spent fuel assemblies or by immediate ultimate storage. To get a general idea the pros and cons are arranged and analysed according to the following subjects: - technology/science, - safety/environment, - profitability, - political aspects.

  18. Reprocessing of spent nuclear fuel; Prerada isluzenog nuklearnog goriva

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorija za hemiju visoke aktivnosti, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    This volume contains the following reports: Experimental facility for testing and development of pulsed columns and auxiliary devices; Chemical-technology study of the modified 'Purex' process; Chemical and radiometric control analyses; Chromatographic separation of rare earth elements on paper treated by di-n butylphosphate; Preliminary study of some organic nitrogen extracts significant in fuel reprocessing.

  19. General criteria for the project of nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    1979-01-01

    Recommendations are presented establishing the general criteria for the project of nuclear fuel reprocessing plants to be licensed according to the legislation in effect. They apply to all the plant's systems, components and structures which are important to operation safety and to the public's health and safety. (F.E.) [pt

  20. Materials management in an internationally safeguarded fuels reprocessing plant

    International Nuclear Information System (INIS)

    Hakkila, E.A.; Cobb, D.D.; Dayem, H.A.; Dietz, R.J.; Kern, E.A.; Markin, J.T.; Shipley, J.P.; Barnes, J.W.; Scheinman, L.

    1980-04-01

    The first volume of this report summarizes the results and conclusions for this study of conventional and advanced nuclear materials accounting systems applicable for both large (1500 MTHM/y) and small (210 MTHM/y) spent-fuel reprocessing facilities subject to international verification

  1. Spent fuel reprocessing system security engineering capability maturity model

    International Nuclear Information System (INIS)

    Liu Yachun; Zou Shuliang; Yang Xiaohua; Ouyang Zigen; Dai Jianyong

    2011-01-01

    In the field of nuclear safety, traditional work places extra emphasis on risk assessment related to technical skills, production operations, accident consequences through deterministic or probabilistic analysis, and on the basis of which risk management and control are implemented. However, high quality of product does not necessarily mean good safety quality, which implies a predictable degree of uniformity and dependability suited to the specific security needs. In this paper, we make use of the system security engineering - capability maturity model (SSE-CMM) in the field of spent fuel reprocessing, establish a spent fuel reprocessing systems security engineering capability maturity model (SFR-SSE-CMM). The base practices in the model are collected from the materials of the practice of the nuclear safety engineering, which represent the best security implementation activities, reflect the regular and basic work of the implementation of the security engineering in the spent fuel reprocessing plant, the general practices reveal the management, measurement and institutional characteristics of all process activities. The basic principles that should be followed in the course of implementation of safety engineering activities are indicated from 'what' and 'how' aspects. The model provides a standardized framework and evaluation system for the safety engineering of the spent fuel reprocessing system. As a supplement to traditional methods, this new assessment technique with property of repeatability and predictability with respect to cost, procedure and quality control, can make or improve the activities of security engineering to become a serial of mature, measurable and standard activities. (author)

  2. Development of a pneumatic transfer system for HTGR recycle fuel particles

    International Nuclear Information System (INIS)

    Mack, J.E.; Johnson, D.R.

    1978-02-01

    In support of the High-Temperature Gas-Cooled Reactor (HTGR) Fuel Refabrication Development Program, an experimental pneumatic transfer system was constructed to determine the feasibility of pneumatically conveying pyrocarbon-coated fuel particles of Triso and Biso designs. Tests were conducted with these particles in each of their nonpyrophoric forms to determine pressure drops, particle velocities, and gas flow requirements during pneumatic transfer as well as to evaluate particle wear and breakage. Results indicated that the material can be pneumatically conveyed at low pressures without excessive damage to the particles or their coatings

  3. Historical fuel reprocessing and HLW management in Idaho

    International Nuclear Information System (INIS)

    Knecht, D.A.; Staiger, M.D.; Christian, J.D.

    1997-01-01

    This article review some of the key decision points in the historical development of spent fuel reprocessing and waste management practices at the Idaho Chemical Processing Plant that have helped ICPP to successfully accomplish its mission safely and with minimal impact on the environment. Topics include ICPP reprocessing development; batch aluminum-uranium dissolution; continuous aluminum uranium dissolution; batch zirconium dissolution; batch stainless steel dissolution; semicontinuous zirconium dissolution with soluble poison; electrolytic dissolution of stainless steel-clad fuel; graphite-based rover fuel processing; fluorinel fuel processing; ICPP waste management consideration and design decisions; calcination technology development; ICPP calcination demonstration and hot operations; NWCF design, construction, and operation; HLW immobilization technology development. 80 refs., 4 figs

  4. An Experiment on the Carbonization of Fuel Compact Matrix Graphite for HTGR

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, Joo Hyoung; Cho, Moon Sung

    2012-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a properly prepared matrix graphite powder, pressed into a spherical shape or a cylindrical compact, and finally heat-treated at about 1800 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, over coating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K, In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is of extreme importance to investigate the relationship among the process parameters of the matrix graphite powder preparation, fabrication parameters of fuel element green compact and the carbonization condition, which has a strong influence on further steps and the material properties of fuel element. In this work, the carbonization behavior of green compact samples prepared from the matrix graphite powder mixtures with different binder materials was investigated in order to elucidate the behavior of binders during the carbonization heat treatment by analyzing the change in weight, density and its

  5. TAPIR, Thermal Analysis of HTGR with Graphite Sleeve Fuel Elements

    International Nuclear Information System (INIS)

    Weicht, U.; Mueller, W.

    1983-01-01

    1 - Nature of the physical problem solved: Thermal analysis of a reactor core containing internally and/or externally gas cooled prismatic fuel elements of various geometries, rating, power distribution, and material properties. 2 - Method of solution: A fuel element in this programme is regarded as a sector of a fuelled annulus with graphite sleeves of any shape on either side and optional annular gaps between fuel and graphite and/or within the graphite. It may have any centre angle and the fuelled annulus may become a solid cylindrical rod. Heat generation in the fuel is assumed to be uniform over the cross section and peripheral heat flux into adjacent sectors is ignored. Fuel elements and coolant channels are treated separately, then linked together to fit a specified pattern. 3 - Restrictions on the complexity of the problem: Maxima of: 50 fuel elements; 50 cooled channels; 25 fuel geometries; 25 coolant channel geometries; 10 axial power distributions; 10 graphite conductivities

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

    International Nuclear Information System (INIS)

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

    1979-06-01

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

  7. Scoping study of flowpath of simulated fission products during secondary burning of crushed HTGR fuel in a quartz fluidized-bed burner

    International Nuclear Information System (INIS)

    Rindfleisch, J.A.; Barnes, V.H.

    1976-04-01

    The results of four experimental runs in which isotopic tracers were used to simulate fission products during fluidized bed secondary burning of HTGR fuel were studied. The experimental tests provided insight relative to the flow path of fission products during fluidized-bed burning of HTGR fuel

  8. Multiservice utility plug for remote fuel reprocessing

    International Nuclear Information System (INIS)

    Goldmann, L.H. Jr.; Jensen, D.A.

    1979-10-01

    This paper presents the design of a multiservice utility plug and drive system to be used for reliably engaging and disengaging all utility connections automatically that serve large portable equipment modules. The modules are arranged into a fuel processing production line within the Fuels and Materials Examination Laboratory. The utility plugs allow the modules to be easily replaced, rearranged or removed for maintenance

  9. 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

  10. 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)

  11. 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)

  12. Experimental determination of thermal conductivity and gap conductance of fuel rod for HTGR

    International Nuclear Information System (INIS)

    Kikuchi, Teruo; Iwamoto, Kazumi; Ikawa, Katsuichi; Ishimoto, Kiyoshi

    1985-01-01

    The thermal conductivity of fuel compacts and the gap conductance between the fuel compact and the graphite sleeve in fuel rods for a high-temperature gas-cooled reactor (HTGR) were measured by the center heating method. These measurements were made as functions of volume percent particle loading and temperature for thermal conductivity and as functions of gap distance and gas composition for gap conductance. The thermal conductivity of fuel compacts decreases with increasing temperature and with increasing particle loading. The gap conductance increases with increasing temperature and decrease with increasing gap distance. A good gap conductance was observed with helium fill gas. It was seen that the gap conductance was dependent on the thermal conductivity of fill gas and conductance by radiation and could be neglected the conductance through solid-solid contact points of fuel compact and graphite sleeve. (author)

  13. Irradiation performance of HTGR fuel rods in HFIR experiments HRB-7 and -8

    International Nuclear Information System (INIS)

    Valentine, K.H.; Homan, F.J.; Long, E.L. Jr.; Tiegs, T.N.; Montgomery, B.H.; Hamner, R.L.; Beatty, R.L.

    1977-05-01

    The HRB-7 and -8 experiments were designed as a comprehensive test of mixed thorium-uranium oxide fissile particles with Th:U ratios from 0 to 8 for HTGR recycle application. In addition, fissile particles derived from Weak-Acid Resin (WAR) were tested as a potential backup type of fissile particle for HTGR recycle. These experiments were conducted at two temperatures (1250 and 1500 0 C) to determine the influence of operating temperature on the performance parameters studied. The minor objectives were comparison of advanced coating designs where ZrC replaced SiC in the Triso design, testing of fuel coated in laboratory-scale equipment with fuel coated in production-scale coaters, comparison of the performance of 233 U-bearing particles with that of 235 U-bearing particles, comparison of the performance of Biso coatings with Triso coatings for particles containing the same type of kernel, and testing of multijunction tungsten-rhenium thermocouples. All objectives were accomplished. As a result of these experiments the mixed thorium-uranium oxide fissile kernel was replaced by a WAR-derived particle in the reference recycle design. A tentative decision to make this change had been reached before the HRB-7 and -8 capsules were examined, and the results of the examination confirmed the accuracy of the previous decision. Even maximum dilution (Th/U approximately equal to 8) of the mixed thorium-uranium oxide kernel was insufficient to prevent amoeba of the kernels at rates that are unacceptable in a large HTGR. Other results showed the performance of 233 U-bearing particles to be identical to that of 235 U-bearing particles, the performance of fuel coated in production-scale equipment to be at least as good as that of fuel coated in laboratory-scale coaters, the performance of ZrC coatings to be very promising, and Biso coatings to be inferior to Triso coatings relative to fission product retention

  14. Fuel reprocessing at THORP: profitability and public liabilities

    International Nuclear Information System (INIS)

    Berkhout, F.

    1992-01-01

    Since the economics of British Nuclear Fuels Limited's (BNFL) Thermal Oxide Reprocessing Plant (THORP) were analysed in an earlier report, a number of domestic and international developments have affected the prospects for THORP. The present report outlines these changes, and analyses their implications for the profits and public liabilities associated with the project. Timing is of some significance because once THORP becomes radioactive (planned to occur in March 1993) the bill for decommissioning the plant will rise from a trivial sum to a very large one - Pound 900 million (1992 prices) in BNFL's own estimates. The report begins with a brief outline of reprocessing and the THORP project. It then examines the market prospects for reprocessing beyond THORP's first ten years and revises BNFL's own projections. It then considers the potential profitability of THORP in relation to various possible cost increases and finally outlines the possible implications of different THORP scenarios for the public purse. (author)

  15. The refurbishment of the D1206 fuel reprocessing plant

    International Nuclear Information System (INIS)

    Bailey, G.

    1988-01-01

    The term decommissioning can be applied not only to reactors but to any nuclear plant, laboratory, building or part of a building that may have been associated with radioactive material and needs to be restored to clean conditions. In this case the decommissioning and reconstruction of the Dounreay Fast Reactor fuel reprocessing plant, so that plutonium oxide could be reprocessed as well as enriched uranium fuel, is described. The work included improving containment and shielding, building a new head-end treatment cave for the more complex and larger fuel elements, improving the ventilation and constructing a new dissolver. In this paper the breakdown cave and dissolver cell are described and compared and the work done explained. (U.K.)

  16. Remote maintenance in TOR fast reactor fuel reprocessing facility

    International Nuclear Information System (INIS)

    Eymery, R.; Constant, M.; Malterre, G.

    1986-11-01

    The TOR facility which is undergoing commissioning tests has a capacity of 5 T. HM/year which is enough for reprocessing all the Phenix fuel, with an excess capacity which is to be used for other fast reactors fuels. It is the result of enlargement and renovation of the old Marcoule pilot facility. A good load factor is expected through the use of equipment with increased reliability and easy maintenance. TOR will also be used to test new equipment developed for the large breeder fuel reprocessing plant presently in the design stage. The latter objective is specifically important for the parts of the plant involving mechanical equipment which are located in a new building: TOR 1. High reliability and flexibility will be obtained in this building thanks to the attention given to the integrated remote handling system [fr

  17. Reprocessing technology of liquid metal cooled fast breeder reactor fuel

    International Nuclear Information System (INIS)

    Baetsle, L.H.; Broothaerts, J.; Heylen, P.R.; Eschrich, H.; Geel, J. van

    1974-11-01

    All the important aspects of LMFBR fuel reprocessing are critically reviewed in this report. Storage and transportation techniques using sodium, inert gas, lead, molten salts and organic coolants are comparatively discussed in connection with cooling time and de-activation techniques. Decladding and fuel disaggregation of UO 2 -PuO 2 fuel are reviewed according to the present state of R and D in the main nuclear powers. Strong emphasis is put on on voloxidation, mechanical pulverization and molten salt disaggregation in connection with volatilization of gaseous fission products. Release of fission gases and the resulting off-gas treatment are discussed in connection with cooling time, burn up and dissagregation techniques. The review is limited to tritium, iodine xenon-krypton and radioactive airborne particulates. Dissolution, solvent extraction and plutonium purification problems specifically connected to LMFBR fuel are reviewed with emphasis on the differences between LWR and fast fuel reprocessing. Finally the categories of wastes produced by reprocessing are analysed according to their origin in the plant and their alpha emitters content. The suitable waste treatment techniques are discussed in connection with the nature of the wastes and the ultimate disposal technique. (author)

  18. Methodology for estimating reprocessing costs for nuclear fuels

    International Nuclear Information System (INIS)

    Carter, W.L.; Rainey, R.H.

    1980-02-01

    A technological and economic evaluation of reprocessing requirements for alternate fuel cycles requires a common assessment method and a common basis to which various cycles can be related. A methodology is described for the assessment of alternate fuel cycles utilizing a side-by-side comparison of functional flow diagrams of major areas of the reprocessing plant with corresponding diagrams of the well-developed Purex process as installed in the Barnwell Nuclear Fuel Plant (BNFP). The BNFP treats 1500 metric tons of uranium per year (MTU/yr). Complexity and capacity factors are determined for adjusting the estimated facility and equipment costs of BNFP to determine the corresponding costs for the alternate fuel cycle. Costs of capacities other than the reference 1500 MT of heavy metal per year are estimated by the use of scaling factors. Unit costs of reprocessed fuel are calculated using a discounted cash flow analysis for three economic bases to show the effect of low-risk, typical, and high-risk financing methods

  19. An overview on dry reprocessing of irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Ouyang Yinggen

    2002-01-01

    Although spent nuclear fuels have been reprocessed successfully for many years by the well-know Purex process based on solvent extraction, other reprocessing method which do not depend upon the use of organic solvents and aqueous media appear to have important potential advantage. There are two main non-aqueous methods for the reprocessing of spent fuel: fluoride-volatility process and pyro-electrochemical process. The presence of a poser in the process is that PuF 6 is obviously thermodynamically stable only in the presence of a large excess of fluorine. Pyro-electrochemical process is suited to processing metallic, oxide and carbide fuels. First, the fuel is dissolved in fresh salts, then, electrodes are introduced into the bath, U and Pu are deposited on the cathode, third, separation and refinement U and Pu are deposited on the cathode. There is a couple of contradictions in the process that are not in harmonious proportion in the fields on the nuclear fuel is dissolved the ability in the molten salt and corrosiveness of the molten salt for equipment used in the process

  20. Safety aspects of a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Donoghue, J.K.; Charlesworth, F.R.; Fairbairn, A.

    1977-01-01

    decommissioning and demolition procedures must be anticipated and suitable provision made. Application of these principles is illustrated by experience gained in the surveillance of reprocessing plants. United Kingdom regulatory procedures for the licensing and inspection of reprocessing plant, and statutory requirements influencing safety in design, construction and operation are reviewed. Recent developments in safety legislation including the Health and Safety at Work Act, 1974, are discussed

  1. An investigation on technical feasibilities of fuel cycle for high temperature gas-cooled reactor (Case study)

    International Nuclear Information System (INIS)

    Sumita, Junya; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sawa, Kazuhiro

    2008-03-01

    In accordance with the basic policy of effectively using nuclear fuel resources, the FBR cycle, one of the most possible fuel cycle in the future, will be adapted after plu-thermal program by LWR in Japanese nuclear cycle plan. In this paper, a case study of technical investigation of HTGR fuel cycle based on HTGR fuel cycle proposed to adapt to Japanese nuclear fuel cycle plan were carried out from the viewpoint of effective utilization of uranium, fabrication technologies of MOX fuel, reprocessing technologies, amount of interim storage of HTGR fuel and graphite waste. As a result, the fuel cycle for HTGR is expected to be possible technically. (author)

  2. 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)

  3. Design and main characteristics of HTGR fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Permyakov, L.N.; Koshelev, Yu.V.; Mikhajlichenko, L.I.

    1983-01-01

    Two types of spherical fuel elements and coated particles were investigated under the operating conditions of the high temperature reactors in the Soviet Union (VGR-50 and VG-400). This paper gives the main characteristics of spherical fuel elements (thermal conductivity, static and dynamic strength, wear resistance, release of gaseous fission products, etc.) as determined in test facilities. (author)

  4. Development of a fissile particle for HTGR fuel recycle

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.; Lindemer, T.B.; Beatty, R.L.; Tiegs, T.N.

    1976-12-01

    Recycle fissile fuel particles for high-temperature gas-cooled reactors (HTGRs) have been under development since the mid-1960s. Irradiation performance on early UO 2 and Th 0 . 8 U 0 . 2 O 2 kernels is described in this report, and the performance limitations associated with the dense oxide kernels are presented. The development of the new reference fuel kernel, the weak-acid-resin-derived (WAR) UO 2 --UC 2 , is discussed in detail, including an extensive section on the irradiation performance of this fuel in HFIR removable beryllium capsules HRB-7 through -10. The conclusion is reached that the irradiation performance of the WAR fissile fuel kernel is better than that of any coated particle fuel yet tested. Further, the present fissile kernel is adequate for steam cycle HTGRs as well as for many advanced applications such as gas turbine and process heat HTGRs

  5. Measurement of Weight of Kernels in a Simulated Cylindrical Fuel Compact for HTGR

    International Nuclear Information System (INIS)

    Kim, Woong Ki; Lee, Young Woo; Kim, Young Min; Kim, Yeon Ku; Eom, Sung Ho; Jeong, Kyung Chai; Cho, Moon Sung; Cho, Hyo Jin; Kim, Joo Hee

    2011-01-01

    The TRISO-coated fuel particle for the high temperature gas-cooled reactor (HTGR) is composed of a nuclear fuel kernel and outer coating layers. The coated particles are mixed with graphite matrix to make HTGR fuel element. The weight of fuel kernels in an element is generally measured by the chemical analysis or a gamma-ray spectrometer. Although it is accurate to measure the weight of kernels by the chemical analysis, the samples used in the analysis cannot be put again in the fabrication process. Furthermore, radioactive wastes are generated during the inspection procedure. The gamma-ray spectrometer requires an elaborate reference sample to reduce measurement errors induced from the different geometric shape of test sample from that of reference sample. X-ray computed tomography (CT) is an alternative to measure the weight of kernels in a compact nondestructively. In this study, X-ray CT is applied to measure the weight of kernels in a cylindrical compact containing simulated TRISO-coated particles with ZrO 2 kernels. The volume of kernels as well as the number of kernels in the simulated compact is measured from the 3-D density information. The weight of kernels was calculated from the volume of kernels or the number of kernels. Also, the weight of kernels was measured by extracting the kernels from a compact to review the result of the X-ray CT application

  6. The regulations concerning the reprocessing business of spent fuels

    International Nuclear Information System (INIS)

    1978-01-01

    In compliance with ''The law for the regulations of nuclear source material, nuclear fuel material and reactors'' these regulations prescribe concerning reprocessing facilities: The procedures to apply for the approval of the design and method of construction and the approval of the change thereof; as well as the procedure to apply for the inspection of the facilities, and details of the inspection (in sections 2-6). After that, the regulations require the enterpriser of reprocessing business to keep necessary records and take necessary measures for safety concerning the facilities, operation of reprocessing equipments, and transportation, storage on disposal of used fuel, materials separated therefrom or materials contaminated by either of them (in sections 8-16). Further, the regulations prescribe the procedure to apply for the approval of the safety rule required to the enterpriser of reprocessing business by above mentioned law and specifies items which should be included into the rule (section 17). Moreover, the regulations require the enterpriser to submit reports of each use of the internationally controllled material and specifies the items which should be included into these reports (section 19). (Matsushima, A.)

  7. Air conditioning facilities in a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Kawasaki, Michitaka; Oka, Tsutomu

    1987-01-01

    Reprocessing plants are the facilities for separating the plutonium produced by nuclear reaction and unconsumed remaining uranium from fission products in the spent fuel taken out of nuclear reactors and recovering them. The fuel reprocessing procedure is outlined. In order to ensure safety in handling radioactive substances, triple confinement using vessels, concrete cells and buildings is carried out in addition to the prevention of criticality and radiation shielding, and stainless steel linings and drip trays are installed as occasion demands. The ventilation system in a reprocessing plant is roughly divided into three systems, that is, tower and tank ventilation system to deal with offgas, cell ventilation system for the cells in which main towers and tanks are installed, and building ventilation system. Air pressure becomes higher from tower and tank system to building system. In a reprocessing plant, the areas in a building are classified according to dose rate. The building ventilation system deals with green and amber areas, and the cell ventilation system deals with red area. These three ventilation systems are explained. Radiation monitors are installed to monitor the radiation dose rate and air contamination in working places. The maintenance and checkup of ventilation systems are important. (Kako, I.)

  8. Fission product release from HTGR coated microparticles and fuel elements

    International Nuclear Information System (INIS)

    Gusev, A.A.; Deryugin, A.I.; Lyutikov, R.A.; Chernikov, A.S.

    1991-01-01

    The article presents the results of the investigation of fission products release from microparticles with UO 2 core and five-layer HII PyC- and SiC base protection layers of TRICO type as well as from spherical fuel elements based thereon. It is shown that relative release of short-lived xenon and crypton from microparticles does not exceed (2-3) 10 -7 . The release of gaseous fission products from fuel elements containing no damaged coated microparticles, is primarily determined by the contamination of matrix graphite with fuel. An analytical dependence is derived, the dependence described the relation between structural parameters of coated microparticles, irradiation conditions and fuel burnup at which depressurization of coated microparticles starts

  9. Fuel reprocessing: safety analysis of extraction cycles

    International Nuclear Information System (INIS)

    Dinh, B.; Mauborgne, B.; Baron, P.; Mercier, J.P.

    1991-01-01

    An essential part of the safety analysis related to the extraction cycles of reprocessing plants, is the analysis of their behaviour during steady-state and transient operations, by means of simulation codes. These codes are based on the chemical properties of the main species involved (distribution coefficient and kinetics) and the hydrodynamics inside the contactors (mixer-settlers and pulsed columns). These codes have been consolidated by comparison of calculations with experimental results. The safety analysis is essentially performed in two steps. The first step is a parametric sensitivity analysis of the chemical flowsheet operated: the effect of a misadjustment (flowrate of feed, solvent, etc) is evaluated by successive steady-state calculations. These calculations help the identification of the sensitive parameters for the risk of plutonium accumulation, while indicating the permissible level of misadjustment. These calculations also serve to identify the parameters which should be measured during plant operation. The second step is the study of transient regimes, for the most sensitive parameters related to plutonium accumulation risk. The aim is to confirm the conclusions of the first step and to check that the characteristic process parameters chosen effectively allow, the early and reliable detection of any drift towards a plutonium accumulating regime. The procedures to drive the process backwards to a specified convenient steady-state regime from a drifting-state are also verified. The identification of the sensitive parameters, the process status parameters and the process transient analysis, allow a good control of process operation. This procedure, applied to the first purification cycle of COGEMA's UP3-A La Hague plant has demonstrated the total safety of facility operations

  10. Analysis and study of spent fuel reprocessing technology from birth to present

    International Nuclear Information System (INIS)

    Takahashi, Keizo

    2006-01-01

    As for the nuclear fuel reprocessing of the spent fuel, although there was argument of pros and cons, it was decided to start Rokkasho reprocessing project further at the Japan Atomic Energy Commission of ''Long-Term Program for Research, Development and Utilization of Nuclear Energy'' in year 2004. The operation of Tokai Reprocessing is going steadily to reprocess spent fuel more than 1,100 tons. In this paper, history, present status and future of reprocessing technology is discussed focusing from military Pu production, Magnox fuel reprocessing to oxide fuel reprocessing. Amount of reprocessed fuel are estimated based on fuel type. Then, history of reprocessing, US, UK, France, Germany, Russian, Belgian and Japan is presented and compared on technology, national character, development organization, environmental protection, and high active waste vitrification. Technical requirements are increased from Pu production fuel, Magnox fuel and oxide fuel mainly because of higher burnup. Reprocessing technology is synthetic of engineering and accumulation of operational experience. The lessons learned from the operational experience of the world will be helpful for establishment of nuclear fuel reprocessing technology in Japan. (author)

  11. HTGR fuel development: investigations of breakages of uranium-loaded weak acid resin microspheres

    International Nuclear Information System (INIS)

    Carpenter, J.A. Jr.

    1977-11-01

    During the HTGR fuel development program, a high percentage of uranium-loaded weak acid resin microspheres broke during pneumatic transfer, carbonization, and conversion. One batch had been loaded by the UO 3 method; the other by the ammonia neutralization method. To determine the causes of failure, samples of the two failed batches were investigated by optical microscopy, scanning electron microscopy, electron beam microprobe, and other techniques. Causes of failure are postulated and methods are suggested to prevent recurrence of this kind of failure

  12. Fission product behavior in HTGR fuel particles made from weak-acid resins

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Henson, T.J.

    1979-04-01

    Fission product retention and behavior are of utmost importance in HTGR fuel particles. The present study concentrates on particles made from weak-acid resins, which can vary in composition from 100% UO 2 plus excess carbon to 100% UC 2 plus excess carbon. Five compositions were tested: UC 4 58 O 2 04 , UC 3 68 O 0 01 , UC 4 39 O 1 72 , UC 4 63 O 0 97 , and UC 4 14 O 1 53 . Metallographically sectioned particles were examined with a shielded electron microprobe. The distributions of the fission products were determined by monitoring characteristic x-ray lines while scanning the electron beam over the particle surface

  13. 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

  14. Remotex and servomanipulator needs in nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Garin, J.

    1981-01-01

    Work on the conceptual design of a pilot-scale plant for reprocessing breeder reactor fuels is being performed at Oak Ridge National Laboratory. The plant design will meet all current federal regulations for repocessing plants and will serve as prototype for future production plants. A unique future of the concept is the incorporation of totally remote operation and maintenance of the process equipment within a large barn-like hot cell. This approach, caled Remotex, utilizes servomanipulators coupled with television viewing to extend man's capabilities into the hostile cell environment. The Remotex concept provides significant improvements for fuel reprocessing plants and other nuclear facilities in the areas of safeguarding nuclear materials, reducing radiation exposure, improving plant availability, recovering from unplanned events, and plant decommissioning

  15. Criticality considerations for 233U fuels in an HTGR fuel refabrication facility

    International Nuclear Information System (INIS)

    McNeany, S.R.; Jenkins, J.D.

    1978-01-01

    Eleven 233 U solution critical assemblies spanning an H/ 233 U ratio range of 40 to 2000 and a bare metal 233 U assembly have been calculated with the ENDF/B-IV and Hansen-Roach cross sections. The results from these calculations are compared with the experimental results and with each other. An increasing disagreement between calculations with ENDF/B and Hansen-Roach data with decreasing H/ 233 U ratio was observed, indicative of large differences in their intermediate energy cross sections. The Hansen-Roach cross sections appeared to give reasonably good agreement with experiments over the whole range; whereas the ENDF/B calculations yielded high values for k/sub eff/ on assemblies of low moderation. It is concluded that serious problems exist in the ENDF/B-IV representation of the 233 U cross sections in the intermediate energy range and that further evaluation of this nuclide is warranted. In addition, it is recommended that an experimental program be undertaken to obtain 233 U criticality data at low H/ 233 U ratios for verification of generalized criticality safety guidelines. Part II of this report presents the results of criticality calculations on specific pieces of equipment required for HTGR fuel refabrication. In particular, fuel particle storage hoppers and resin carbonization furnaces are criticality safe up to 22.9 cm (9.0 in.) in diameter providing water or other hydrogenous moderators are excluded. In addition, no criticality problems arise due to accumulation of particles in the off-gas scrubber reservoirs provided reasonable administrative controls are exercised

  16. Abnormal reactions in a evaporator in a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Kida, Takashi; Umeda, Miki; Sugikawa, Susumu

    2003-01-01

    In order to evaluate a self-accelerated reaction in an evaporator in a fuel reprocessing plant due to organic-nitric acid reactions, a development of a calculation code is under way. Mock-up tests were performed to investigate the fluid dynamic behavior of the organic solvent in the evaporator. Based on these results, the model of the calculation code was constructed. This report describes the results of mock-up tests and the model of the calculation code. (author)

  17. Retention of gaseous fission products in reprocessing LMFBR fuels

    International Nuclear Information System (INIS)

    Burch, W.D.; Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-05-01

    The report is devoted to status of the development programme at the Oak Ridge National Laboratory on methods for retaining iodine-131 and 129, Krypton-85, Tritium and Carbon-14 in reprocessing LMFBR fuels. The Iodox process, Fluorocarbon absorption process and Voloxidation process are described for retention of iodine, Krypton-85 and Tritium, respectively. Flowsheets for the different processes are given and results of experimental runs in small engineering-scale equipment are reported

  18. Assembly of laboratory line for nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Fidler, J.; Chotivka, V.

    1979-01-01

    The dismantling of a laboratory line for spent fuel reprocessing after the termination of the research programme and the procedures for hot and semi-hot cell decontamination are described. The equipment was mostly disassembled in smaller parts which were then decontaminated by wiping them with cotton wool soaked in detergent and citric acid, varnished with two-component epoxi varnish, wrapped into multiple polyethylene foils, sealed in PVC bags and thus ready for transport. (B.S.)

  19. The reasons for and against reprocessing of spent fuel elements

    International Nuclear Information System (INIS)

    Gries, W.

    1983-01-01

    In the following the reasons for and against the main methods of waste disposal are compred. The author examines the advantages and disadvantages of waste disposal by reprocessing of spent fuel assemblies or by immediate ultimate storage. To get a general idea the pros and cons are arranged and analysed according to the following subjects: - technology/science, - safety/environment, - profitability, - political aspects. (orig./UA) [de

  20. Loading ion exchange resins with uranium for HTGR fuel kernels

    International Nuclear Information System (INIS)

    Notz, K.J.; Greene, C.W.

    1976-12-01

    Uranium-loaded ion exchange beads provide an excellent starting material in the production of uranium carbide microspheres for nuclear fuel applications. Both strong-acid (sulfonate) and weak-acid (carboxylate) resins can be fully loaded with uranium from a uranyl nitrate solution utilizing either a batch method or a continuous column technique

  1. Fuel reprocessing data validation using the isotope correlation technique

    International Nuclear Information System (INIS)

    Persiani, P.J.; Bucher, R.G.; Pond, R.B.; Cornella, R.J.

    1990-01-01

    The Isotope Correlation Technique (ICT), in conjunction with the gravimetric (Pu/U ratio) method for mass determination, provides an independent verification of the input accountancy at the dissolver or accountancy stage of the reprocessing plant. The Isotope Correlation Technique has been applied to many classes of domestic and international reactor systems (light-water, heavy-water, and graphite reactors) operating in a variety of modes (power, research, and production reactors), and for a variety of reprocessing fuel cycle management strategies. Analysis of reprocessing operations data based on isotopic correlations derived for assemblies in a PWR environment and fuel management scheme, yielded differences between the measurement-derived and ICT-derived plutonium mass determinations of (- 0.02 ± 0.23)% for the measured U-235 and (+ 0.50 ± 0.31)% for the measured Pu-239, for a core campaign. The ICT analyses has been implemented for the plutonium isotopics in a depleted uranium assembly in a heavy-water, enriched uranium system and for the uranium isotopes in the fuel assemblies in light-water, highly-enriched systems

  2. A comprehensive fuel nuclide analysis at the reprocessing plant

    International Nuclear Information System (INIS)

    Arenz, H.J.; Koch, L.

    1983-01-01

    The composition of spent fuel can be determined by various methods. They rely partially on different information. Therefore the synopsis of the results of all methods permits a detection of systematic errors and their explanation. Methods for determining the masses of fuel nuclides at the reprocessing input point range from pure calculations (shipper data) to mere experimental determinations (volumetric analysis). In between, a mix of ''fresh'' experimental results and ''historical'' data is used to establish a material balance. Deviations in the results obtained by the individual methods can be attributed to the information source, which is unique for the method in question. The methodology of the approach consists of three steps: by paired comparison of the operator analysis (usually volumetric or gravimetric) with remeasurements the error components are determined on a batch-by-batch basis. Using the isotope correlation technique the operator data as well as the remeasurements are checked on an inter-batch basis for outliers, precision and bias. Systematic errors can be uncovered by inter-lab comparison of remeasurements and confirmed by using historical information. Experience collected during the reprocessing of LWR fuel at two reprocessing plants prove the flexibility and effectiveness of this approach. An example is presented to demonstrate its capability in detecting outliers and determining systematic errors. (author)

  3. Engineering study: Fast Flux Test Facility fuel reprocessing

    International Nuclear Information System (INIS)

    Beary, M.M.; Raab, G.J.; Reynolds, W.R. Jr.; Yoder, R.A.

    1974-01-01

    Several alternatives were studied for reprocessing FFTF fuels at Hanford. Alternative I would be to decontaminate and trim the fuel at T Plant and electrolytically dissolve the fuel at Purex. Alternative II would be to decontaminate and shear leach the fuels in a new facility near Purex. Alternative III would be to decontaminate and store fuel elements indefinitely at T Plant for subsequent offsite shipment. Alternative I, 8 to 10 M$ and 13 quarter-years; for Alternative II, 24 to 28 M$ and 20 quarter-years; for Alternative III, 3 to 4 M$ and 8 quarter-years. Unless there is considerable slippage in the FFTF shipping schedule, it would not be possible to build a new facility as described in Alternative II in time without building temporary storage facilities at T Plant, as described in Alternative III

  4. Ventilating system for reprocessing of nuclear fuel rods

    International Nuclear Information System (INIS)

    Szulinski, M.J.

    1981-01-01

    In a nuclear facility such as a reprocessing plant for nuclear fuel rods, the central air cleaner discharging ventilating gas to the atmosphere must meet preselected standards not only as to the momentary concentration of radioactive components, but also as to total quantity per year. In order to comply more satisfactorily with such standards, reprocessing steps are conducted by remote control in a plurality of separate compartments. The air flow for each compartment is regulated so that the air inventory for each compartment has a slow turnover rate of more than a day but less than a year, which slow rate is conveniently designated as quasihermetic sealing. The air inventory in each such compartment is recirculated through a specialized processing unit adapted to cool and/or filter and/or otherwise process the gas. Stale air is withdrawn from such recirculating inventory and fresh air is injected (eg., By the less than perfect sealing of a compartment) into such recirculating inventory so that the air turnover rate is more than a day but less than a year. The amount of air directed through the manifold and duct system from the reprocessing units to the central air cleaner is less than in reprocessing plants of conventional design

  5. Performance of HTGR fuel in HFIR capsule HT-33

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Robbins, J.M.

    1979-06-01

    Irradiation capsule HT-33 was a cooperative effort between General Atomic Company (GA) and Oak Ridge National Laboratory (ORNL). In this capsule ThO 2 particles (fabricated by GA), low-enriched uranium particles, inert carbon particles, and various fuel rod matrices were tested under accelerated irradiation in the High-Flux Isotope Reactor. Visual examination showed good irradiation behavior for fuel rods with slug-injected matrices (using a pitch binder) and warm-molded matrices (using a thermosetting resin binder). Rod debonding improved somewhat with fuel rods that used GLCC H-451 ground graphite shim particles rather than Speer fluid coke shim particles. Measurements of permeability (by inert gas intrusion) of the pyrocarbon on the inert particles showed that the disorder created by the neutron flux did not increase the inert gas permeability. Metallographic examination of Triso-coated particles irradiated both with and without an outer pyrocarbon coating revealed that the outer coating is necessary to suppress SiC degradation at temperatures above approximately 1375 0 C. The fission product behavior (determined by the electron microprobe) was similar in both low-enriched and high-enriched uranium particles made from weak-acid resins. Furthermore, fission product palladium caused severe SiC corrosion at time-averaged temperatures above 1400 0 C

  6. Airborne effluent control for LMFBR fuel reprocessing plants

    International Nuclear Information System (INIS)

    Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-01-01

    A significant part of the LMFBR fuel reprocessing development program has been devoted to the development of efficient removal systems for the volatile fission products, including 131 I, krypton, tritium, 129 I, and most recently 14 C. Flowsheet studies have indicated that very significant reductions of radioactive effluents can be achieved by integrating advanced effluent control systems with new concepts of containment and ventilation; however, the feasibility of such has not yet been established, nor have the economics been examined. This paper presents a flowsheet for the application of advanced containment systems to the processing of LMFBR fuels and summarizes the status and applicability of specific fission product removal systems

  7. Thermal-stress analysis of HTGR fuel and control rod fuel blocks in in-block carbonization and annealing furnace

    International Nuclear Information System (INIS)

    Gwaltney, R.C.; McAfee, W.J.

    1977-01-01

    The equivalent solid plate method, in conjunction with two-dimensional plane stress and plane strain analyses, was used in assessing the thermal stress behavior of HTGR fuel and control rod fuel blocks. For the control rod fuel blocks, particular attention was given to ascertaining the effects of the reserve shutdown hole and the control rod channel holes. The assumed safety factor of 2 on the failure criteria was considered adequate to account for neglecting the axial temperature gradient in the plane analyses of the ends of the blocks. The analyses indicated that the maximum calculated tensile stress values were smaller than the criteria values except for the plane strain analysis of the control rod fuel block end surfaces and the axisymmetric analysis of the fuel block as a circular cylinder. However, most of the maximum calculated strain values were greater than the criteria values

  8. Irradiation performance of HTGR fuel in HFIR capsule HT-31

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Robbins, J.M.; Hamner, R.L.; Montgomery, B.H.; Kania, M.J.; Lindemer, T.B.; Morgan, C.S.

    1979-05-01

    The capsule was irradiated in the High Flux Isotope Reactor at ORNL to peak particle temperatures up to 1600 0 C, fast neutron fluences (0.18 MeV) up to 9 x 10 25 n/m 2 , and burnups up to 8.9% FIMA for ThO 2 particles. The oxygen release from plutonium fissions was less than calculated, possibly because of the solid solution of SrO and rare earth oxides in UO 2 . Tentative results show that pyrocarbon permeability decreases with increasing fast neutron fluence. Fission products in sol-gel UO 2 particles containing natural uranium mostly behaved similarly to those in particles containing highly enriched uranium (HEU). Thus, much of the data base collected on HEU fuel can be applied to low-enriched fuel. Fission product palladium penetrated into the SiC on Triso-coated particles. Also the SiC coating provided some retention of /sup 110m/Ag. Irradiation above about 1200 0 C without an outer pyrocarbon coating degraded the SiC coating on Triso-coated particles

  9. Safety recycling of reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Weinlaender, W.

    Additionally to the measures, which descent from the conventional safety techniques, a series of supplementary protective measures have to be taken in connection with the Atomic Energy Law, the Radiation Protection Ordinance and the nuclear-technical practice, which in particular guarantee a safe enclosure and a safe residual heat rejection of the handled radioactive material and an avoidance of nuclear chain reactions. The most important plant malfunctions to be considered within the scope of the plant safety control according to the atomic law are, the radioactivity release due to mechanical damage of fuel elements, containment leakage, explosions in process equipment and/or vessels, burning of run out organic solvents, criticality malfunctions, and the already mentioned accidental failure of after-heat removal. If we let alone the extremely low probabilities for the occurrence of such accidents due to the selected methods, the layout of the equipment and by taking the required quality warranty measures into consideration, and infer such accidents in spite of this, the resulting radiation doses outside the plant are in all cases much lower than 5 rem, which is the design limit according to the regulations for radiation protection. (orig./HP) [de

  10. Measurement and behaviour of technetium in fast reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Ferguson, C.; Kyffin, T.W.

    1986-02-01

    A method is described for the spectrophotometric measurement of technetium in plant solutions from the reprocessing of fast reactor fuel. The technetium is selectively extracted using tri-iso-octylamine. After back extraction, thiocyanate is added, in the presence of tetrabutyl-ammonium hydroxide, to form the red hexa-thiocyanato anionic complex in a chloroform medium. The concentration of the technetium is then calculated from the spectrophotometric measurement of this complex. This method was applied to bulk samples, collected during a PFR fuel reprocessing campaign, to identify the main routes followed by technetium through the reprocessing plant. In order to understand the probable behaviour of technetium in the process plant streams, an investigation into the influence of plutonium IV nitrate on the extraction of Tc (VII) into 20%v/v tributyl phosphate/odourless kerosene solution from nitric acid solutions, was initiated. The results of this investigation, along with the known distribution coefficient for the extraction of the uranyl/technetium complex U0 2 (N0 3 )(Tc0 4 ).2TBP and the redox chemistry of technetium, are used to predict the probable behaviour of technetium in the process plant streams. This predicted behaviour is compared with the experimental results and reasonable agreement is obtained between experiment and theory, considering the history of the samples analysed. (author)

  11. Radioactive Iodine and Krypton Control for Nuclear Fuel Reprocessing Facilities

    Directory of Open Access Journals (Sweden)

    Nick R. Soelberg

    2013-01-01

    Full Text Available The removal of volatile radionuclides generated during used nuclear fuel reprocessing in the US is almost certain to be necessary for the licensing of a reprocessing facility in the US. Various control technologies have been developed, tested, or used over the past 50 years for control of volatile radionuclide emissions from used fuel reprocessing plants. The US DOE has sponsored, since 2009, an Off-gas Sigma Team to perform research and development focused on the most pressing volatile radionuclide control and immobilization problems. In this paper, we focus on the control requirements and methodologies for 85Kr and 129I. Numerous candidate technologies have been studied and developed at laboratory and pilot-plant scales in an effort to meet the need for high iodine control efficiency and to advance alternatives to cryogenic separations for krypton control. Several of these show promising results. Iodine decontamination factors as high as 105, iodine loading capacities, and other adsorption parameters including adsorption rates have been demonstrated under some conditions for both silver zeolite (AgZ and Ag-functionalized aerogel. Sorbents, including an engineered form of AgZ and selected metal organic framework materials (MOFs, have been successfully demonstrated to capture Kr and Xe without the need for separations at cryogenic temperatures.

  12. Reprocessing of AHWR spent-fuel: Challenges and strategies

    International Nuclear Information System (INIS)

    Kant, S.

    2005-01-01

    Reprocessing of advanced heavy water reactor (AHWR) spent-fuel involves separation of Th, 233 U and Pu, from the fission products and from one another. A proper combination of Purex and Thorex processes is required. The technology development for a reprocessing facility is extremely complex owing to high fissile content, high levels of irradiation, presence high of levels of 232 U, difficulty in thoria dissolution, presence of thorium as the major constituent, problems due to third phase formation with Th, etc. It demands for development of suitable dissolution, solvent extraction, criticality control, U-Pu partitioning, and other equipments and/or techniques. Process modelling, simulation and optimisation are crucial in predicting behaviour of equipments/cycles, and in arriving at safe and optimum flowsheet. A significant success in this field has been achieved. This paper describes the reprocessing aspects pertaining to AHWR spent-fuel, indicating the major technological challenges, strategies to be followed and development requirements. A schematic flowsheet is proposed for Th- 233 U-Pu separation. (author)

  13. Methods and data for HTGR fuel performance and radionuclide release modeling during normal operation and accidents for safety analysis

    International Nuclear Information System (INIS)

    Verfondern, K.; Martin, R.C.; Moormann, R.

    1993-01-01

    The previous status report released in 1987 on reference data and calculation models for fission product transport in High-Temperature, Gas-Cooled Reactor (HTGR) safety analyses has been updated to reflect the current state of knowledge in the German HTGR program. The content of the status report has been expanded to include information from other national programs in HTGRs to provide comparative information on methods of analysis and the underlying database for fuel performance and fission product transport. The release and transport of fission products during normal operating conditions and during the accident scenarios of core heatup, water and air ingress, and depressurization are discussed. (orig.) [de

  14. Present status of fuel reprocessing plant in PNC

    International Nuclear Information System (INIS)

    Koyama, Kenji

    1981-01-01

    In the fuel reprocessing plant of the Power Reactor and Nuclear Fuel Development Corporation, its hot test has now been completed. For starting its full-scale operation duly, the data are being collected on the operation performance and safety. The construction was started in June, 1971, and completed in October, 1974. In July, 1977, spent fuel was accepted in the plant, and the hot test was started. In September, the same year, the first fuel shearing was made. So far, a total of about 31 t U from both BWR and PWR plants has been processed, thus the hot test was entirely completed. The following matters are described: hot test and its results, research on Pu and U mixed extraction, utilization of product plutonium, development of safeguard technology, and repair work on the acid recovery evaporation tank. (J.P.N.)

  15. The measurement of neptunium in fast reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Mair, M.A.; Savage, D.J.; Kyffin, T.W.

    1986-02-01

    Analytical techniques have been developed to measure neptunium in the feed, waste and product streams of a fast reactor fuel reprocessing plant. The estimated level of one microgram per milligram of plutonium in some solutions presented severe separation and measurement problems. An initial separation stage was essential, and both ion exchange and solvent extraction using thenoyltrifluoroacetone were studied. The redox chemistry of neptunium necessary to achieve good separation is considered. Spectrophotometry measurement of the stable neptunium/arsenazo III complex was selected for the final neptunium determination with additional analysis by radiometric methods. Incomplete recovery of neptunium during the separation stages necessitated yield measurements, using either neptunium-237 as an internal standard or the short lived gamma active neptunium-239 isotope as a tracer. The distribution of neptunium between the waste and product streams is discussed, in relation to the chemistry of neptunium in the reprocessing plant. (author)

  16. Chemical engineering in fuel reprocessing. The French experience

    International Nuclear Information System (INIS)

    Viala, M.; Sombret, C.; Bernard, C.; Miquel, P.; Moulin, J.P.

    1992-01-01

    Reprocessing is the back-end of the nuclear fuel cycle, designed to recover valuable fissile materials, especially plutonium, and to condition safely all the wastes ready for disposal. For its new commercial reprocessing plants (UP 3 and UP 2 800) COGEMA decided to include many engineering innovations as well as new processes and key-components developed by CEA. UP 3 is a complete new plant with a capacity of 800 t/y which was put in operation in August 1990. UP 2 800 is an extension of the existing UP 2 facility, designed to achieve the same annual capacity of 800 t/y, to be put in operation at the end of 1993 by the commissioning of a new head-end and highly active chemical process facilities

  17. Evaluation of temperature coefficients of reactivity for 233U--thorium fueled HTGR lattices. Final report

    International Nuclear Information System (INIS)

    Newman, D.F.; Leonard, B.R. Jr.; Trapp, T.J.; Gore, B.F.; Kottwitz, D.A.; Thompson, J.K.; Purcell, W.L.; Stewart, K.B.

    1977-05-01

    A comparison of calculated and measured neutron multiplication factors as a function of temperature was made for three graphite-moderated lattices in the High Temperature Lattice Test Reactor (HTLTR) using 233 UO 2 --ThO 2 fuels in varying amounts and configurations. Correlation of neutronic analysis methods and cross section data with the experimental measurements forms the basis for assessing the accuracy of the methods and data and developing confidence in the ability to predict the temperature coefficient of reactivity for various High Temperature Gas-Cooled Reactor (HTGR) conditions in which 233 U and thorium are present in the fuel. The calculated values of k/sub infinity/(T) were correlated with measured values using two least-squares-fitted correlation coefficients: (1) a normalization factor, and (2) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross section data

  18. Automatic particle-size analysis of HTGR nuclear fuel microspheres

    International Nuclear Information System (INIS)

    Mack, J.E.

    1977-01-01

    An automatic particle-size analyzer (PSA) has been developed at ORNL for measuring and counting samples of nuclear fuel microspheres in the diameter range of 300 to 1000 μm at rates in excess of 2000 particles per minute, requiring no sample preparation. A light blockage technique is used in conjunction with a particle singularizer. Each particle in the sample is sized, and the information is accumulated by a multi-channel pulse height analyzer. The data are then transferred automatically to a computer for calculation of mean diameter, standard deviation, kurtosis, and skewness of the distribution. Entering the sample weight and pre-coating data permits calculation of particle density and the mean coating thickness and density. Following this nondestructive analysis, the sample is collected and returned to the process line or used for further analysis. The device has potential as an on-line quality control device in processes dealing with spherical or near-spherical particles where rapid analysis is required for process control

  19. Spent fuel management in France: Reprocessing, conditioning, recycling

    International Nuclear Information System (INIS)

    Giraud, J.P.; Montalembert, J.A. de

    1994-01-01

    The French energy policy has been based for 20 years on the development of nuclear power. The some 75% share of nuclear in the total electricity generation, representing an annual production of 317 TWh requires full fuel cycle control from the head-end to the waste management. This paper presents the RCR concept (Reprocessing, Conditioning, Recycling) with its industrial implementation. The long lasting experience acquired in reprocessing and MOX fuel fabrication leads to a comprehensive industrial organization with minimized impact on the environment and waste generation. Each 900 MWe PWR loaded with MOX fuel avoids piling up 2,500 m 3 per year of mine tailings. By the year 2000, less than 500 m 3 of high-level and long-lived waste will be annually produced at La Hague for the French program. The fuel cycle facilities and the associated MOX loading programs are ramping-up according to schedule. Thus, the RCR concept is a reality as well as a policy adopted in several countries. Last but not least, RCR represents a strong commitment to non-proliferation as it is the way to fully control and master the plutonium inventory

  20. Handbook on process and chemistry of nuclear fuel reprocessing version 2

    International Nuclear Information System (INIS)

    2008-10-01

    Aqueous nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of aqueous reprocessing, because it contributes to establish and develop fuel reprocessing technology and nuclear fuel cycle treating high burn-up UO 2 fuel and spent MOX fuel, and to utilize aqueous reprocessing technology much widely. This handbook is the second edition of the first report, which summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing' from FY 1993 until FY 2000. (author)

  1. Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Carlsen, Brett; Chiguer, Mustapha; Grahn, Per; Sampson, Michele; Wolff, Dietmar; Bevilaqua, Arturo; Wasinger, Karl; Saegusa, Toshiari; Seelev, Igor

    2016-09-01

    Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on

  2. Development of remote fuel pushing system in Reprocessing Plant, Tarapur

    International Nuclear Information System (INIS)

    Chandra, Munish; Coelho, G.; Kodilkar, S.S.; Mishra, A.K.; Bajpai, D.D.; Nair, M.K.T.

    1990-01-01

    Power Reactor Fuel Reprocessing Plant (PREFRE), Tarapur has been processing spent fuel arising from Pressurized Heavy Water Reactors for quite some time. The process adopted in the plant is purex process with chopleach head end treatment. The head end treatment involves loading of ten spent fuel bundles in the charging cask at a time in the fuel bay and aligning the cask with the transfer port and subsequently pushing all the ten bundles together into the fuel magazine. At present the fuel is pushed into the magazine manually. Since the ten bundles weigh approximately 200 Kg. and involves pushing of 9.4 meters length, the operation is carried out using stainless steel screwed pipes, in steps of five lengths. The entire operation requires a large number of trained skilled workers and is found to be tedious. To solve this problem a hydraulic cum pneumatic fuel pushing system has been designed, fabricated, tested and is in the process of installation in the fuel handling area. This paper describes various requirements, constraints and dimensional details arising in the incorporation of such a system to be back fitted in an existing plant, though many of these constraints can be avoided in future plants. Further, complete sequence of operations, technical specifications regarding the telescopic hydraulic power pack and associated controls incorporated in the system are highlighted. (author). 2 figs

  3. Pilot and pilot-commercial plants for reprocessing spent fuels of FBR type reactors

    International Nuclear Information System (INIS)

    Shaldaev, V.S.; Sokolova, I.D.

    1988-01-01

    A review of modern state of investigations on the FBR mixed oxide uranium-plutonium fuel reprocessing abroad is given. Great Britain and France occupy the leading place in this field, operating pilot plants of 5 tons a year capacity. Technology of spent fuel reprocessing and specific features of certain stages of the technological process are considered. Projects of pilot and pilot-commercial plants of Great Britain, France, Japan, USA are described. Economic problems of the FBR fuel reprocessing are touched upon

  4. The main chemical safety problems in main process of nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Song Fengli; Zhao Shangui; Liu Xinhua; Zhang Chunlong; Lu Dan; Liu Yuntao; Yang Xiaowei; Wang Shijun

    2014-01-01

    There are many chemical reactions in the aqueous process of nuclear fuel reprocessing. The reaction conditions and the products are different so that the chemical safety problems are different. In the paper the chemical reactions in the aqueous process of nuclear fuel reprocessing are described and the main chemical safety problems are analyzed. The reference is offered to the design and accident analysis of the nuclear fuel reprocessing plant. (authors)

  5. Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Perkins, W.C.; Durant, W.S.; Dexter, A.H.

    1980-12-01

    The occurrence of certain potential events in nuclear fuel reprocessing plants could lead to significant consequences involving risk to operating personnel or to the general public. This document is a compilation of such potential initiating events in nuclear fuel reprocessing plants. Possible general incidents and incidents specific to key operations in fuel reprocessing are considered, including possible causes, consequences, and safety features designed to prevent, detect, or mitigate such incidents

  6. Apparatus and method for reprocessing and separating spent nuclear fuels

    International Nuclear Information System (INIS)

    Krikorian, O.H.; Grens, J.Z.; Parrish, W.H.; Coops, M.S.

    1983-01-01

    A method and apparatus for separating and reprocessing spent nuclear fuels includes a separation vessel housing a molten metal solvent in a reaction region, a reflux region positioned above and adjacent to the reaction region, and a porous filter member defining the bottom of the separation vessel in a supporting relationship with the metal solvent. Spent fuels are added to the metal solvent. A non-oxidizing nitrogen-containing gas is introduced into the separation vessel, forming solid actinide nitrides in the metal solvent from actinide fuels, while leaving other fission products in solution. A pressure of about 1.1 to 1.2 atm is applied in the reflux region, forcing the molten metal solvent and soluble fission products out of the vessel, while leaving the solid actinide nitrides in the separation vessel. (author)

  7. Development of remote maintenance technology for nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Kawahara, Akira; Saito, Masayuki; Kawamura, Hironobu; Yamade, Atsushi; Sugiyama, Sen; Sugiyama, Sakae.

    1986-01-01

    In the plants for reprocessing spent nuclear fuel containing fission products, due to the facts that the facilities are in high radiations fields, and the surfaces of equipments are contaminated with radioactive substances, the troubles of process equipments are directly connected to the remarkable drop of the rate of operation of the facilities. Therefore, the development of various remote maintenance techniques has been carried out so far, but this time, Hitachi Ltd. got a chance to take part in the repair of spent fuel dissolving tanks in the Tokai Reprocessing Plant of Power Reactor and Nuclear Fuel Development Corp. and the development of several kinds of remote checkup equipment related to the repair work. Especially in the repair of the dissolving tanks, a radiation-withstanding checkup and repair apparatus which has high remote operability taking the conditions of radioactive environment and the restriction of the repaired objects in consideration was required, and a dissolving tank repairing robot composed of six kinds has been developed. The key points of the development were the selective use of high radiation-withstanding parts and materials, small size structure and the realization of full remote operability. The full remote maintenance apparatus of this kind is unique in the world, and applicable to wide fields. (Kako, I.)

  8. Pyrolytic electrochemical process for the reprocessing of irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Brambilla, G.; Sartorelli, A.

    1980-01-01

    The reprocessing is aimed at synthetic UO 2 -PuO 2 mixed oxides, UC-PuC mixed carbides and at oxides and carbides of U, Pu and Th from fast nuclear reactors. The nuclear fuel is dissolved in a salt melting bath. The conversion of the Pu(SO 4 ) 2 is done thermally and that of UO 2 is done electrolytically. The molten salts are returned to the input of the process and the fission products and the molten salts are conditioned. (DG) [de

  9. In-line analytical instrumentation in nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Rao, V.K.; Bhargava, V.K.; Marathe, S.G.

    1979-01-01

    In nuclear fuel reprocessing plants where uranium and plutonium are separated from highly radioactive fission products, continuous monitoring of these constituents is helpful in many ways. Apart from quick detection of possible process malfunctions, in-line monitoring protects operating personnel from radiation hazards, reduces the cost of laboratory analysis and increases the overall efficiency of the process. A review of a proqramme of work on the design, fabrication and testing of some in-line instruments viz. gamma absorptiometer for uranium, neutron monitor for plutonium, acidity monitor for scrub nitric acid etc., their feasibility studies in the laboratory as well as in the pilot plant is presented. (auth.)

  10. Behavior of Nb fission product during nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Gue, J.P.

    1977-02-01

    Investigations on niobium fission product behavior in nitric acid and tributyl phosphate media have been carried out in order to explain the difficulties encountered in separating this element from fissile materials during spent nuclear fuel reprocessing. The studies have shown that in nitric acid solution, pentavalent niobium has a colloidal hydroxide form. The so-obtained sols were characterized by light scattering, electronic microscopy, electrophoresis and ultracentrifugation methods. In heterogeneous extracting media containing tributyl phosphate and dibutyl phosphoric acid the niobium hydroxide sols could be flocculated by low dibutyl phosphoric acid concentration or extracted into the organic phase containing an excess of dibutyl phosphoric acid [fr

  11. The reprocessing of fast reactor fuels - the TOR project

    International Nuclear Information System (INIS)

    Calame-Longjean, A.; Le Bouhellec, J.; Schwob, Y.

    1982-01-01

    A description is given of development work on the proposed new French facility for the reprocessing of fast reactor fuel. This is the TOR facility (Traitement des Oxydes Rapides). Block diagrams give details of the TOR project as a whole and of the main line and R and D line of the TOR 1 facility which is a new works devoted to the head of the process. Modifications to existing plant which will form the TOR 2 and TOR 3 facilities are also described. (U.K.)

  12. Acid in perchloroethylene scrubber solutions used in HTGR fuel preparation processes. Analytical chemistry studies

    International Nuclear Information System (INIS)

    Lee, D.A.

    1979-02-01

    Acids and corrosion products in used perchloroethylene scrubber solutions collected from HTGR fuel preparation processes have been analyzed by several analytical methods to determine the source and possible remedy of the corrosion caused by these solutions. Hydrochloric acid was found to be concentrated on the carbon particles suspended in perchloroethylene. Filtration of carbon from the scrubber solutions removed the acid corrosion source in the process equipment. Corrosion products chemisorbed on the carbon particles were identified. Filtered perchloroethylene from used scrubber solutions contained practically no acid. It is recommended that carbon particles be separated from the scrubber solutions immediately after the scrubbing process to remove the source of acid and that an inhibitor be used to prevent the hydrolysis of perchloroethylene and the formation of acids

  13. Thermodynamic assessment of the HTGR fuel system Th-U-C-O

    International Nuclear Information System (INIS)

    Ugajin, M.; Shiba, K.

    1978-01-01

    Carbon monoxide pressures and uranium segregation at 2000 K have been calculated for the three-phase equilibria [(ThU)O 2 + (ThU)C 2 + C] in the Th-U-C-O system. This study is concerned with the thermochemical behavior of (Th, U)O 2 particle fuel for the high-temperature gas-cooled reactor (HTGR). The following two points are considered: (1) Reduction of the in-particle CO pressure of (Th, U)O 2 kernels by doping (Th, U)C 2 to make it an oxygen getter. (2) Prediction of U segregation between (Th, U)O 2 and (Th, U)C 2 , doped in the kernel. (Auth.)

  14. Design and evaluation of an on-line fuel rod assay device for an HTGR fuel refabrication plant

    International Nuclear Information System (INIS)

    Rushton, J.E.; Allen, E.J.; Chiles, M.M.; Jenkins, J.D.

    1979-11-01

    Refabricated HTGR fuel rods will contain from approx. 0.15 to 0.5 g 233 U and/or 235 U. The fuel rods are approx. 16 mm in diameter and 62 mm long. A typical commercial fuel refabrication facility will have six fuel rod production lines, each producing approximately one fuel rod every 4 seconds at design capacity. One on-line assay device will be present for each two production lines. The relative standard deviation in an individual fuel rod fissile material measurement must be less than 3% to satisfy process and quality control requirements. Systematic errors must be kept less than approx. 0.3% for fissile material measured in fuel rods produced over two months to satisfy material accountability requirements. Several nondestructive assay (NDA) methods were investigated. Because the gamma-ray activity of the refabricated fuel is relatively high due to the presence of 232 U in the fuel and because the gamma-ray activity is not directly related to total or fissile uranium content, NDA methods employing gamma-ray detection did not appear practicable. A method using thermal neutron irradiation and fast-fission neutron detection was selected. An experimental assay device was fabricated based on this NDA method. Experiments were performed to determine the precision and accuracy of the measurements and to investigate potential interferences and systematic errors. Operating procedures were evaluated, and analysis procedures were identified

  15. Reprocessing ability of high density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Gay, A.; Belieres, M.

    1997-01-01

    The development of a new high density fuel is becoming a key issue for Research Reactors operators. Such a new fuel should be a Low Enrichment Uranium (LEU) fuel with a high density, to improve present in core performances. It must be compatible with the reprocessing in an industrial plant to provide a steady back-end solution. Within the framework of a work group CEA/CERCA/COGEMA on new fuel development for Research Reactors, COGEMA has performed an evaluation of the reprocessing ability of some fuel dispersants selected as good candidates. The results will allow US to classify these fuel dispersants from a reprocessing ability point of view. (author)

  16. Reprocessing of spent nuclear fuels in OECD countries

    International Nuclear Information System (INIS)

    1977-01-01

    This report deals with the adequacy of projected reprocessing capacity, the short-term measures proposed in view of the lack of sufficient reprocessing capacity, the longer term measures proposed in view of the lack of sufficient reprocessing capacity, the alternatives to reprocessing and the cooperative arrangements

  17. Critical experiment needs and plans of the consolidated fuel reprocessing program

    International Nuclear Information System (INIS)

    Primm, R.T.

    1984-01-01

    An integral part of the United States Department of Energy (DOE) plan for the development of breeder reactors is the development of the capability for fuel reprocessing. The Consolidated Fuel Reprocessing Program (CFRP) was established by the DOE to identify and conduct research and development activities in this area. The DOE is currently proposing that a capability to reprocess fast reactor fuel be established in the Fuels and Materials Examination Facility at the Hanford Engineering Development Laboratory. This capability would include conversion of plutonium nitrate to plutonium oxide. The reprocessing line is designated the Breeder Reprocessing Engineering Test (BRET). Criticality safety remains an important critetion in the design of the BRET. The different steps in the reprocessing are reviewed and areas where additional critical experiments are needed have been indentified as also areas where revision or clarification of existing criticality safety standards are desirable

  18. Uranium dispersion in the coating of weak-acid-resin-deprived HTGR fuel microspheres

    International Nuclear Information System (INIS)

    Weber, G.W.; Beatty, R.L.; Tennery, V.J.; Lackey, W.J. Jr.

    1976-02-01

    The current reference HTGR recycle fuel particle is a UO 2 /UC 2 kernel with a Triso coating comprising a low-density pyrocarbon (PyC) buffer, a high-density PyC inner LTI coating, SiC, and a high-density PyC outer LTI. The kernel is fabricated from a weak-acid ion exchange resin (WAR). Microradiographic examination of coated WAR particles has demonstrated that considerable U can be transferred from the kernel to the buffer coating during fabrication. Investigation of causes of fuel dispersion has indicated several different factors that contribute to fuel redistribution if not properly controlled. The presence of a nonequilibrium UC/sub 1-x/O/sub x/ (0 less than or equal to x less than or equal to 0.3) phase had no significant effect on initiating fuel dispersion. Gross exposure of the completed fuel kernel to ambient atmosphere or to water vapor at room temperature produced very minimal levels of dispersion. Exposure of the fuel to perchloroethylene during buffer and inner LTI deposition produced massive redistribution. Fuel redistribution observed in Triso-coated particles results from permeation of the inner LTI by HCl during SiC deposition. As the decomposition of CH 3 Cl 3 Si is used to deposit SiC, chlorine is readily available during this process. The permeability of the inner LTI coating has a marked effect on the extent of this mode of fuel dispersion. LTI permeability was determined by chlorine leaching studies to be a strong function of density, coating gas dilution, and coating temperature but relatively unaffected by application of a seal coat, variations in coating thickness, and annealing at 1800 0 C. Mechanical attrition of the kernels during processing was identified as a potential source of U-bearing fines that may be incorporated into the coating in some circumstances

  19. HTGR fuel rods: carbon-carbon composites designed for high weight and low strength

    International Nuclear Information System (INIS)

    Bullock, R.E.

    1977-01-01

    The evolution of the process for fabricating fuel rods for the high-temperature gas-cooled reactor (HTGR) by injection and carbonization of a thermoplastic matrix that bonds close-packed beds of pyrocarbon-coated fuel particles together is reviewed for the fresh-fuel cycle, and a variant process involving a thermosetting matrix that would allow free-standing carbonization of refabricated fuel is discussed. Previous attempts to fabricate such injection-bonded fuel rods from undiluted thermosetting binders filled with powdered graphite were unsuccessful, because of damage to coatings on fuel particles that resulted from strong particle-to-matrix bonding in conjunction with large matrix shrinkage on carbonization and subsequent irradiation. These problems have now been overcome through the use of a diluted thermosetting matrix with a low-char-yield additive (fugitive), which produces a more porous char similar to that from the pitch-based thermoplastic used in fabrication of fresh fuel. A 1-to-1 dilution of resin with fugitive produced the optimum binder for injection and carbonization, where the fired matrix in such rods contained about 20 wt% binder char and 80 wt% powdered graphite. Thermosetting fuel rods diluted with various amounts of fugitive to give binder chars that range from 12 to 48 wt% of the fired matrix have been subjected to irradiation screening tests, and rods with no more than 32 wt% binder char appear to perform about as well under irradiation as do pitch-based rods. However, particle damage does begin to occur in those lightly diluted rods in which the less-stable binder char constitutes more than 32 wt% of the fired matrix. (author)

  20. A review of reprocessing, partitioning, and transmutation of spent nuclear fuel and the implications for Canada

    International Nuclear Information System (INIS)

    Jackson, D.P.

    2006-01-01

    The current status of the reprocessing, partitioning, and transmutation of used nuclear fuel are reviewed in the context of assessing the possible application of these technologies to used CANDU fuel. The status of commercial reprocessing is briefly surveyed and recent progress in world R and D programs on the transmutation of FP's and actinides using Accelerator Driven Systems is summarized. The implications of reprocessing for Canada are explored from the point of view of a long strategy for managing used CANDU fuel in terms of the costs of initiating reprocessing domestically at some time in the future including public and occupational radiation doses, and the wastes generated. (author)

  1. Computational analysis of modern HTGR fuel performance and fission product release during the HFR-EU1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl, E-mail: k.verfondern@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Xhonneux, André, E-mail: xhonneux@lrst.rwth-aachen.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Nabielek, Heinz, E-mail: heinznabielek@me.com [Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Allelein, Hans-Josef, E-mail: h.j.allelein@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, Chair for Reactor Safety and Reactor Technology, 52072 Aachen (Germany)

    2014-07-01

    Highlights: • HFR-EU1 irradiation test demonstrates high quality of HTGR spherical fuel elements. • Irradiation performance is in good agreement with German fuel performance modeling. • International benchmark exercise expected first particle to fail at ∼13–17% FIMA. • EOL silver release is predicted to be in the percentage range. • EOL cesium and strontium are expected to remain at a low level. - Abstract: Various countries engaged in the development and fabrication of modern HTGR fuel have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under HTGR operating and accident conditions. Verification and validation studies are conducted by code-to-code benchmarking and code-to-experiment comparisons as part of international exercises. The methodology developed in Germany since the 1980s represents valuable and efficient tools to describe fission product release from spherical fuel elements and TRISO fuel performance, respectively, under given conditions. Continued application to new results of irradiation and accident simulation testing demonstrates the appropriateness of the models in terms of a conservative estimation of the source term as part of interactions with HTGR licensing authorities. Within the European irradiation testing program for HTGR fuel and as part of the former EU RAPHAEL project, the HFR-EU1 irradiation experiment explores the potential for high performance of the presently existing German and newly produced Chinese fuel spheres under defined conditions up to high burnups. The fuel irradiation was completed in 2010. Test samples are prepared for further postirradiation examinations (PIE) including heatup simulation testing in the KÜFA-II furnace at the JRC-ITU, Karlsruhe, to be conducted within the on-going ARCHER Project of the European Commission. The paper will describe the application of the German computer models to the HFR-EU1 irradiation test and

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  3. Administrative and managerial controls for the operation of nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    Guidelines are provided for the administrative and managerial controls necessary for the safe and efficient operation of nuclear fuel reprocessing plants. Topics covered include: administrative organization; review and audit; facility administrative policies and procedures; and tests and inspections. Recognizing that administrative practices vary among organizations operating nuclear fuel reprocessing plants, the standard incorporates flexibility that provides for compliance by any organization

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

    International Nuclear Information System (INIS)

    Mayorshin, A.A.; Skiba, O.V.; Tsykanov, V.A.; Golovanov, V.N.; Bychkov, A.V.; Kisly, V.A.; Bobrov, D.A.

    2000-01-01

    For advanced nuclear fuel cycle in SSC RIAR there is developed the pyroelectrochemical process to reprocess irradiated fuel and produce granulated oxide fuel UO 2 , PuO 2 or (U,Pu)O 2 from chloride melts. The basic technological stage is the extraction of oxides as a crystal product with the methods either of the electrolysis (UO 2 and UO 2 -PuO 2 ) or of the precipitating crystalIization (PuO 2 ). After treating the granulated fuel is ready for direct use to manufacture vibropacking fuel pins. Electrochemical model for (U,Pu)O 2 coprecipitation is described. There are new processes being developed: electroprecipitation of mixed oxides - (U,Np)O 2 , (U,Pu,Np)O 2 , (U,Am)O 2 and (U,Pu,Am)O 2 . Pyroelectrochemical production of mixed actinide oxides is used both for reprocessing spent fuel and for producing actinide fuel. Both the efficiency of pyroelectrochemical methods application for reprocessing nuclear fuel and of vibropac technology for plutonium recovery are estimated. (author)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  6. High-temperature gas reactor (HTGR) market assessment, synthetic fuels analysis

    International Nuclear Information System (INIS)

    1980-08-01

    This study is an update of assessments made in TRW's October 1979 assessment of overall high-temperature gas-cooled reactor (HTGR) markets in the future synfuels industry (1985 to 2020). Three additional synfuels processes were assessed. Revised synfuel production forecasts were used. General environmental impacts were assessed. Additional market barriers, such as labor and materials, were researched. Market share estimates were used to consider the percent of markets applicable to the reference HTGR size plant. Eleven HTGR plants under nominal conditions and two under pessimistic assumptions are estimated for selection by 2020. No new HTGR markets were identified in the three additional synfuels processes studied. This reduction in TRW's earlier estimate is a result of later availability of HTGR's (commercial operation in 2008) and delayed build up in the total synfuels estimated markets. Also, a latest date for HTGR capture of a synfuels market could not be established because total markets continue to grow through 2020. If the nominal HTGR synfuels market is realized, just under one million tons of sulfur dioxide effluents and just over one million tons of nitrous oxide effluents will be avoided by 2020. Major barriers to a large synfuels industry discussed in this study include labor, materials, financing, siting, and licensing. Use of the HTGR intensifies these barriers

  7. Cost Savings of Nuclear Power with Total Fuel Reprocessing

    International Nuclear Information System (INIS)

    Solbrig, Charles W.; Benedict, Robert W.

    2006-01-01

    The cost of fast reactor (FR) generated electricity with pyro-processing is estimated in this article. It compares favorably with other forms of energy and is shown to be less than that produced by light water reactors (LWR's). FR's use all the energy in natural uranium whereas LWR's utilize only 0.7% of it. Because of high radioactivity, pyro-processing is not open to weapon material diversion. This technology is ready now. Nuclear power has the same advantage as coal power in that it is not dependent upon a scarce foreign fuel and has the significant additional advantage of not contributing to global warming or air pollution. A jump start on new nuclear plants could rapidly allow electric furnaces to replace home heating oil furnaces and utilize high capacity batteries for hybrid automobiles: both would reduce US reliance on oil. If these were fast reactors fueled by reprocessed fuel, the spent fuel storage problem could also be solved. Costs are derived from assumptions on the LWR's and FR's five cost components: 1) Capital costs: LWR plants cost $106/MWe. FR's cost 25% more. Forty year amortization is used. 2) The annual O and M costs for both plants are 9% of the Capital Costs. 3) LWR fuel costs about 0.0035 $/kWh. Producing FR fuel from spent fuel by pyro-processing must be done in highly shielded hot cells which is costly. However, the five foot thick concrete walls have the advantage of prohibiting diversion. LWR spent fuel must be used as feedstock for the FR initial core load and first two reloads so this FR fuel costs more than LWR fuel. FR fuel costs much less for subsequent core reloads ( 6 /MWe. The annual cost for a 40 year licensed plant would be 2.5 % of this or less if interest is taken into account. All plants will eventually have to replace those components which become radiation damaged. FR's should be designed to replace parts rather than decommission. The LWR costs are estimated to be 2.65 cents/kWh. FR costs are 2.99 cents/kWh for the first

  8. Confinement of ruthenium oxides volatilized during nuclear fuels reprocessing

    International Nuclear Information System (INIS)

    Maas, E.T. Jr.; Longo, J.M.

    1980-01-01

    While many materials have been suggested and employed as trapping agents for gaseous oxides of fission product ruthenium volatilized during nuclear fuels reprocessing, none that is known to form thermodynamically stable compounds with rutheniunm has been utilized. We have employed alkaline earth metal compounds for this purpose because of their ability to form stable mixed metal oxide phases with ruthenium. Results of experiments in which RuO 4 was volatilized from either a solid source (RuO 2 .xH 2 O) or from solution [Ru(NO)(NO 3 ) 3 ] in HNO 3 and passed through beds of alkaline earth metal carbonates and calcium oxide held at 600 to 750 0 C have demonstrated that compounds of formulation MRuO 3 (M = calcium, strontium, barium) are formed. Under oxidizing conditions, these materials exist as stable ceramic phases, whereas under reducing conditions, they are transformed into intimate mixtures of the alkaline earth metal oxide and nonvolatile ruthenium metal

  9. Sludge behavior in centrifugal contactor operation for nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Sakamoto, Atsushi; Sano, Yuichi; Takeuchi, Masayuki; Okamura, Nobuo; Koizumi, Kenji

    2015-01-01

    The Japan Atomic Energy Agency (JAEA) has been developing the centrifugal contactor for spent fuel reprocessing. In this study, we investigated the sludge behavior in centrifugal contactors at three different scales. The operational conditions (the flow rate and rotor speed) were varied. Most insoluble particles such as sludge remained in the rotor via centrifugal force. The capture ratio of sludge in the contactor was measured as a function of particle size at various flow rates, rotor speeds, and contactor scales. The sludge adhered and accumulated inside the rotor as the operational time increased, and the operational conditions influenced the capture ratio of the sludge; a lower flow rate and higher rotor speed increased the capture ratio. The results confirmed that Stokes' law can be applied to estimate the experimental result on the behavior of the capture ratio for centrifugal contactors with different scales. (author)

  10. Studies on application on airlift in fuel reprocessing engineering

    International Nuclear Information System (INIS)

    Prasad, A.N.; Balasubramanian, G.R.; Ranganathan, K.

    1977-01-01

    The experiments have been conducted to study the possibility of using airlift for: (1) metering the radioactive fluids by metering the prime air used and (2) transport of these fluids. It is found that airlift can be used for metering directly or a part of a metering system. It can transport radioactive fluids e.g. concentrated plutonium solutions. It can be adopted to transfer completely solutions between tanks at the same level. The problem of entrainment of liquid by air can be sufficiently reduced by introducing suitable de-entrainers. The major advantage is the absence of any moving parts and its wider flow rate ranges. It is, thus, a valuable tool for a fuel reprocessing engineer. (M.G.B.)

  11. Microbial transformations of radionuclides released from nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Francis, A.J.

    2007-01-01

    Microorganisms can affect the stability and mobility of the actinides U, Pu, Cm, Am, Np, and the fission products Tc, I, Cs, Sr, released from nuclear fuel reprocessing plants. Under appropriate conditions, microorganisms can alter the chemical speciation, solubility and sorption properties and thus could increase or decrease the concentrations of radionuclides in solution and the bioavailability. Dissolution or immobilization of radionuclides is brought about by direct enzymatic action or indirect non-enzymatic action of microorganisms. Although the physical, chemical, and geochemical processes affecting dissolution, precipitation, and mobilization of radionuclides have been investigated, we have only limited information on the effects of microbial processes. The mechanisms of microbial transformations of the major and minor actinides and the fission products under aerobic and anaerobic conditions in the presence of electron donors and acceptors are reviewed. (author)

  12. Alternate extractants to tributyl phosphate for reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Crouse, D.J.; Arnold, W.D.; Hurst, F.J.

    1983-01-01

    Both tri(n-hexyl) phosphate (THP) and tri(2-ethylhexyl) phosphate (TEHP) have some important potential process advantages over TBP for reactor fuel reprocessing. These include negligible aqueous phase solubility and less tendency toward third phase and crud formation. The alkyl chain branching of TEHP makes it much more stable to chemical degradation than TBP and probably also accounts for its much weaker ruthenium extraction. The higher uranium and plutonium extraction power of THP and TEHP allows higher solvent loadings in extraction but makes them somewhat more difficult to strip. The phase separation properties of 1.09 M solutions of THP and TEHP are inferior to those of 1.09 M TBP (30 vol %) but are favorable at lower concentrations. Use of more dilute THP and TEHP solutions is recommended for this reason and to obtain a better balance of extraction power in the extraction versus stripping steps

  13. On-line control of nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Parus, I.; Kierzek, J.; Zoltowski, T.

    1977-01-01

    The development trends in the field of chemical processes control and the present state of the development of continuous composition analysers has been described. On this background the peculiarities of on-line control methods for spent nuclear fuel reprocessing have been discussed. The measuring methods for direct and indirect determination of chemical composition and nuclear safety are reviewed in detail. The review comprises such methods as: measurement of α, γ and neutron radiation emitted both by nuclides present in technological solutions and using external sources of different radiation, X-ray fluorescence, measurements of physicochemical parameters connected with the composition (pH, density, electrical conductivity), polarography and spectrophotometry. At the end of this review some new trends in process control based on dynamic process models have been presented. (author)

  14. Advanced teleoperation in nuclear applications: consolidated fuel reprocessing program

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  15. Falling film evaporators: organic solvent regeneration in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Garcin, I.

    1989-01-01

    The aim of this work was to improve knowledge about working of falling film evaporators used in nuclear fuel reprocessing plants for organic solvent regeneration. The first part deals with a non evaporation film. An original film thickness measuring technique was used; infrared thermography. It gave indications on hydrodynamics and wave amplitude and pointed out thermocapillary forces to be the cause of bad wetting of the heated wall. By another way we showed that a small slit spacing on the film distributor, an enhanced surface roughness and an important liquid flow rate favour a better wetting. The second part deals with evaporation of a binary solvent mixture. Experiments in an industrial evaporator corroborated the fact that it is essential for the efficiency of the apparatus to work at high flow rates. We propose an over-simple model which can be used to estimate performances of co-current falling film evaporators of the process [fr

  16. Cleaning and extraction apparatus in a nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Nakamura, Yoshiaki.

    1983-01-01

    Purpose : To eliminate the requirement for the decomposition and cleaning of a centrifugal extractor upon re-processing of FBR type reactor fuels, by preventing solid fission products from depositing on a rotary body of the centrifugal extractor. Constitution : A cleaning and extraction apparatus comprising a combination of a centrifugal cleaner and a centrifugal extractor is used for shortening the contact time between the process liquid and the extraction solvent in FBR type reactor fuel re-processing, and variable parameters are adjusted so that the following equation can be satisfied for avoiding the deposition of solids onto the rotary body of the centrifugal extractor: lsub(e). (rsub(le) 2 + rsub(2r) 2 ) . Nsub(e) . Qsub(c)/ lsub(c) (rsub(lc) 2 + rsub(2c) 2 ) . Nsub(c) . Qsub(e) < 0.8 where Qsub(c) : flow rate to be processed in a centrifugal cleaner, lsub(c) : length of the rotary body, rsub(2c) : radius of a rotary body, rsub(le) : distance from the center to the liquid-extracting hole of the rotary body center to the liquid-extraction hole, Nsub(c) : number of revolution of the rotary body, Qsub(e) : amount of flowrate to be treated in the centrifugal extractor, lsub(e) : length of the rotary body, rsub(2e) : radius for the rotary body, rsub(le) : distance from the center of the rotary body to the liquid discharging aperture and Nsub(e) : number of rotation of the rotary body. (Ikeda, J.)

  17. EdF speaks about economic advantages of fuel reprocessing as compared with interim storage

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    The French company Electricite de France (EdF) will prefer nuclear fuel reprocessing and plutonium recycling to spent fuel storage also in the years after 2000. This option is economically advantageous if the proportional cost of reprocessing does not exceed 1900 FRF/kg heavy metal. Economic analysis shows that this is feasible. EdF will soon have to reprocess annually about 1000 Mt spent fuel to supply enough plutonium for MOX fuel fabrication to feed as many as 28 PWR units and the Superphenix reactor. Spent fuel reprocessing is seen as promising as long as the efficiency of the MOX fuel approaches that of natural uranium based fuel. The French national industrial, political and legal context of EdF operations is also considered. (P.A.)

  18. Status of nuclear fuel reprocessing, spent fuel storage, and high-level waste disposal. Nuclear Fuel Cycle Committee, California Energy Resources Conservation and Development Commission. Draft report

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    An analysis of the current status of technologies and issues in the major portions of the back-end of the nuclear fuel cycle is presented. The discussion on nuclear fuel reprocessing covers the reprocessing requirement, reprocessing technology assessment, technology for operation of reprocessing plants, and approval of reprocessing plants. The chapter devoted to spent fuel storage covers the spent fuel storge problem, the legislative response, options for maintaining full core discharge capacity, prospective availability of alterntive storage options, and the outlook for California. The existence of a demonstrated, developed high-level waste disposal technology is reviewed. Recommendations for Federal programs on high-level waste disposal are made

  19. Monitoring of releases from an irradiated fuel reprocessing plant

    International Nuclear Information System (INIS)

    Fitoussi, L.

    1978-01-01

    At its UP 2 plant, the La Hague facility reprocesses irradiated fuel by the PUREX process. The fuel stems from graphite/gas, natural-uranium reactors and pressurized or boiling water enriched-uranium reactors. The gaseous effluents are collected and purified by high-efficiency washing and filtration. After purification the gas stream is discharged into the atmosphere by a single stack, 100m high and 6m in diameter, located at a high point on the site (184m). The radionuclides released into the air are: krypton-85, iodine-129 and -131, and tritium. The liquid effluents are collected by drainage systems, which transfer them to the effluent treatment station in the case of active or suspect solutions. Active solutions undergo treatment by chemical and physical processes. After purification the waste water is released into the sea by an underwater drainage system 5km long, which brings the outlet point into the middle of a tidal current 2km offshore. The radionuclides contained in the purified waste water are fission products originating from irradiated fuels in only slightly variable proportions, in which ruthenium-rhodium-106 predominates. Traces of the transuranium elements are also found in these solutions

  20. Hydrothermal synthesis for fabrication and reprocessing of MOX nuclear fuel

    International Nuclear Information System (INIS)

    Ohta, Suguru; Yamamura, Tomoo; Shirasaki, Kenji; Satoh, Isamu; Shikama, Tatsuo

    2011-01-01

    To improve the nuclear proliferation resistance and to minimize use of chemicals, a new reprocessing and fabrication process of 'mixed oxide' (MOX) fuel was proposed and studied by using simulated spent fuel solutions. The process is consisting of the two steps, i.e. the removal of fission product (FP) from dissolved spent fuel by using carbonate solutions (Step-1), and hydrothermal synthesis of uranium dioxides (Step-2). In Step-1, rare earth (the precipitation ratio: 90%) and alkaline earth (10-50% for Sr) as FP were removed based on their low solubility of hydroxides and carbonate salts, with uranium kept dissolved for the certain carbonate solutions of weak base (Type 2) or mixtures of relatively strong base and weak base (Type 3). In Step-2, the features of uranium dioxides UO 2+x particles, i.e. stoichiometry (x=0.05-0.2), size (0.2-3 μm) and shape (cubic, spherical, rectangular parallelpiped, etc.), were controlled, and the cesium was removed down to 40 ppm by an addition of organic additives. The decontamination factors (DF) for cesium exceeds 10 5 , whereas the total DF of all the simulated FP were as low as the order of 10 which requires future studies for removal of alkaline earth, Re and Tc etc. (author)

  1. Processes for the control of 14CO2 during reprocessing

    International Nuclear Information System (INIS)

    Notz, K.J.; Holladay, D.W.; Forsberg, C.W.; Haag, G.L.

    1980-01-01

    The fixation of 14 CO 2 may be required at some future time because of the significant fractional contribution of 14 C, via the ingestion pathway, to the total population dose from the nuclear fuel cycle, even though the actual quantity of this dose is very small when compared to natural background. The work described here was done in support of fuel reprocessing development, of both graphite fuel (HTGRs) and metal-clad fuel (LWRs and LMFBRs), and was directed to the control of 14 CO 2 released during reprocessing operations. However, portions of this work are also applicable to the control of 14 CO 2 released during reactor operation. The work described falls in three major areas: (1) The application of liquid-slurry fixation with Ca(OH) 2 , which converts the CO 2 to CaCO 3 , carried out after treatment of the CO 2 -containing stream to remove other gaseous radioactive components, mainly 85 Kr. This approach is primarily for application to HTGR fuel reprocessing. (2) The above process for CO 2 fixation, but used ahead of Kr removal, and followed by a molecular sieve process to take out the 85 Kr. This approach was developed for use with HTGR reprocessing, but certain aspects also have application to metal-clad fuel reprocessing and to reactor operation. (3) The use of solid Ba(OH) 2 hydrate reacting directly with the gaseous phase. This process is generally applicable to both reprocessing and to reactor operation

  2. Development of some operations in technological flowsheet for spent VVER fuel reprocessing at a pilot plant

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Galkin, B.Ya; Lyubtsev, R.I.; Romanovskii, V.N.; Velikhov, E.P.

    1981-01-01

    The fuel reprocessing pilot plants for high active materials would permit the study and development or particular processing steps and flowsheet variations; in some cases, these experimental installations realize on a small scale practically all technological chains of large reprocessing plants. Such a fuel reprocessing pilot plant with capacity of 3 kg U/d has been built at V. G. Khlopin Radium Institute. The pilot plant is installed in the hot cell of radiochemical compartment, and is composed of the equipments for fuel element cutting and dissolving, the preparation of feed solution (clarification, correction), extraction reprocessing and the production of uranium, plutonium and neptunium concentrates, the complex processing of liquid and solid wastes and a special unit for gas purification and analysis. In the last few years, a series of experiments have been carried out on the reprocessing of spent VVER fuel. (J.P.N.)

  3. Review of monitoring instruments for transuranics in fuel fabrication and reprocessing plants. A progress report to the physical and technological programs, Division of Biomedical and Environmental Research, U.S. Energy Research and Development Administration

    International Nuclear Information System (INIS)

    Kordas, J.F.; Phelps, P.L.

    1977-01-01

    A comprehensive review of the monitoring instruments for transuranic elements released from nuclear fuel fabrication and reprocessing plants has been compiled. The extent of routine operational releases has been reviewed for the light water reactor (LWR) fuel cycle (including plutonium recycle), the breeder reactor fuel cycle, and the high-temperature gas cooled reactor (HTGR) fuel cycle. The stack monitoring instrumentation that is presently in use at the various fabrication and reprocessing plants around the country is examined. Sampling difficulties including the inlet-probe arrangement and the effectiveness of the entire sampling system are discussed, as are the measurement problems for alpha-emitting, long-lived, transuranic aerosols, 129 I, 106 Ru, and tritium oxide. The potential problems in the HTGR fuel cycle such as the measurement of releases of alpha-emitting aerosols and of gaseous releases of 220 Rn and 14 C are also considered. Monitoring requirements range from the detection of low-level, routine releases to high-level accidental releases. Both first and second kinds of detection errors are considered in a discussion of adequate detection limits. The presently deployed monitors are critically examined in this light and the drawbacks and limitations of each are noted. Prototype instrumentation is studied, including Argonne's mechanical separation technique, Battelle's mass separation by surface ionization method, and in particular, LLL's transuranic aerosol measurement system. The potentials, sensitivities, advantages, and limitations of each system are enumerated. The additional potential uses of the LLL system are also discussed

  4. Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants

    International Nuclear Information System (INIS)

    Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.

    1977-01-01

    The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of

  5. Remote repair robots for dissolvers in nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Sugiyama, Sen; Hirose, Yasuo; Kawamura, Hironobu; Minato, Akira; Ozaki, Norihiko.

    1984-01-01

    In nuclear facilities, for the purpose of the reduction of radiation exposure of workers, the shortening of working time and the improvement of capacity ratio of the facilities, the technical development of various devices for remote maintenance and inspection has been advanced so far. This time, an occasion came to inspect and repair the pinhole defects occurred in spent fuel dissolving tanks in the reprocessing plant of Tokai Establishment, Power Reactor and Nuclear Fuel Development Corp. However, since the radiation environmental condition and the restricting condition due to the object of repair were extremely severe, it was impossible to cope with them using conventional robot techniques. Consequently, a repair robot withstanding high level radiation has been developed anew, which can work by totally remote operation in the space of about 270 mm inside diameter and about 6 m length. The repair robot comprises a periscope reflecting mirror system, a combined underwater and atmospheric use television, a grinder, a welder, a liquid penetrant tester and an ultrasonic flaw detector. The key points of the development were the parts withstanding high level radiation and the selection of materials, to make the mechanism small size and the realization of totally remote operation. (Kako, I.)

  6. Irradiated-Microsphere Gamma Analyzer (IMGA): an integrated system for HTGR coated particle fuel performance assessment

    International Nuclear Information System (INIS)

    Kania, M.J.; Valentine, K.H.

    1980-02-01

    The Irradiated-Microsphere Gamma Analyzer (IMGA) System, designed and built at ORNL, provides the capability of making statistically accurate failure fraction measurements on irradiated HTGR coated particle fuel. The IMGA records the gamma-ray energy spectra from fuel particles and performs quantitative analyses on these spectra; then, using chemical and physical properties of the gamma emitters it makes a failed-nonfailed decision concerning the ability of the coatings to retain fission products. Actual retention characteristics for the coatings are determined by measuring activity ratios for certain gamma emitters such as 137 Cs/ 95 Zr and 144 Ce/ 95 Zr for metallic fission product retention and 134 Cs/ 137 Cs for an indirect measure of gaseous fission product retention. Data from IMGA (which can be put in the form of n failures observed in N examinations) can be accurately described by the binomial probability distribution model. Using this model, a mathematical relationship between IMGA data (n,N), failure fraction, and confidence level was developed. To determine failure fractions of less than or equal to 1% at confidence levels near 95%, this model dictates that from several hundred to several thousand particles must be examined. The automated particle handler of the IMGA system provides this capability. As a demonstration of failure fraction determination, fuel rod C-3-1 from the OF-2 irradiation capsule was analyzed and failure fraction statistics were applied. Results showed that at the 1% failure fraction level, with a 95% confidence level, the fissile particle batch could not meet requirements; however, the fertile particle exceeded these requirements for the given irradiation temperature and burnup

  7. 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

  8. Ministerial ordinance on the establishment of a reserve fund for spent nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    1984-01-01

    The ministerial ordinance provides for a reserve fund for spent nuclear fuel reprocessing, according to the Electricity Enterprises Act. The Government designates an electricity enterprise that must deposit a reserve fund for spent nuclear fuel reprocessing. The electricity enterprise concerned must deposit a certain sum of money as a reserve fund which is the payment left over from spent fuel reprocessing at the end of a fiscal year minus the same at the end of the preceding year less a certain sum, when the former exceeds the latter. Then, concerning the remainder of the reserve fund in the preceding year, a certain sum must be subtracted from this reserve fund. (Mori, K.)

  9. Distribution of 60Co and 54Mn in graphite material of irradiated HTGR fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Kikuchi, Teruo; Kobayashi, Fumiaki; Minato, Kazuo; Fukuda, Kousaku; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-05-01

    Distribution of 60 Co and 54 Mn was measured in the graphite sleeves and blocks of the third and fourth HTGR fuel assemblies irradiated in the Oarai Gas Loop-1 (OGL-1), which is a high temperature inpile gas loop installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Axial and circumferential profiles were obtained by gamma spectrometry, and radial profiles by lathe sectioning with gamma spectrometry. Distribution of 60 Co is in good agreement with that of thermal neutron flux, and the Co content in the graphite is estimated to be -- 1 x 10 -9 in weight fraction. Concentration of 54 Mn decreases toward the axial center in its axial profile, and radially is almost uniform inside and appreciably higher at free surfaces. An estimated Fe content of --10 -8 in wight fraction is smaller by two orders of magnitude than that from chemical analysis. Higher concentraion of 60 Co and 54 Mn at the free surfaces suggests the importance of transportation process of these nuclides in the coolant loop. (author)

  10. ORCOST-2, PWR, BWR, HTGR, Fossil Fuel Power Plant Cost and Economics

    International Nuclear Information System (INIS)

    Fuller, L.C.; Myers, M.L.

    1975-01-01

    1 - Description of problem or function: ORCOST2 estimates the cost of electrical energy production from single-unit steam-electric power plants. Capital costs and operating and maintenance costs are calculated using base cost models which are included in the program for each of the following types of plants: PWR, BWR, HTGR, coal, oil, and gas. The user may select one of several input/output options for calculation of capital cost, operating and maintenance cost, levelized energy costs, fixed charge rate, annual cash flows, cumulative cash flows, and cumulative discounted cash flows. Options include the input of capital cost and/or fixed charge rate to override the normal calculations. Transmission and distribution costs are not included. Fuel costs must be input by the user. 2 - Method of solution: The code follows the guidelines of AEC Report NUS-531. A base capital-cost model and a base operating- and maintenance-cost model are selected and adjusted for desired size, location, date, etc. Costs are discounted to the year of first commercial operation and levelized to provide annual cost of electric power generation. 3 - Restrictions on the complexity of the problem: The capital cost models are of doubtful validity outside the 500 to 1500 MW(e) range

  11. Design method of control system for HTGR fuel handling process with control Petri net

    International Nuclear Information System (INIS)

    Han Zandong; Luo Sheng; Liu Jiguo

    2008-01-01

    As a complex mechanical system,the fuel handling system (FHS) of pebble-bed high temperature gas cooled reactor (HTGR) is with the features of complicated structure, numerous control devices and strict working scheduling. It is very important to precisely describe the function of FHS and effectively design its control system. A design method of control system based on control Petri net (CPN) is introduced in this paper. By associating outputs and operations with places, associating inputs and conditions with transitions, and introducing macro-places and macro-actions, the CPN realizes hierarchy design of complex control system. Based on the analysis of basic functions and working flow of FHS, its control system is described and designed by CPN. According to the firing regulation of transition,the designed CPN can be easily converted into LAD program of PLC, which can be implemented on the FHS simulating control test-bed. Application illuminates that proposed method has the advantages of clear design structure, exact description power and effective design ability of control program, which can meet the requirements of FHS control sys-tem design. (authors)

  12. Influence of process variables on permeability and anisotropy of Biso-coated HTGR fuel particles

    International Nuclear Information System (INIS)

    Stinton, D.P.; Lackey, W.J.; Thiele, B.A.

    1977-11-01

    The effect of several important process variables on the fraction of defective particles and anisotropy of the low-temperature isotropic (LTI) coating layer was determined for Biso-coated HTGR fuel particles. Process variables considered are deposition temperature, hydrocarbon type, diluent type, and percent diluent. The effect of several other variables such as coating rate and density that depend on the process variables were also considered in this analysis. The fraction of defective particles was controlled by the dependent variables coating rate and LTI density. Coating rate was also the variable controlling the anisotropy of the LTI layer. Diluent type and diluent concentration had only a small influence on the deposition rate of the LTI layer. High-quality particles in terms of anisotropy and permeability can be produced by use of a porous plate gas distributor if the coating rate is between 3 and 5 μm/min and the coating density is between about 1.75 and 1.95 g/cm 3

  13. Advanced Fuel UCO Preparation Technology for HTGR (Characteristics of Carbon Black)

    International Nuclear Information System (INIS)

    Jeong, Kyung Chai; Oh, S. C.; Kim, Y. K.; Cho, M. S.; Kim, W. K.; Kim, Y. M.; Lee, Y. W.; Cho, H. J.; Shin, E. J.

    2010-06-01

    NGNP program for high specification of HTGR nuclear fuel through the GEN IV study is be progressed. Furthermore, because the NGNP program have a highly focused goal like UCO kernel, kernel fabrication and coating types varied which made selection of a US reference fabrication process. In this study, it was evaluated from the reviews on the UO2 and UCO kernel fabrication technologies and its particle characteristics. For improving the UCO qualities, first it was improved the kernel fabrication processes and carbon dispersion method also. New method for carbon dispersion in broth solution was developed, and its characteristics was evaluated from the AGR irradiation tests used the UCO kernel. In fabrication process, also process parameter variation tests in both forming and sintering steps led to an increased understanding of the acceptable ranges for process parameters and additional reduction in required operating times. Another result of this test program was to double the kernel production rate. Following the development tests, approximately 40 kg of natural uranium UCO kernels have been produced for use in coater scale up tests, and approximately 10 kg of low enriched uranium UCO kernels for use in the AGR-2 experiment

  14. The regulations concerning the reprocessing business of spent fuels

    International Nuclear Information System (INIS)

    1987-01-01

    Regulations specified here cover application for such matters as designation of reprocessing undertaking, permission of construction of reprocessing facilities, permission and approval of alteration (of plan for reprocessing facilities), etc. The regulations also cover application for prior inspection, execution of prior inspection, technical standards concerning performance of reprocessing facilities, certificate of prior inspection, reprocessing facilities subject to welding inspection, application for welding inspection, execution of welding inspection, facilities not subject to welding inspection, approval of welding method, welding inspection for imported equipment, certificate of welding inspection, reprocessing facilities subject to regular inspection, application for regular inspection, technical standards for regular inspection, operation plan, application for approval of joint management, record keeping, restriction on access to areas under management, measures concerning exposure to radioactive rays, patrol and checking in reprocessing facilities, operation of reprocessing facilities, self-imposed regular inspection of reprocessing facilities, transportation within plant or operation premises, storage, waste disposal within plant or operation premises, safety rules, notice of disassembly, measures for emergency, notice of abolition of business, notice of disorganization, measures concerning cancellation of designation, submission of report, etc. (Nogami, K.)

  15. The transport of irradiated fuel. An activity closely related to reprocessing

    International Nuclear Information System (INIS)

    Lenail, B.; Curtis, H.W.

    1987-01-01

    With a proven reprocessing capacity of 400 tonnes of uranium per year and the rapid expansion of this capacity, the need to feed the reprocessing plants at La Hague has become vital to ensure continuous and economic reprocessing. The programming of transports by the reprocessor and transporter to ensure a constant supply of fuel for reprocessing has therefore become increasingly important. These transports use the public roads and the railway system and the reprocessor and transporter must cooperate in maintaining the highest possible standards of safety. Safety must take priority over all other factors, including the economics of the operation

  16. On permission of reprocessing project change at the Reprocessing Works of the Japan Nuclear Fuel Ltd. (Reply)

    International Nuclear Information System (INIS)

    1997-01-01

    The Nuclear Safety Commission replied as follows to the Prime Minister on July 14, 1997 on permission of reprocessing project change at the Reprocessing Works of the Japan Nuclear Fuel Ltd. inquired on Dec. 26, 1996. Contents of the inquiry consisted of change of refinery facility and its related instruments, integration of low level wasted liquid treating instrument and change of low level solid waste treating instrument, integration of high level wasted liquid storing building and high level wasted liquid glassification building, installation of used fuel transporting container maintenance instrument and its relating instruments, and so forth. As a result of careful discussion at the Commission for these items, they were admitted to be valid on her technical ability and her safety. (G.K.)

  17. EDRP public local inquiry, UKAEA/BNFL precognition on: PFR fuel reprocessing and radioactive waste management at Dounreay

    International Nuclear Information System (INIS)

    Pugh, O.

    1986-01-01

    A description of PFR fuel reprocessing at Dounreay is given, including brief details of fuel assembly transport, dismantling, chemical separation processes and reprocessing experience. The origin of radioactive wastes from PFR reprocessing, and the types of radioactive waste are outlined. The management of radioactive waste, including storage, treatment and disposal is described. (U.K.)

  18. Continuous solvent extraction feed adjustment for HTGR fuel reprocessing. Interim development report

    International Nuclear Information System (INIS)

    Olguin, L.J.

    1978-06-01

    The two-cycle Acid-Thorex solvent extraction process requires that the feed stream to each thorium cycle be processed to reduce its nitric acid concentration (feed adjustment). This interim development report presents the results of bench-scale and pilot-plant-scale feed adjustment experiments using a continuous mode of operation. An examination of formic acid denitration and fluoride ion volatilization is also included

  19. Recent R/D towards aqueous reprocessing of FBR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Mallika, C.; Pandey, N.K.; Kumar, S.; Kamachi Mudali, U. [Materials, Process and Equipment Development Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2016-07-01

    The mixed Pu-rich carbide spent fuel with a burn up of 155 GWd/t from the Fast Breeder Test Reactor is being reprocessed in a hot-cell facility by PUREX process. Based on the input from the operation of this facility, research and development activities were carried out to improve the recovery, decontamination factors, economy and to reduce the waste volumes. Reduction of uranyl ions in a continuous flow electrochemical reactor and electrolytic as well as chemical reduction of 4 M HNO{sub 3} from liquid waste could be performed in continuous mode. Using the optimized parameters, suitable electrolytic cells/experimental setups were designed for the plant capacity of 6 L/h. Studies on the extraction kinetics of Ru with 30% TBP (tributyl phosphate) in NPH revealed that better decontamination factor with respect to Ru can be achieved using fast contactors like centrifugal extractors (CEs). Towards developing a spent solvent recovery system to reduce organic waste volumes, a pilot plant was set up, which could recover diluent as top product of distillation column and 40% TBP as bottom product from inactive degraded solvent. A solvent recovery system using short path distillation was also developed for installation in hot cells. (authors)

  20. Radiation protection aspects in decommissioning of a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Kotrappa, P.; Joshi, P.P.; Theyyunni, T.K.; Sidhwa, B.M.; Nadkarni, M.N.

    1980-01-01

    The decontamination of a fuel reprocessing plant which underwent partial decommissioning is described. The following radiation protection aspects of the work are discussed: dismantling and removal of process vessels, columns and process off-gas filters; decontamination of various process areas; and management of liquid and solid wastes. The work was completed safely by using personnel protective equipment such as plastic suits and respirators (gas, particulate and fresh air). Total dose commitment for this work was around 3000 man-rems, including dose received by staff for certain jobs related to the operation of a section of the plant. The external dose was kept below the annual limit of 5000 mrems for any individual. No internal contamination incident occurred which caused a dose commitment in excess of 10% of the annual limit. The fact that all the work was completed by the staff normally associated with the operation of the plant contributed significantly to the management and control of personnel exposures. (H.K.)

  1. Alpha-contaminated waste from reprocessing of nuclear fuel

    International Nuclear Information System (INIS)

    Sumner, W.

    1982-01-01

    The anticipated alpha-waste production rates from the Barnwell Nuclear Fuel Reprocessing plant is discussed. The estimated alpha-waste production rate from the 1500 metric ton/year plant is about 85,000 ft 3 /year at the 10 nCi/g limit. Most of this waste is estimated to come from the separation facility, and the major waste sources were cladding, which was 27%, and low-level contact-handled general process trash, which was estimated at 32% of the total. It was estimated that 45% of the waste was combustible and 72% of the waste was compactible. These characteristics could have a significant impact on the final volumes as disposed. Changing the alpha-waste limit from 10 nCi/g to 100 nCi/g was estimated to reduce the amount of alpha waste produced by about 20%. Again, the uncertainty in this value obviously has to be substantial. One has to recognize that these estimates were just that; they were not based on any operating experience. The total plutonium losses to waste, including the high-level waste, was estimated to be 1.5%. The cladding waste was estimated to be contaminated with alpha emitters to the extent of 10 4 to 10 5 nCi/g

  2. Reprocessability of molybdenum and magnesia based inert matrix fuels

    Directory of Open Access Journals (Sweden)

    Ebert Elena L.

    2015-12-01

    Full Text Available This work focuses on the reprocessability of metallic 92Mo and ceramic MgO, which is under investigation for (Pu,MA-oxide (MA = minor actinide fuel within a metallic 92Mo matrix (CERMET and a ceramic MgO matrix (CERCER. Magnesium oxide and molybdenum reference samples have been fabricated by powder metallurgy. The dissolution of the matrices was studied as a function of HNO3 concentration (1-7 mol/L and temperature (25-90°C. The rate of dissolution of magnesium oxide and metallic molybdenum increased with temperature. While the MgO rate was independent of the acid concentration (1-7 mol/L, the rate of dissolution of Mo increased with acid concentration. However, the dissolution of Mo at high temperatures and nitric acid concentrations was accompanied by precipitation of MoO3. The extraction of uranium, americium, and europium in the presence of macro amounts of Mo and Mg was studied by three different extraction agents: tri-n-butylphosphate (TBP, N,Nʹ-dimethyl-N,Nʹ-dioctylhexylethoxymalonamide (DMDOHEMA, and N,N,N’,N’- -tetraoctyldiglycolamide (TODGA. With TBP no extraction of Mo and Mg occurred. Both matrix materials are partly extracted by DMDOHEMA. Magnesium is not extracted by TODGA (D < 0.1, but a weak extraction of Mo is observed at low Mo concentration.

  3. Development of pulsed plate columns for fast reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Jenkins, J.A.; Logsdail, D.H.; Lyall, E.; Myers, P.E.; Partridge, B.A.

    1987-01-01

    The UK Atomic Energy Authority has undertaken a development programme on solvent extraction equipment for reprocessing fast reactor fuels. As part of this programme a solvent extraction pilot plant has been built at Harwell in which a variety of flowsheet conditions can be simulated using the system uranyl nitrate/nitric acid (UN/HNO 3 ) - 20% tri-n-butyl phosphate in odourless kerosene (TBP/OK). The main purpose of present pilot plant operations is to study the performance of pulsed plate columns, with the following specific objectives: to measure the volumetric throughput capacity of the columns, - to study the effect of scale-up of column diameter on U mass transfer performance, - to provide hydraulic and mass transfer data for a dynamic simulation model of pulsed column operation, - to develop and test instruments and ancillary equipment. This poster describes the pilot plant and is illustrated by experimental data, with particular reference to an external settler for controlling the removal of aqueous phase from columns operated with the aqueous phase dispersed

  4. Crud in the solvent extraction process for spent fuel reprocessing

    International Nuclear Information System (INIS)

    Chen Jing

    2004-01-01

    The crud occurred in Purex process is caused by the degradations of extractant and solvent and the existence of insoluble solid particle in the nuclear fuel reprocessing. The crud seriously affects the operation of the extraction column. The present paper reviews the study status on the crud in the Purex process. It is generally accepted that in the Purex process, particularly in the first cycle, the crud occurrence is related to the capillary chemistry phenomena resulting from the deposits of Zr with TBP degradation products HDBP, H 2 MBP, H 3 PO 4 and the insoluble particle RuO 2 and Pd. The occurrence of deposits and the type of crud are tightly related to the molar ratio of HDBP and Zr, and the aqueous pH. In addition, the effect of degradation products from the diluent, such as kerosene, is an unnegligible factor to cause the crud. The crud can be discharged from the extraction equipment with Na 2 CO 3 or oxalic acid. In the study on simulating the crud, the effects of the deposits of Zr with TBP degradation products HDBP, H 2 MBP and H 2 PO 4 , and the insoluble particle RuO 2 and Pd should be considered at the same time. (authors)

  5. Thermochemical properties of media for pyrometallurgical nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Hosoya, Yuji; Terai, Takayuki

    1998-01-01

    Molten chloride/cadmium system is considered to be applied to a solvent in pyrochemical reprocessing of spent nuclear fuel. In this work, phase diagrams for molten chloride systems were constructed, using NdCl 3 as an imitative substance in place of UCl 3 or PuCl 3 . Hastelloy-X (Ni/Cr21/Fe18/Mo9/W) was examined as a structural material for the corrosion-resistance against molten chloride baths containing NdCl 3 . The process of corrosion was thermochemically discussed and the form of the corrosion was illustrated. Rutherford backscattering spectroscopy was successfully applied to determine the elemental distribution profile of specimens tested on the compatibility with molten chloride mixture at elevated temperature. Ferritic steel was also examined as another candidate material for the compatibility with molten cadmium covered with LiCl-KCl eutectic salt. Variation of near-surface composition was observed by comparing the results of Rutherford backscattering spectroscopy obtained before and after the dipping. (author)

  6. Status of radioiodine control for nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Burger, L.L.; Scheele, R.D.

    1983-07-01

    This report summarizes the status of radioiodine control in a nuclear fuel reprocessing plant with respect to capture, fixation, and disposal. Where possible, we refer the reader to a number of survey documents which have been published in the last four years. We provide updates where necessary. Also discussed are factors which must be considered in developing criteria for iodine control. For capture from gas streams, silver mordenite and a silver nitrate impregnated silica (AC-6120) are considered state-of-the-art and are recommended. Three aqueous scrubbing processes have been demonstrated: Caustic scrubbing is simple but probably will not give an adequate iodine retention by itself. Mercurex (mercuric nitrate-nitric acid scrubbing) has a number of disadvantages including the use of toxic mercury. Iodox (hyperazeotropic nitric acid scrubbing) is effective but employs a very corrosive and hazardous material. Other technologies have been tested but require extensive development. The waste forms recommended for long-term storage or disposal are silver iodide, the iodates of barium, strontium, or calcium, and silver loaded sorbents, all fixed in cement. Copper iodide in bitumen (asphalt) is a possibility but requires testing. The selection of a specific form will be influenced by the capture process used

  7. Solvent extraction for spent nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Masui, Jinichi

    1986-01-01

    The purex process provides a solvent extraction method widely used for separating uranium and plutonium from nitric acid solution containing spent fuel. The Tokai Works has adopted the purex process with TPB-n dodecane as the extraction agent and a mixer settler as the solvent extraction device. The present article outlines the solvent extraction process and discuss the features of various extraction devices. The chemical principle of the process is described and a procedure for calculating the number of steps for countercurrent equilibrium extraction is proposed. Discussion is also made on extraction processes for separating and purifying uranium and plutonium from fission products and on procedures for managing these processes. A small-sized high-performance high-reliability device is required for carrying out solvent extraction in reprocessing plants. Currently, mixer settler, pulse column and centrifugal contactor are mainly used in these plants. Here, mixer settler is comparted with pulse column with respect to their past achievements, design, radiation damage to solvent, operation halt, controllability and maintenance. Processes for co-extraction, partition, purification and solvent recycling are described. (Nogami, K.)

  8. Design aspects of water usage in the Windscale nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Wharton, J.; Bullock, M.J.

    1982-01-01

    The safeguard requirements of a nuclear fuel reprocessing plant place unique constraints on a designer which, in turn, affect the scope for the exercise of water economy. These constraints are examined within the context of the British Nuclear Fuels Limited reprocessing plants at Windscale and indicate the scope for water conservation. The plants and their design principles are described with particular reference to water services and usage. Progressive design development is discussed to illustrate the increasing importance of water economy. (author)

  9. Laser-enhanced chemical reactions and the liquid state. II. Possible applications to nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    DePoorter, G.L.; Rofer-DePoorter, C.K.

    1976-01-01

    Laser photochemistry is surveyed as a possible improvement upon the Purex process for reprocessing spent nuclear fuel. Most of the components of spent nuclear fuel are photochemically active, and lasers can be used to selectively excite individual chemical species. The great variety of chemical species present and the degree of separation that must be achieved present difficulties in reprocessing. Lasers may be able to improve the necessary separations by photochemical reaction or effects on rates and equilibria of reactions

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

    International Nuclear Information System (INIS)

    Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya

    2001-05-01

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

  11. Nondestructive determination of residual fuel on leached hulls and dissolver sludges from LWR fuel reprocessing

    International Nuclear Information System (INIS)

    Wuerz, H.; Wagner, K.; Becker, H.J.

    1990-01-01

    In reprocessing plants leached hulls and dissolver sludges represent rather important intermediate level α-waste streams. A control of the Pu content of these waste streams is desirable. The nondestructive assay method to be preferred would be passive neutron counting. However, before any decision on passive neutron monitoring becomes possible a characterization of hulls and sludges in terms of Pu content and neutron emission is necessary. For the direct determination of plutonium on hulls and in sludges, as coming from reprocessing, an active neutron measurement is required. A simple, and sufficiently sensitive active neutron method which can easily be installed uses as stationary Cf-252 neutron source. This method was used for the characterization of hulls and sludges in terms of plutonium content and total neutron emission in the WAK. Meanwhile a total of 28 batches of leached hulls and 22 batches of dissolver sludges from reprocessing of PWR fuel have been assayed. The paper describes the assay method used and gives an analysis of the error sources together with a discussion of the results and the accuracies obtained in a reprocessing plant. (orig./HP)

  12. Economic assessment factors relating to spent nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    This paper is in two parts. Part I discusses the factors to be applied in an economic assessment of reprocessing. It sets forth three basic cost components, namely capital costs, operating costs and the cost of capital utilization. It lists the various components of each cost area. Part II proposes a relationship between these respective cost areas, tabulates a range of costs and then develops unit costs for reprocessing operations. Finally, an addendum to the paper gives a more detailed breakdown of the capital costs of a reprocessing plant

  13. Status and trends in spent fuel reprocessing. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    1999-08-01

    Spent fuel management has always been an important part of the nuclear fuel cycle and is still one of the most important activities in all countries exploiting the peaceful use of nuclear energy. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for exchanging information and to coordinate and encourage closer co-operation among Member States in certain research and developing activities that are of common interest. As part of spent fuel management, reprocessing activities have been reviewed from time to time on a low profile level under the terminology 'spent fuel treatment'. However, spent fuel treatment covers, in broad terms, spent fuel storage (short, interim and long term), fuel rod consolidation, reprocessing and, in case the once-through cycle is selected, conditioning of the spent fuel for disposal. Hence the reprocessing activities under the heading 'spent fuel treatment' were somewhat misleading. Several meetings on spent fuel treatment have been organized during the fast decade: an Advisory Group meeting (AGM) in 1992, a Technical Committee meeting in 1995 and recently an Advisory Group meeting from 7 to 10 September 1998. The objectives of the meetings were to review the status and trends of spent fuel reprocessing, to discuss the environmental impact and safety aspects of reprocessing facilities and to define the most important issues in this field. Notwithstanding the fact that the Summary of the report does not include aspects of military reprocessing, some of the national presentations do refer to some relevant aspects (e.g. experience, fissile stockpiles)

  14. Spent fuel handling and storage facility for an LWR fuel reprocessing plant

    International Nuclear Information System (INIS)

    Baker, W.H.; King, F.D.

    1979-01-01

    The facility will have the capability to handle spent fuel assemblies containing 10 MTHM/day, with 30% if the fuel received in legal weight truck (LWT) casks and the remaining fuel received in rail casks. The storage capacity will be about 30% of the annual throughput of the reprocessing plant. This size will provide space for a working inventory of about 50 days plant throughput and empty storage space to receive any fuel that might be in transit of the reprocessing plant should have an outage. Spent LWR fuel assemblies outside the confines of the shipping cask will be handled and stored underwater. To permit drainage, each water pool will be designed so that it can be isolated from the remaining pools. Pool water quality will be controlled by a filter-deionizer system. Radioactivity in the water will be maintained at less than or equal to 2 x 10 -4 Ci/m 3 ; conductivity will be maintained at 1 to 2 μmho/cm. The temperature of the pool water will be maintained at less than or equal to 40 0 C to retard algae growth and reduce evaporation. Decay heat will be transferred to the environment via a heat exchanger-cooling tower system

  15. Simulation of spent fuel reprocessing processes: Realizations and prospects

    International Nuclear Information System (INIS)

    Boullis, B.

    1986-12-01

    The separation of uranium and plutonium in the Purex process is very complex and for the extension of reprocessing plants optimization of the process requires mathematical modelling. The development of this model is reviewed [fr

  16. Biodegradation of radioactive organic liquid waste from spent fuel reprocessing

    International Nuclear Information System (INIS)

    Ferreira, Rafael Vicente de Padua

    2008-01-01

    The research and development program in reprocessing of low burn-up spent fuel elements began in Brazil in 70's, originating the lab-scale hot cell, known as Celeste located at Nuclear and Energy Research Institute, IPEN - CNEN/SP. The program was ended at the beginning of 90's, and the laboratory was closed down. Part of the radioactive waste generated mainly from the analytical laboratories is stored waiting for treatment at the Waste Management Laboratory, and it is constituted by mixture of aqueous and organic phases. The most widely used technique for the treatment of radioactive liquid wastes is the solidification in cement matrix, due to the low processing costs and compatibility with a wide variety of wastes. However, organics are generally incompatible with cement, interfering with the hydration and setting processes, and requiring pre -treatment with special additives to stabilize or destroy them. The objective of this work can be divided in three parts: organic compounds characterization in the radioactive liquid waste; the occurrence of bacterial consortia from Pocos de Caldas uranium mine soil and Sao Sebastiao estuary sediments that are able to degrade organic compounds; and the development of a methodology to biodegrade organic compounds from the radioactive liquid waste aiming the cementation. From the characterization analysis, TBP and ethyl acetate were chosen to be degraded. The results showed that selected bacterial consortia were efficient for the organic liquid wastes degradation. At the end of the experiments the biodegradation level were 66% for ethyl acetate and 70% for the TBP. (author)

  17. Removal of actinides from selected nuclear fuel reprocessing wastes

    International Nuclear Information System (INIS)

    Navratil, J.D.; Thompson, G.H.

    1979-01-01

    The US Department of Energy awarded Oak Ridge National Laboratory a program to develop a cost-risk-benefit analysis of partitioning long-lived nuclides from waste and transmuting them to shorter lived or stable nuclides. Two subtasks of this program were investigated at Rocky Flats. In the first subtask, methods for solubilizing actinides in incinerator ash were tested. Two methods appear to be preferable: reaction with ceric ion in nitric acid or carbonate-nitrate fusion. The ceric-nitric acid system solubilizes 95% of the actinides in ash; this can be increased by 2 to 4% by pretreating ash with sodium hydroxide to solubilize silica. The carbonate-nitrate fusion method solubilizes greater than or equal to 98% of the actinides, but requires sodium hydroxide pretreatment. Two additional disadvantages are that it is a high-temperature process, and that it generates a lot of salt waste. The second subtask comprises removing actinides from salt wastes likely to be produced during reactor fuel fabrication and reprocessing. A preliminary feasibility study of solvent extraction methods has been completed. The use of a two-step solvent extraction system - tributyl phosphate (TBP) followed by extraction with a bidentate organophosphorous extractant (DHDECMP) - appears to be the most efficient for removing actinides from salt waste. The TBP step would remove most of the plutonium and > 99.99% of the uranium. The second step using DHDECMP would remove > 99.91% of the americium and the remaining plutonium (> 99.98%) and other actinides from the acidified salt waste. 8 figures, 11 tables

  18. Radiation resistant polymers and coatings for nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Kamachi Mudali, U.; Mallika, C.; Lawrence, Falix

    2014-01-01

    Polymer based materials are extensively used in the nuclear industry for the reprocessing of spent fuels in highly radioactive and corrosive environment. Hence, these polymer materials are susceptible to damage by ionizing radiation, resulting in the degradation in properties. Polymers containing aromatic molecules generally possess higher resistance to radiation degradation than the aliphatic polymers. For improving the radiation resistance of polymers various methods are reported in the literature. Among the aromatic polymers, polyetheretherketone (PEEK) has the radiation tolerance up to 10 Mega Grey (MGy). To explore the possibility of enhancing the radiation resistance of PEEK, a study was initiated to develop PEEK - ceramic composites and evaluate the effect of radiation on the properties of the composites. PEEK and PEEK - alumina (micron size) composites were irradiated in a gamma chamber using 60 Co source and the degradation in mechanical, structural, electrical and thermal properties, gel fraction, coefficient of friction and morphology were investigated. The degradation in the mechanical properties owing to radiation could be reduced by adding alumina filler to PEEK. Nano alumina filler was observed to be more effective in suppressing the damage caused by radiation on the polymer, when compared to micron alumina filler. For the protection of aluminium components in the manipulators and the rotors and stators of the motors of the centrifugal extractors employed in the plant from the attack by nitric acid vapour, PEEK coating based on liquid dispersion was developed, which has resistance to radiation, chemicals and wear. The effect of radiation and chemical vapour on the properties of the PEEK coating was estimated. The performance of the coating in the plant was evaluated and the coating was found to give adequate protection to the motors of centrifugal extractors against corrosion. (author)

  19. Management and disposal of used nuclear fuel and reprocessing wastes

    International Nuclear Information System (INIS)

    1983-01-01

    The subject is dealt with in chapters, entitled: introduction (general statement of problem); policy framework (criteria for waste management policy); waste management and disposal, as practised and planned (general; initial storage; reprocessing and conditioning of reprocessing wastes; intermediate storage; transportation; packaging; disposal); international co-operation. Details of the situation in each country concerned (Australia, Belgium, Canada, France, Federal Republic of Germany, Spain, Sweden, Switzerland and United Kingdom) are included as annexes. (U.K.)

  20. Fuel salt reprocessing influence on the MSFR behavior and on its associated reprocessing unit

    International Nuclear Information System (INIS)

    Doligez, X.

    2010-10-01

    In order to face with the growing of the energy demand, the nuclear industry has to reach the fourth generation technology. Among those concept, molten salt reactor, and especially the fast neutron spectrum configuration, seems very promising: indeed breeding is achievable while the feedback coefficient are still negative. However, the reprocessing salt scheme is not totally set down yet. A lot of uncertainties remain on chemical properties of the salt. Thanks to numerical simulation we studied the behavior of the molten Salt Fast Reactor coupled to a nominal reprocessing unit. We are now able to determine heat transfer and radiation in each elementary step of the unit and, by this way determine those that need special study for radioprotection. We also studied which elements are fundamental to extract for the reactor operation. Finally, we present a sensibility analysis of the chemical uncertainties to few relevant properties of the reactor behavior. (author)

  1. Economic feasibility study of regional centers for nuclear fuel reprocessing in the developing countries

    International Nuclear Information System (INIS)

    Bakeshloo, A.A.

    1977-01-01

    The fuel cycle costs for the following three different economic alternatives were studied: (1) Reprocessing in an industrialized country (such as the U.S.); (2) Reprocessing in the individual developing country; (3) Reprocessing in a regional center. The nuclear fuel cycle cost for the ''Throw-away'' fuel cycle was evaluated. Among the six regions which were considered in this study, region one (South America including Mexico) was selected for the economic analysis of the nuclear fuel cycle for the above three alternatives. For evaluation of the cases where the fuel is reprocessed in a regional center or in an individual developing country, a unit reprocessing cost equation was developed. An economic evaluation was developed to estimate the least expensive method for transporting radioactive nuclear material by either leased or purchased shipping casks. The necessary equations were also developed for estimating plutonium transportation and the safeguard costs. On the basis of nuclear material and services requirements and unit costs for each component, the levelized nuclear fuel cycle costs for each alternative were estimated. Finally, by a comparison of cost, among these three alternatives plus the ''Throw-away'' case,it was found that it is not at all economical to build individual reprocessing plants inside the developing countries in region one. However, it also was found that the economic advantage of a regional center with respect to the first alternative is less than a 4% difference between their total fuel cycle costs. It is concluded that there is no great economic advantage in any developing countries to seek to process their fuel in one of the advanced countries. Construction of regional reprocessing centers is an economically viable concept

  2. The economics of reprocessing versus direct disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Bunn, M.; Holdren, J.P.; Fetter, S.; Zwaan, B. van der

    2007-01-01

    The economics of reprocessing versus direct disposal of spent nuclear fuel are assessed. The break-even uranium price at which reprocessing spent nuclear fuel from existing light water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is estimated for a wide range of reprocessing prices and other fuel cycle costs and parameters. The contribution of each fuel cycle option to the cost of electricity is also estimated. A similar analysis is performed for the breakeven uranium price at which deploying fast neutron reactors (FRs) would become competitive compared with a once-through fuel cycle in LWRs, for a range of differences in capital cost between LWRs and FRs. Available information about reprocessing prices and various other fuel cycle costs and input parameters are reviewed, as well as the quantities of uranium likely to be recoverable worldwide at a range of different possible future prices. It is concluded that the once-through fuel cycle is likely to remain significantly cheaper than reprocessing and recycling in either LWRs or FRs for at least the next 50 years. Finally, there is a discussion of how scarce and expensive repository space would have to become before separation and transmutation would be economically attractive. (author)

  3. Prospect of spent fuel reprocessing and back-end cycling in China in 1990's

    International Nuclear Information System (INIS)

    Ke Youzhi; Wang Rengtao

    1987-01-01

    According to the CHinese Program of nuclear energy in 1990's, the amount of spent fuel by the year 2000 is estimated in this paper. Reprocessing is considered as an important link in the back-end fuel cycle. A pilot plant is scheduled for hot start up in 1996. The main goal of the study is LWR spent fuel reprocessing. We will use the experience gained from reprocessing of production reactor fuel and last research results. The advanced foreign technigue and experience will be introduced. The study emphasizes on the test of technology, equipments, instrumentation and automation, development of remote maintenance and decontamination. China will start to demonstrate the way for fuel cycle. (author)

  4. Overview of reductants utilized in nuclear fuel reprocessing/recycling

    Energy Technology Data Exchange (ETDEWEB)

    Patricia Paviet-Hartmann; Catherine Riddle; Keri Campbell; Edward Mausolf

    2013-10-01

    promises as a replacement for AHA. FHA undergoes hydrolysis to formic acid which is volatile, thus allowing the recycling of nitric acid. Unfortunately, FHA powder was not stable in the experiments we ran in our laboratory. In addition, AHA and FHA also decompose to hydroxylamine which may undergo an autocatalytic reaction. Other reductants are available and could be extremely useful for actinides separation. The review presents the current plutonium reductants used in used nuclear fuel reprocessing and will introduce innovative and novel reductants that could become reducers for future research on UNF separation.

  5. Consolidated fuel reprocessing programme: Analysis of various options for the breeder fuel cycle in the USA

    International Nuclear Information System (INIS)

    Stradley, J.G.; Burch, W.D.; Yook, H.R.

    1986-01-01

    The United States Department of Energy (DOE) has established a programme to develop innovative liquid metal reactor (LMR) designs to assist in developing future U.S. reactor strategy. The paper describes studies in progress to examine various fuel cycle strategies that relate to the reactor strategy. Three potential fuel cycle options that focus on supporting an initial 1300 MW(e) reactor station have been defined: (1) Completion and utilization of the Breeder Reprocessing Engineering Test/Secure Automated Fabrication (BRET/SAF) in the Fuels and Materials Examination Facility (FMEF) at Hanford, Washington; (2) a co-located fuel cycle facility; and (3) delayed closure of the fuel cycle for five to ten years. The BRET, designed as a development facility, has sufficient capacity to service the needs of an initial module at an LMR station. It appears feasible to increase this capacity and to utilize SAF in the FMEF to accommodate the projected output (up to 35 MtHM/year) from the 1300 MW(e) liquid-metal concepts under study. Plans developed within the United States Consolidated Management Office for an initial reactor project have envisioned that cost savings could be realized by delaying the closure of the fuel cycle as long as supplies of plutonium could be obtained relatively inexpensively. This might prove to be only five to ten years, but even that period might be long enough for the fuel cycle costs to be spread over more than one reactor rather than loaded on the initial project. This concept is being explored as is the question of the future coupling of a light water reactor reprocessing industry for plutonium supply to breeder recycle

  6. Reprocessing techniques of LWR spent fuel for reutilization in hybrid systems and IV generation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Aruquipa, Wilmer; Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Barros, Graiciany de P. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    Since the era of nuclear technology begins, nuclear reactors have been produced spent fuel. This spent fuel contains material that could be recycle and reprocessed by different processes. All these processes aim to reduce the contribution to the final repository through the re-utilization of the nuclear material. Therefore, some new reprocessing options with non-proliferation characteristics have been proposed and the goal is to compare the different techniques used to maximize the effectiveness of the spent fuel utilization and to reduce the volume and long-term radiotoxicity of high-level waste by irradiation with neutron with high energy such as the ones created in hybrid reactors. In order to compare different recovery methods, the cross sections of fuels are calculated with de MCNP code, the first set consists of thorium-232 spiked with the reprocessed material and the second set in depleted uranium that containing 4.5% of U-235 spiked with the reprocessed material; These sets in turn are compared with the cross section of the UO{sub 2} in order to evaluate the efficiency of the reprocessed fuel as nuclear fuel. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  9. Preliminary concepts: coordinated safeguards for materials management in a thorium--uranium fuel reprocessing plant

    International Nuclear Information System (INIS)

    Hakkila, E.A.; Barnes, J.W.; Dayem, H.A.; Dietz, R.J.; Shipley, J.P.

    1978-10-01

    This report addresses preliminary concepts for coordinated safeguards materials management in a typical generic thorium--uranium-fueled light-water reactor (LWR) fuels reprocessing plant. The reference facility is designed to recover thorium and uranium from first-generation (denatured 235 U) startup fuels, first-recycle and equilibrium (denatured 233 U) thorium--uranium LWR fuels, and to recover the plutonium generated in the 238 U denaturant as well. 12 figures, 3 tables

  10. Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    1978-01-01

    The report gives a general summary of the Swedish KBS-project on management and disposal of vitrified reprocessed waste. Its final aim is to demostrate that the means of processing and managing power reactor waste in an absolutely safe way, as stipulated in the Swedish so called Conditions Act, already exist. Chapters on Storage facility for spent fuel, Intermidiate storage of reprocessed waste, Geology, Final repository, Transportation, Protection, and Siting. (L.E.)

  11. Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    1978-01-01

    A summary of the planning of transportation and plant design in the Swedish KBS project on management and disposal reprocessed radioactive waste. It describes a transportation system, a central storage facility for used fuel elements, a plant for intermediate storage and encapsulation and a final repository for the vitrified waste. Accounts are given for the reprocessing and vitrification. The safety of the entire system is discussed

  12. Nuclear material inventory estimation in a nuclear fuel reprocessing facility

    International Nuclear Information System (INIS)

    Bennett, J.E.; Beyerlein, A.L.

    1981-01-01

    A new approach in the application of modern system identification and estimation techniques is proposed to help nuclear reprocessing facilities meet the nuclear accountability requirement proposed by the International Atomic Energy Agency. The proposed identification and estimation method considers the material inventory in a portion of the chemical separations area of a reprocessing facility. The method addresses the nonlinear aspects of the problem, the time delay through the separation facility, and the lack of measurement access. The method utilizes only input-output measured data and knowledge of the uncertainties associated with the process and measured data. 14 refs

  13. Extraction of Uranium Using Nitrogen Dioxide and Carbon Dioxide for Spent Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Kayo Sawada; Daisuke Hirabayashi; Youichi Enokida [EcoTopia Science Institute, Nagoya University, Nagoya, 464-8603 (Japan)

    2008-07-01

    For the reprocessing of spent nuclear fuels, a new method to extract actinides from spent fuel using highly compressed gases, nitrogen dioxide and carbon dioxide was proposed. Uranium extraction from broken pieces, whose average grain size was 5 mm, of uranium dioxide pellet with nitrogen dioxide and carbon dioxide was demonstrated in the present study. (authors)

  14. Simulation of facility operations and materials accounting for a combined reprocessing/MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Coulter, C.A.; Whiteson, R.; Zardecki, A.

    1991-01-01

    We are developing a computer model of facility operations and nuclear materials accounting for a facility that reprocesses spent fuel and fabricates mixed oxide (MOX) fuel rods and assemblies from the recovered uranium and plutonium. The model will be used to determine the effectiveness of various materials measurement strategies for the facility and, ultimately, of other facility safeguards functions as well. This portion of the facility consists of a spent fuel storage pond, fuel shear, dissolver, clarifier, three solvent-extraction stages with uranium-plutonium separation after the first stage, and product concentrators. In this facility area mixed oxide is formed into pellets, the pellets are loaded into fuel rods, and the fuel rods are fabricated into fuel assemblies. These two facility sections are connected by a MOX conversion line in which the uranium and plutonium solutions from reprocessing are converted to mixed oxide. The model of the intermediate MOX conversion line used in the model is based on a design provided by Mike Ehinger of Oak Ridge National Laboratory (private communication). An initial version of the simulation model has been developed for the entire MOX conversion and fuel fabrication sections of the reprocessing/MOX fuel fabrication facility, and this model has been used to obtain inventory difference variance estimates for those sections of the facility. A significant fraction of the data files for the fuel reprocessing section have been developed, but these data files are not yet complete enough to permit simulation of reprocessing operations in the facility. Accordingly, the discussion in the following sections is restricted to the MOX conversion and fuel fabrication lines. 3 tabs

  15. FRESCO-II: A computer program for analysis of fission product release from spherical HTGR-fuel elements in irradiation and annealing experiments

    International Nuclear Information System (INIS)

    Krohn, H.; Finken, R.

    1983-06-01

    The modular computer code FRESCO has been developed to describe the mechanism of fission product release from a HTGR-Core under accident conditions. By changing some program modules it has been extended to take into account the transport phenomena (i.e. recoil) too, which only occur under reactor operating conditions and during the irradiation experiments. For this report, the release of cesium and strontium from three HTGR-fuel elements has been evaluated and compared with the experimental data. The results show that the measured release can be described by the considered models. (orig.) [de

  16. Technical specifications on the welding in fuel reprocessing plants

    Energy Technology Data Exchange (ETDEWEB)

    Karino, Motonobu; Uryu, Mitsuru; Matsui, N.; Nakazawa, Fumio; Imanishi, Makoto; Koizumi; Kazuhiko; Sugawara, Junichi; Tanaka, Hideo

    1999-04-01

    The past specifications SGN of the welding in JNC was reexamined for the reprocessing plants in order to further promote the quality control. The specification first concerns the quality of raw materials, items of the quality tests, material management, and qualification standards of the welders. It extends over details of the welding techniques, welding design, welding testings, inspection and the judgment standards. (H. Baba)

  17. Development of centrifugal contactor for FBR fuel reprocessing

    International Nuclear Information System (INIS)

    Washiya, Tadahiro; Takeuchi, Masayuki; Suganuma, Takashi; Aose, Shinichi; Ogino, Hideki

    2003-01-01

    In the Feasibility Study on Commercialized Fast Reactor Cycle Systems, the aqueous reprocessing technology is nominated as a candidate for future reprocessing system, which supposes to apply a centrifugal contactor in the extraction process. For the reprocessing plant, the centrifugal contactor has great advantages such as reducing solvent degradation, improving of equipment utilization rate, compact designing of equipment layout and critical safety domination. From these advantages, the centrifugal contactor is crucial equipment in the aqueous reprocessing process. Since 1985, JNC has been developing the centrifugal contactor. The single unit development has been accomplished and basic characteristics such as extraction performance, fluidic performance and remote maintenance performance have been determined. A durability test has been conducted for high longevity, with consideration given to the nitric acid mist and estimation of the equipment lifetime. System test equipment with centrifugal contactors of engineering scale was installed, and uranium test was conducted. Up to now, a standard flow sheet test in the extraction process and mal-operation test assuming the one stage shutdown condition have been performed. (author)

  18. In-line monitoring of effluents from HTGR fuel particle preparation processes using a time-of-flight mass spectrometer

    International Nuclear Information System (INIS)

    Lee, D.A.; Costanzo, D.A.; Stinton, D.P.; Carpenter, J.A.; Rainey, W.T. Jr.; Canada, D.C.; Carter, J.A.

    1976-08-01

    The carbonization, conversion, and coating processes in the manufacture of HTGR fuel particles have been studied with the use of a time-of-flight mass spectrometer. Non-condensable effluents from these fluidized-bed processes have been monitored continuously from the beginning to the end of the process. The processes which have been monitored are these: uranium-loaded ion exchange resin carbonization, the carbothermic reduction of UO 2 to UC 2 , buffer and low temperature isotropic pyrocarbon coatings of fuel kernels, SiC coating of the kernels, and high-temperature particle annealing. Changes in concentrations of significant molecules with time and temperature have been useful in the interpretation of reaction mechanisms and optimization of process procedures

  19. Role of the consolidated fuel reprocessing program in the United States Breeder Reactor Program

    International Nuclear Information System (INIS)

    Ballard, W.W.; Burch, W.D.

    1980-01-01

    While present US policy precludes the commercial reprocessing of LWR fuels and the recycle of plutonium, the policy does encompass the need to continue a program to develop the technology for reprocessing breeder fuels. Some questions have again risen this year as to the pace of the entire breeder program, including recycle, and the answers are evolving. This paper and the other companion papers which describe several aspects of the Consolidated Fuel Reprocessing Program take a longer-range perspective on the total program. Whether the program is implemented in the general time frame described is dependent on future government actions dedicated to carrying out a systematic program that would permit breeders to be commercialized early in the next century

  20. Methods and calculations for regional, continental, and global dose assessments from a hypothetical fuel reprocessing facility

    International Nuclear Information System (INIS)

    Schubert, J.F.; Kern, C.D.; Cooper, R.E.; Watts, J.R.

    1978-01-01

    The Savannah River Laboratory (SRL) is coordinating an interlaboratory effort to provide, test, and use state-of-the-art methods for calculating the environmental impact to an offsite population from the normal releases of radionuclides during the routine operation of a fuel-reprocessing plant. Results of this effort are the estimated doses to regional, continental, and global populations. Estimates are based upon operation of a hypothetical reprocessing plant at a site in the southeastern United States. The hypothetical plant will reprocess fuel used at a burn rate of 30 megawatts/metric ton and a burnup of 33,000 megawatt days/metric ton. All fuel will have been cooled for at least 365 days. The plant will have a 10 metric ton/day capacity and an assumed 3000 metric ton/year (82 percent online plant operation) output. Lifetime of the plant is assumed to be 40 years

  1. Removal of actinides from high-level wastes generated in the reprocessing of commercial fuels

    International Nuclear Information System (INIS)

    Bond, W.D.; Leuze, R.E.

    1975-09-01

    Progress is reported on a technical feasibility study of removing the very long-lived actinides (uranium, neptunium, plutonium, americium, and curium) from high-level wastes generated in the commercial reprocessing of spent nuclear fuels. The study was directed primarily at wastes from the reprocessing of light water reactor (LWR) fuels and specifically to developing satisfactory methods for reducing the actinide content of these wastes to values that would make 1000-year-decayed waste comparable in radiological toxicity to natural uranium ore deposits. Although studies are not complete, results thus far indicate the most promising concept for actinide removal includes both improved recovery of actinides in conventional fuel reprocessing and secondary processing of the high-level wastes. Secondary processing will be necessary for the removal of americium and curium and perhaps some residual plutonium. Laboratory-scale studies of separations methods that appear most promising are reported and conceptual flowsheets are discussed. (U.S.)

  2. Experimental determination of the Koo fuel temperature coefficient for an HTGR lattice

    Energy Technology Data Exchange (ETDEWEB)

    Agostini, P.; Benedetti, F.; Brighenti, G.; Chiodi, P. L.; Dell' Oro, P.; Giuliani, C.; Tassan, S.

    1974-10-15

    This paper describes temperature-dependent k-infinity measurements conducted using an assembly of loose HTGR coated particles in the BR-2 reactor by means of null reactivity oscillating method comparing the effect of poisoned and unpoisoned lattices like tests performed in the Physical Constants Test Reactor (PCTR) at Hanford. The RB-2 reactor was the property of the Italian firm AGIP NUCLEARE and operated at the Montecuccolino Center in Bologna.

  3. Evaluation of nuclear fuel reprocessing strategies. 2. LWR fuel storage, recycle economics and plutonium logistics

    International Nuclear Information System (INIS)

    Prince, B.E.; Hadley, S.W.

    1983-01-01

    This is the second of a two-part report intended as a critical review of certain issues involved with closing the Light Water Reactor (LWR) fuel cycle and establishing the basis for future transition to commercial breeder applications. The report is divided into four main sections consisting of (1) a review of the status of the LWR spent fuel management and storage problem; (2) an analysis of the economic incentives for instituting reprocessing and recycle in LWRs; (3) an analysis of the time-dependent aspects of plutonium economic value particularly as related to the LWR-breeder transition; and (4) an analysis of the time-dependent aspects of plutonium requirements and supply relative to this transition

  4. Operation of Nuclear Fuel Based on Reprocessed Uranium for VVER-type Reactors in Competitive Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Troyanov, V.; Molchanov, V.; Tuzov, A. [TVEL Corporation, 49 Kashirskoe shosse, Moscow 115409 (Russian Federation); Semchenkov, Yu.; Lizorkin, M. [RRC ' Kurchatov Institute' (Russian Federation); Vasilchenko, I.; Lushin, V. [OKB ' Gidropress' (Russian Federation)

    2009-06-15

    Current nuclear fuel cycle of Russian nuclear power involves reprocessed low-enriched uranium in nuclear fuel production for some NPP units with VVER-type LWR. This paper discusses design and performance characteristics of commercial nuclear fuel based on natural and reprocessed uranium. It presents the review of results of commercial operation of nuclear fuel based on reprocessed uranium on Russian NPPs-unit No.2 of Kola NPP and unit No.2 of Kalinin NPP. The results of calculation and experimental validation of safe fuel operation including necessary isotope composition conformed to regulation requirements and results of pilot fuel operation are also considered. Meeting the customer requirements the possibility of high burn-up achieving was demonstrated. In addition the paper compares the characteristics of nuclear fuel cycles with maximum length based on reprocessed and natural uranium considering relevant 5% enrichment limitation and necessity of {sup 236}U compensation. The expedience of uranium-235 enrichment increasing over 5% is discussed with the aim to implement longer fuel cycles. (authors)

  5. Status report - expert knowledge of operators in fuel reprocessing plants, enrichment plants and fuel fabrication plants

    International Nuclear Information System (INIS)

    Preuss, W.; Kramer, J.; Wildberg, D.

    1987-01-01

    The necessary qualifications of the responsible personnel and the knowledge required by personnel otherwise employed in nuclear plants are among the requirements for licensing laid down in paragraph 7 of the German Atomic Energy Act. The formal regulations for nuclear power plants are not directly applicable to plants in the fuel cycle because of the differences in the technical processes and the plant and work organisation. The aim of the project was therefore to establish a possible need for regulations for the nuclear plants with respect to the qualification of the personnel, and to determine a starting point for the definition of the required qualifications. An extensive investigation was carried out in the Federal Republic of Germany into: the formal requirements for training; the plant and personnel organisation structures; the tasks carried out by the responsible and otherwise employed personnel; and the state of training. For this purpose plant owners and managers were interviewed and the literature and plant specific documentation (e.g. plant rules) were reviewed. On the basis of literature research, foreign practices were determined and used to make comparative evaluations. The status report is divided into three separate parts for the reprocessing, the uranium enrichment, and the manufacture of the fuel elements. On the basis of the situation for reprocessing plants (particularly that of the WAK) and fuel element manufacturing plants, the development of a common (not uniform) regulation for all the examined plants in the fuel cycle was recommended. The report gives concrete suggestions for the content of the regulations. (orig.) [de

  6. Policy in France regarding the back-end of the fuel cycle reprocessing/recycling route

    International Nuclear Information System (INIS)

    Gloaguen, A.; Lenail, B.

    1991-01-01

    The decision taken in early 1970s to base the French power policy on the use of pressurized water reactors also included the strategy for the back end of the nuclear fuel cycle based on reprocessing, waste conditioning for the final disposal in the most suitable form in terms of safety and plutonium recycling to fast breeder reactors. Twenty years have elapsed, and substantial development and investment have been made. New evidences have emerged especially regarding breeder development, and the initial choice has been proved to be sound. EDF and COGEMA, the French utility and fuel cycle companies, respectively, are working together in order to take the best advantage of past efforts. The good behavior of MOX fuel in EDF reactors and the excellent start of the UP3 reprocessing plant of La Hague, which was completed and commissioned in August, 1990, made EDF and COGEMA extremely confident for future decision. The French choice made in favor of fuel reprocessing the history of fuel reprocessing in France, the policy concerning the back end of nuclear fuel cycle of EDF, and the present consideration and circumstances on this matter are reported. (K.I.)

  7. Methodology for evaluation of alternative technologies applied to nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Selvaduray, G.S.; Goldstein, M.K.; Anderson, R.N.

    1977-07-01

    An analytic methodology has been developed to compare the performance of various nuclear fuel reprocessing techniques for advanced fuel cycle applications including low proliferation risk systems. The need to identify and to compare those processes, which have the versatility to handle the variety of fuel types expected to be in use in the next century, is becoming increasingly imperative. This methodology allows processes in any stage of development to be compared and to assess the effect of changing external conditions on the process

  8. Fuel reprocessing plant - no solution for the economy of the region

    International Nuclear Information System (INIS)

    Elvers, G.

    1986-01-01

    Both for the construction and operation stage, the direct and indirect impact of the fuel reprocessing plant on employment on the whole will be negative. It is not altogether certain either that there will be no adverse effects for the areas of tourism. The top organization of German trade unions (DGB) holds that a different structure-political concept from the one represented by the large-scale project of the fuel reprocessing plant would be more appropriate for the region. Employment in the steel and construction industries must be safeguarded by corresponding programmes, and new employment must be created in small- and medium-size companies. (DG) [de

  9. Remote handling equipment for laboratory research of fuel reprocessing in Nuclear Research Institute at Rez

    International Nuclear Information System (INIS)

    Fidler, J.; Novy, P.; Kyrs, M.

    1985-04-01

    Laboratory installations were developed for two nuclear fuel reprocessing methods, viz., the solvent extraction process and the fluoride volatility process. The apparatus for solvent extraction reprocessing consists of a pneumatically driven rod-chopper, a dissolver, mixer-settler extractors, an automatic fire extinguishing device and other components and it was tested using irradiated uranium. The technological line for the fluoride volatility process consists of a fluorimater, condensers, sorption columns with NaF pellets and a distillation column for the separation of volatile fluorides from UF 6 . The line has not yet been tested using irradiated fuel. Some features of the remote handling equipment of both installations are briefly described. (author)

  10. Fast breeder reactor fuel reprocessing R and D: technological development for a commercial plant

    International Nuclear Information System (INIS)

    Colas, J.; Saudray, D.; Coste, J.A.; Roux, J.P.; Jouan, A.

    1987-01-01

    The technological developments undertaken by the CEA are applied to a plant project of a 50 t/y capacity, having to reprocess in particular the SUPERPHENIX 1 reactor fuel. French experience on fast breeder reactor fuel reprocessing is presented, then the 50 t/y capacity plant project and the research and development installations. The R and D programs are described, concerning: head-end operations, solvent extractions, Pu02 conversion and storage, out-of-specification Pu02 redissolution, fission products solution vitrification, conditioning of stainless steel hulls by melting, development of remote operation equipments, study of corrosion and analytical problems

  11. Costs and the environmental impact of radioactive waste treatment in reprocessing high-temperature gas-cooled reactor fuel

    International Nuclear Information System (INIS)

    Davis, W. Jr.

    1976-01-01

    A cost-benefit analysis and an analysis of the reduction in population dose from the use of different decontamination equipment in the off-gas system of a model plant for processing spent fuel from HTGR type reactors are presented

  12. Waste management considerations in HTGR recycle operations

    International Nuclear Information System (INIS)

    Pence, D.T.; Shefcik, J.J.; Heath, C.A.

    1975-01-01

    Waste management considerations in the recycle of HTGR fuel are different from those encountered in the recycle of LWR fuel. The types of waste associated with HTGR recycle operations are discussed, and treatment methods for some of the wastes are described

  13. Assessment of the insertion of reprocessed fuel spiked with thorium in a PWR core

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Victor F.; Monteiro, Fabiana B.A.; Pereira, Claubia, E-mail: victorfc@fis.grad.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Reprocessed fuel by UREX+ technique and spiked with thorium was inserted in a PWR core and neutronic parameters have been analyzed. Based on the Final Safety Analysis Report (FSAR) of the Angra-2 reactor, the core was modeled and simulated with SCALE6.0 package. The neutronic data evaluation was carried out by the analysis of the effective and infinite multiplication factors, and the fuel evolution during the burnup. The conversion ratio (CR) was also evaluated. The results show that, when inserting reprocessed fuel spiked with thorium, the insertion of burnable poison rods is not necessary, due to the amount of absorber isotopes present in the fuel. Besides, the conversion ratio obtained was greater than the presented by standard UO{sub 2} fuel, indicating the possibility of extending the burnup. (author)

  14. The French R and D programme for fast reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Auchapt, P.; Bourgeois, M.; Calame-Longjean, A.; Miquel, P.; Sauteron, J.

    1979-01-01

    The process employed is the Purex process adapted to the specific case of fast breeder reactor fuels. The results achieved have demonstrated that the aqueous method can be applied to these fuels: nearly ten years of operation in the ATl workshop which reprocesses RAPSODIE fuels, and the good results obtained at the Marcoule pilot facility on large batches of fuel attest to this achievement. The CEA effort continues principally on extrapolation to industrial scale, thanks mainly to experiments conducted on industrial prototypes and to the launching of the TOR project, which will, as of 1984, allow reprocessing of FBR fuels on a significant scale, and which will provide extensive additional resources for R and D activities

  15. A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

    Energy Technology Data Exchange (ETDEWEB)

    Parker, Frank L. [Vanderbilt University (United States)

    2012-07-01

    Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded

  16. STAT, GAPS, STRAIN, DRWDIM: a system of computer codes for analyzing HTGR fuel test element metrology data. User's manual

    Energy Technology Data Exchange (ETDEWEB)

    Saurwein, J.J.

    1977-08-01

    A system of computer codes has been developed to statistically reduce Peach Bottom fuel test element metrology data and to compare the material strains and fuel rod-fuel hole gaps computed from these data with HTGR design code predictions. The codes included in this system are STAT, STRAIN, GAPS, and DRWDIM. STAT statistically evaluates test element metrology data yielding fuel rod, fuel body, and sleeve irradiation-induced strains; fuel rod anisotropy; and additional data characterizing each analyzed fuel element. STRAIN compares test element fuel rod and fuel body irradiation-induced strains computed from metrology data with the corresponding design code predictions. GAPS compares test element fuel rod, fuel hole heat transfer gaps computed from metrology data with the corresponding design code predictions. DRWDIM plots the measured and predicted gaps and strains. Although specifically developed to expedite the analysis of Peach Bottom fuel test elements, this system can be applied, without extensive modification, to the analysis of Fort St. Vrain or other HTGR-type fuel test elements.

  17. Materials management in an internationally safeguarded fuels reprocessing plant

    International Nuclear Information System (INIS)

    Hakkila, E.A.; Baker, A.L.; Cobb, D.D.

    1980-04-01

    The following appendices are included: aqueous reprocessing and conversion technology, reference facilities, process design and operating features relevant to materials accounting, operator's safeguards system structure, design principles of dynamic materials accounting systems, modeling and simulation approach, optimization of measurement control, aspects of international verification problem, security and reliability of materials measurement and accounting system, estimation of in-process inventory in solvent-extraction contactors, conventional measurement techniques, near-real-time measurement techniques, isotopic correlation techniques, instrumentation available to IAEA inspectors, and integration of materials accounting and containment and surveillance

  18. CAD system applications to the nuclear fuel reprocessing facilities

    International Nuclear Information System (INIS)

    Morita, Eiji; Matsumoto, Tadakuni; Shikakura, Sakae; Furuya, Kousei; Sakurai, Shin-ichi.

    1994-01-01

    Effective supporting techniques of design, operation, and maintenance of the reprocessing facility have been developed using the Intergraph CAD system. Two and three dimensional views of the process cells were utilized to rationalize the equipment layout and material handling flows, and to check the piping interference. Interferences of the remote maintenance equipment with the process equipments were also evaluated by the pictures on the CAD display. The newest virtual reality technology will help our future development of the more natural simulation for the remote maintenance operator training. (author)

  19. Materials management in an internationally safeguarded fuels reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, E.A.; Baker, A.L.; Cobb, D.D.

    1980-04-01

    The following appendices are included: aqueous reprocessing and conversion technology, reference facilities, process design and operating features relevant to materials accounting, operator's safeguards system structure, design principles of dynamic materials accounting systems, modeling and simulation approach, optimization of measurement control, aspects of international verification problem, security and reliability of materials measurement and accounting system, estimation of in-process inventory in solvent-extraction contactors, conventional measurement techniques, near-real-time measurement techniques, isotopic correlation techniques, instrumentation available to IAEA inspectors, and integration of materials accounting and containment and surveillance. (DLC)

  20. The reprocessing-recycling of spent nuclear fuel. Actinides separation - Application to wastes management

    International Nuclear Information System (INIS)

    2008-01-01

    After its use in the reactor, the spent fuel still contains lot of recoverable material for an energetic use (uranium, plutonium), but also fission products and minor actinides which represent the residues of nuclear reactions. The reprocessing-recycling of the spent fuel, as it is performed in France, implies the chemical separation of these materials. The development and the industrial implementation of this separation process represent a major contribution of the French science and technology. The reprocessing-recycling allows a good management of nuclear wastes and a significant saving of fissile materials. With the recent spectacular rise of uranium prices, this process will become indispensable with the development of the next generation of fast neutron reactors. This book takes stock of the present and future variants of the chemical process used for the reprocessing of spent fuels. It describes the researches in progress and presents the stakes and recent results obtained by the CEA. content: the separation of actinides, a key factor for a sustainable nuclear energy; the actinides, a discovery of the 20. century; the radionuclides in nuclear fuels; the aquo ions of actinides; some redox properties of actinides; some complexing properties of actinide cations; general considerations about treatment processes; some characteristics of nuclear fuels in relation with their reprocessing; technical goals and specific constraints of the PUREX process; front-end operations of the PUREX process; separation and purification operations of the PUREX process; elaboration of finite products in the framework of the PUREX process; management and treatment of liquid effluents; solid wastes of the PUREX process; towards a joint management of uranium and plutonium: the COEX TM process; technical options of treatment and recycling techniques; the fuels of generation IV reactors; front-end treatment processes of advanced fuels; hydrometallurgical processes for future fuel cycles

  1. Commissioning and Operational Experience in Power Reactor Fuel Reprocessing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Pradhan, S., E-mail: spradhan@barctara.gov.in [Tarapur Based Reprocessing Plant, Bhabha Atomic Research Centre, Tarapur (India)

    2014-10-15

    After completing design, construction, commissioning, operation and maintenance experience of the reprocessing plants at Tarapur, Mumbai and Kalpakkam a new reprocessing plant is commissioned and put into operation at BARC, Tarapur since 2011. Subsequent to construction clearance, commissioning of the plant is taken in many steps with simultaneous review by design and safety committees. In spite of vast experience, all the staff was retrained in various aspects of process and utility operations and in operation of innovative changes incorporated in the design. Operating personnel are licensed through an elaborate procedure consisting of various check lists followed by personnel interview. Commissioning systems were divided in sub-systems. Sub-systems were commissioned independently and later integrated testing was carried out. For commissioning, extreme operating conditions were identified in consultation with designers and detailed commissioning procedures were made accordingly. Commissioning was done in different conditions to ensure safety, smooth operation and maintainability. Few modifications were carried out based on commissioning experience. Technical specifications for operation of the plant are made in consultation with designers and reviewed by safety committees. Operation of the plant was carried out after successful commissioning trials with Deep Depleted Uranium (DDU). Emergency operating procedures for each design basis accident were made. Performance of various systems, subsystems are quite satisfactory and the plant has given very good capacity factor. (author)

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

    International Nuclear Information System (INIS)

    Monteiro, Fabiana B.A.; Castro, Victor F.; Faria, Rochkhudson B. de; Pereira, Claubia; Fortini, Angela

    2015-01-01

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

  3. Study on the abnormal reaction in an evaporator at a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Kida, Takashi; Sugikawa, Susumu; Ohsaki, Hiroshi

    2004-01-01

    The calculation code was constructed in order to evaluate a self-accelerated reaction in an evaporator in a fuel reprocessing plant due to organic-nitric acid reactions. This report describes the model of the calculation code and the result of the trial calculation. (author)

  4. Radioactive waste management: a series of bibliographies. Nuclear fuel cycle: reprocessing. Supplement 1

    International Nuclear Information System (INIS)

    McLaren, L.H.

    1984-09-01

    This bibliography contains information on spent fuel reprocessing included in the Department of Energy's Energy Data Base from December 1982 through December 1983. The 555 citations in this bibliography are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number

  5. Nondestructive measurement of spent fuel assemblies at the Tokai Reprocessing and Storage Facility

    International Nuclear Information System (INIS)

    Phillips, J.R.; Bosler, G.E.; Halbig, J.K.; Lee, D.M.

    1979-12-01

    Nondestructive verification of irradiated fuel assemblies is an integral part of any safeguards system for a reprocessing facility. Available techniques are discussed with respect to the level of verification provided by each. A combination of high-resolution gamma spectrometry, neutron detectors, and gross gamma activity profile monitors provide a maximum amount of information in a minimum amount of time

  6. Mechanism of 232U production in MTR fuel evolution of activity in reprocessed uranium

    International Nuclear Information System (INIS)

    Harbonnier, G.; Lelievre, B.; Fanjas, Y.; Naccache, S.J.P.

    1993-01-01

    The use of reprocessed uranium for research reactor fuel fabrication implies to keep operators safe from the hard gamma rays emitted by 232 U daughter products. CERCA has carried out, with the help of French CEA and COGEMA, a detailed study to determine the evolution of the radiation dose rate associated with the use of this material. (author)

  7. Fuel reprocessing plant: No qualitative differences as compared to other sensitive process plants

    International Nuclear Information System (INIS)

    Schweinoch, J.

    1986-01-01

    Nuclear power plants like the fuel reprocessing plant belong to the highly sensitive installations in respect of safety, but involve the same risks qualitatively as liquid-gas plants or chemical plants. Therefore no consequences for basic rights are discernible. The police can take adequate preventive measures. The regulations governing police action provide proper and sufficient warrants. (DG) [de

  8. Trends for minimization of radioactive waste arising from spent nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Polyakov, A.S.; Koltunov, V.S.; Marchenko, V.I.; Ilozhev, A.P.; Mukhin, I.V.

    2000-01-01

    Research and development of technologies for radioactive waste (RAW) minimization arising from spent nuclear fuel reprocessing are discussed. Novel reductants of Pu and Np ions, reagents of purification recycled extractant, possibility of the electrochemical methods are studied. The partitioning of high activity level waste are considered. Examples of microbiological methods decomposition of radioactive waste presented. (authors)

  9. Reprocessing of spent nuclear fuels. Status and trends; Upparbetning av anvaent kaernbraensle. Laege och trender

    Energy Technology Data Exchange (ETDEWEB)

    Hultgren, Aa

    1993-01-01

    The report gives a short review of the status for industrial reprocessing and recycling of Uranium/Plutonium. The following countries are covered: Belgium, France, Germany, Great Britain, India, Japan, Russia, USA. Different fuel cycle strategies are accounted for, and new developments outlined. 116 refs, 27 figs, 12 tabs.

  10. Corrosion resistance of metallic materials for use in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Legry, J.P.; Pelras, M.; Turluer, G.

    1989-01-01

    This paper reviews the corrosion resistance properties required from metallic materials to be used in the various developments of the PUREX process for nuclear fuel reprocessing. Stainless steels, zirconium or titanium base alloys are considered for the various plant components, where nitric acid is the main electrolyte with differing acid and nitrate concentrations, temperature and oxidizing species. (author)

  11. Evaluation of methods for seismic analysis of nuclear fuel reprocessing plants, part 1

    International Nuclear Information System (INIS)

    Tokarz, F.J.; Murray, R.C.; Arthur, D.F.; Feng, W.W.; Wight, L.H.; Zaslawsky, M.

    1975-01-01

    Currently, no guidelines exist for choosing methods of structural analysis to evaluate the seismic hazard of nuclear fuel reprocessing plants. This study examines available methods and their applicability to fuel reprocessing plant structures. The results of this study should provide a basis for establishing guidelines recommending methods of seismic analysis for evaluating future fuel reprocessing plants. The approach taken is: (1) to identify critical plant structures and place them in four categories (structures at or near grade; deeply embedded structures; fully buried structures; equipment/vessels/attachments/piping), (2) to select a representative structure in each of the first three categories and perform static and dynamic analysis on each, and (3) to evaluate and recommend method(s) of analysis for structures within each category. The Barnwell Nuclear Fuel Plant is selected as representative of future commercial reprocessing plants. The effect of site characteristics on the structural response is also examined. The response spectra method of analysis combined with the finite element model for each category is recommended. For structures founded near or at grade, the lumped mass model could also be used. If a time history response is required, a time-history analysis is necessary. (U.S.)

  12. Why reprocess

    International Nuclear Information System (INIS)

    Greenwood, T.

    1977-01-01

    Prospective costs of reprocessing, waste management, and mixed oxide fuel fabrication have risen so much that the costs of U/P recycle and of spent fuel storage are nearly equal. This paper reviews the current state of the reprocessing industry, with a list of facilities all over the world, and examines the incentives and disincentives other than short-term economics that will affect the decision of states to acquire their own reprocessing facilities. Finally, it examines the possibility of avoiding a widespread commercial reprocessing industry

  13. 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)

  14. Environmental survey of the reprocessing and waste management portions of the LWR fuel cycle: a task force report

    International Nuclear Information System (INIS)

    Bishop, W.P.; Miraglia, F.J. Jr.

    1976-10-01

    This Supplement deals with the reprocessing and waste management portions of the nuclear fuel cycle for uranium-fueled reactors. The scope of the report is limited to the illumination of fuel reprocessing and waste management activities, and examination of the environmental impacts caused by these activities on a per-reactor basis. The approach is to select one realistic reprocessing and waste management system and to treat it in enough depth to illuminate the issues involved, the technology available, and the relationships of these to the nuclear fuel cycle in general and its environmental impacts

  15. Environmental survey of the reprocessing and waste management portions of the LWR fuel cycle: a task force report

    Energy Technology Data Exchange (ETDEWEB)

    Bishop, W.P.; Miraglia, F.J. Jr. (eds.)

    1976-10-01

    This Supplement deals with the reprocessing and waste management portions of the nuclear fuel cycle for uranium-fueled reactors. The scope of the report is limited to the illumination of fuel reprocessing and waste management activities, and examination of the environmental impacts caused by these activities on a per-reactor basis. The approach is to select one realistic reprocessing and waste management system and to treat it in enough depth to illuminate the issues involved, the technology available, and the relationships of these to the nuclear fuel cycle in general and its environmental impacts.

  16. Decommissioning alternatives for the West Valley, New York, Fuel Reprocessing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Munson, L F; Nemec, J F; Koochi, A K

    1978-06-01

    The methodology and numerical values of NUREG-0278 were applied to four decommissioning alternatives for the West Valley Fuel Reprocessing Plant. The cost and impacts of the following four alternatives for the process building, fuel receiving and storage, waste tank farm, and auxiliary facilities were assessed: (1) layaway, (2) protective storage, (3) preparation for alternate nuclear use, and (4) dismantlement. The estimated costs are 5.7, 11, 19, and 31 million dollars, respectively. (DLC)

  17. THE ECONOMICS OF REPROCESSING vs. DIRECT DISPOSAL OF SPENT NUCLEAR FUEL

    International Nuclear Information System (INIS)

    Bunn, Matthew; Fetter, Steve; Holdren, John P.; Zwaan, Bob van der

    2003-01-01

    This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recycling to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices

  18. BNFL Sellafield assessment of public radiation exposure due to liquid effluents from fuel reprocessing

    International Nuclear Information System (INIS)

    Hunt, G.J.

    1982-01-01

    Individual (critical group) doses resulting from liquid discharges from the British Nuclear Fuels Limited (BNFL) Sellafield Works have been derived in a form normalised to unit radionuclide discharge rates. This has been done for the purpose of providing a basis for predicting doses in the event of nuclear fuel from a future Sizewell 'B' power station being reprocessed. These doses would have to be reviewed in the light of prevailing circumstances at the time when the actual discharges are known. (author)

  19. Decommissioning alternatives for the West Valley, New York, Fuel Reprocessing Plant

    International Nuclear Information System (INIS)

    Munson, L.F.; Nemec, J.F.; Koochi, A.K.

    1978-06-01

    The methodology and numerical values of NUREG-0278 were applied to four decommissioning alternatives for the West Valley Fuel Reprocessing Plant. The cost and impacts of the following four alternatives for the process building, fuel receiving and storage, waste tank farm, and auxiliary facilities were assessed: (1) layaway, (2) protective storage, (3) preparation for alternate nuclear use, and (4) dismantlement. The estimated costs are 5.7, 11, 19, and 31 million dollars, respectively

  20. Determination of overall decontamination factors for common impurity elements in PHWR spent fuel reprocessing

    International Nuclear Information System (INIS)

    Pant, D.K.; Bhalerao, B.A.; Gupta, K.K.; Kulkarni, P.G.; Gurba, P.B.; Janardan, P.; Changrani, R.D.; Dey, P.K.

    2009-01-01

    An attempt has been made to determine overall decontamination factors for elemental impurities normally encountered in the U 3 O 8 product obtained by reprocessing of PHWR spent fuel. The solution obtained by dissolution of spent fuel and corresponding U 3 O 8 product were analyzed for 24 elemental impurities by ICP-AES for this purpose. Decontamination factors achieved for major neutron poisons are in the range of 200-400. (author)

  1. Legal problems connected with irradiated fuel reprocessing and its waste storage

    International Nuclear Information System (INIS)

    Nercy, B. de.

    1981-10-01

    In view of its nature, an irradiated nuclear fuel reprocessing operation -and the contracts implementing it between the reprocessor and the customer- raises certain difficult legal problems. This paper analyses this question from the legal viewpoint, in particular as regards nuclear fuel and material ownership and products or waste arising therefrom, as well as in the context of rules of international trade and non-proliferation standards. (NEA) [fr

  2. Environmental control aspects for fabrication, reprocessing and waste disposal of alternative LWR and LMFBR fuels

    International Nuclear Information System (INIS)

    Nolan, A.M.; Lewallen, M.A.; McNair, G.W.

    1979-11-01

    Environmental control aspects of alternative fuel cycles have been analyzed by evaluating fabrication, reprocessing, and waste disposal operations. Various indices have been used to assess potential environmental control requirements. For the fabrication and reprocessing operations, 50-year dose commitments were used. Waste disposal was evaluated by comparing projected nuclide concentrations in ground water at various time periods with maximum permissible concentrations (MPCs). Three different fabrication plants were analyzed: a fuel fabrication plant (FFP) to produce low-activity uranium and uranium-thorium fuel rods; a plutonium fuel refabrication plant (PFRFP) to produce plutonium-uranium and plutonium-thorium fuel rods; and a uranium fuel refabrication plant (UFRFP) to produce fuel rods containing the high-activity isotopes 232 U and 233 U. Each plant's dose commitments are discussed separately. Source terms for the analysis of effluents from the fuel reprocessing plant (FRP) were calculated using the fuel burnup codes LEOPARD, CINDER and ORIGEN. Effluent quantities are estimated for each fuel type. Bedded salt was chosen for the waste repository analysis. The repository site is modeled on the Waste Isolation Pilot Program site in New Mexico. Wastes assumed to be stored in the repository include high-level vitrified waste from the FRP, packaged fuel residue from the FRP, and transuranic (TRU) contaminated wastes from the FFP, PFRFP, and UFRFP. The potential environmental significance was determined by estimating the ground-water concentrations of the various nuclides over a time span of a million years. The MPC for each nuclide was used along with the estimated ground-water concentration to generate a biohazard index for the comparison among fuel compositions

  3. A survey of methods to immobilize tritium and carbon-14 arising from a nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Taylor, P.

    1991-02-01

    This report reviews the literature on methods to separate and immobilize tritium ( 3 H) and carbon-14 ( 14 C) released from U0 2 fuel in a nuclear fuel reprocessing plant. It was prepared as part of a broader review of fuel reprocessing waste management methods that might find future application in Canada. The calculated inventories of both 3 H and 14 C in used fuel are low; special measures to limit releases of these radionuclides from reprocessing plants are not currently in place, and may not be necessary in future. If required, however, several possible approaches to the concentration and immobilization of both radionuclides are available for development. Technology to control these radionuclides in reactor process streams is in general more highly developed than for reprocessing plant effluent, and some control methods may be adaptable to reprocessing applications

  4. Safety demonstration test on solvent fire in fuel reprocessing plant

    International Nuclear Information System (INIS)

    Nishio, Gunji; Hashimoto, Kazuichiro

    1989-03-01

    This report summarizes a fundamental of results obtained in the Reprocessing Plant Safety Demonstration Test Program which was performed under the contract between the Science and Technology Agency of Japan and the Japan Atomic Energy Research Institute. In this test program, a solvent fire was hypothesized, and such data were obtained as fire behavior, smoke behavior and integrity of exhaust filters in the ventilation system. Through the test results, it was confirmed that under the fire condition in hypothetical accident, the integrity of the cell and the cell ventilation system were maintained, and the safety function of the exhaust filters was maintained against the smoke loading. Analytical results by EVENT code agreed well with the present test data on the thermofluid flow in a cell ventilation system. (author)

  5. Improved measurement of aluminum in irradiated fuel reprocessed at the Savannah River Site

    International Nuclear Information System (INIS)

    Maxwell, S.L. III.

    1991-01-01

    At the Savannah River Site (SRS), irradiated fuel from research reactor operators or their contract fuel service companies is reprocessed in the H-Canyon Separations Facility. Final processing costs are based on analytical measurements of the amount of total metal dissolved. Shipper estimates for uranium and uranium-235 and measured values at SRS have historically agreed very well. There have occasionally been significant differences between shipper estimates for aluminum and the aluminum content determined at SRS. To minimize analytical error that might contribute to poor shipper-receiver agreement for the reprocessing of off-site fuel, a new analytical method to measure aluminum was developed by SRS Analytical Laboratories at the Central Laboratory Facilities. An EDTA (ethylenediaminetetraacetic acid) titration method, subject to dissolver matrix interferences, was previously used at SRS to measure aluminum in H-Canyon dissolver during the reprocessing of offsite fuel. The new method combines rapid ion exchange technology with direct current argon plasma spectrometry to enhance the reliability of aluminum measurements for off-site fuel. The technique rapidly removes spectral interferences such as uranium and significantly lowers gamma levels due to fission products. Aluminium is separated quantitatively by using an anion exchange technique that employs oxalate complexing, small particle size resin and rapid flow rates. The new method, which has eliminated matrix interference problems with these analyses and improved the quality of aluminum measurements, has improved the overall agreement between shipper-receiver values for offsite fuel processed SRS

  6. Reduction on high level radioactive waste volume and geological repository footprint with high burn-up and high thermal efficiency of HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Yuji, E-mail: fukaya.yuji@jaea.go.jp; Nishihara, Tetsuo

    2016-10-15

    Highlights: • We evaluate the number of canisters and its footprint for HTGR. • We proposed new waste loading method for direct disposal of HTGR. • HTGR can significantly reduce HLW volume compared with LWR. - Abstract: Reduction on volume of High Level radioactive Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and that has particular features such as significantly high burn-up of approximately 120 GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing. By applying the feature, effective waste loading method for direct disposal is proposed in this study. By taking into account these feature, the number of HLW canister generations and its repository footprint are evaluated by burn-up fuel composition, thermal calculation and criticality calculation in repository. As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with Light Water Reactor (LWR) representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. But, the reduced ratios change to 20% and 50% if the long term durability of LWR canister is guaranteed. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.

  7. Work on fuel reprocessing at the Boris Kidric Institute of Nuclear Sciences at Vinca, Yugoslavia

    International Nuclear Information System (INIS)

    Pavasovic, V.

    1969-01-01

    Activity in the region of fuel reprocessing since 1959 up to now has been reported. During that period all necessary conditions were created to enable successful work in that domain (hot laboratory with all necessary devices was constructed, the corresponding staff was trained, also the connections with other research centers were established dealing with these problems). Among the procedures Purex procedure was selected and laboratory plant was constructed to investigate different variants of this procedure. A pre-project has been made in cooperation with the Norway experts covering semi-industrial reprocessing plant. A device for countercurrent extraction is also under development (author) [sr

  8. Historic American Engineering Record, Idaho National Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex

    Energy Technology Data Exchange (ETDEWEB)

    Susan Stacy; Julie Braun

    2006-12-01

    Just as automobiles need fuel to operate, so do nuclear reactors. When fossil fuels such as gasoline are burned to power an automobile, they are consumed immediately and nearly completely in the process. When the fuel is gone, energy production stops. Nuclear reactors are incapable of achieving this near complete burn-up because as the fuel (uranium) that powers them is burned through the process of nuclear fission, a variety of other elements are also created and become intimately associated with the uranium. Because they absorb neutrons, which energize the fission process, these accumulating fission products eventually poison the fuel by stopping the production of energy from it. The fission products may also damage the structural integrity of the fuel elements. Even though the uranium fuel is still present, sometimes in significant quantities, it is unburnable and will not power a reactor unless it is separated from the neutron-absorbing fission products by a method called fuel reprocessing. Construction of the Fuel Reprocessing Complex at the Chem Plant started in 1950 with the Bechtel Corporation serving as construction contractor and American Cyanamid Company as operating contractor. Although the Foster Wheeler Corporation assumed responsibility for the detailed working design of the overall plant, scientists at Oak Ridge designed all of the equipment that would be employed in the uranium separations process. After three years of construction activity and extensive testing, the plant was ready to handle its first load of irradiated fuel.

  9. Milling Behavior of Matrix Graphite Powders with Different Binder Materials in HTGR Fuel Element Fabrication: I. Variation in Particle Size Distribution

    International Nuclear Information System (INIS)

    Lee, Young Woo; Cho, Moon Sung

    2011-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a matrix graphite powder properly prepared and pressed into a spherical shape or a cylindrical compact finally heat-treated at about 1900 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, overcoating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. In order to develop a fuel compact fabrication technology, it is important to develop a technology to prepare the matrix graphite powder (MGP) with proper characteristics, which has a strong influence on further steps and the material properties of fuel element. In this work, the milling behavior of matrix graphite powder mixture with different binder materials and their contents was investigated by analyzing the change in particle size distribution with different milling time

  10. Process options and projected mass flows for the HTGR refabrication scrap recovery system

    International Nuclear Information System (INIS)

    Tiegs, S.M.

    1979-03-01

    The two major uranium recovery processing options reviewed are (1) internal recovery of the scrap by the refabrication system and (2) transfer to and external recovery of the scrap by the head end of the reprocessing system. Each option was reviewed with respect to equipment requirements, preparatory processing, and material accountability. Because there may be a high cost factor on transfer of scrap fuel material to the reprocessing system for recovery, all of the scrap streams will be recycled internally within the refabrication system, with the exception of reject fuel elements, which will be transferred to the head end of the reprocessing system for uranium recovery. The refabrication facility will be fully remote; thus, simple recovery techniques were selected as the reference processes for scrap recovery. Crushing, burning, and leaching methods will be used to recover uranium from the HTGR refabrication scrap fuel forms, which include particles without silicon carbide coatings, particles with silicon carbide coatings, uncarbonized fuel rods, carbon furnace parts, perchloroethylene distillation bottoms, and analytical sample remnants. Mass flows through the reference scrap recovery system were calculated for the HTGR reference recycle facility operating with the highly enriched uranium fuel cycle. Output per day from the refabrication scrap recovery system is estimated to be 4.02 kg of 2355 U and 10.85 kg of 233 U. Maximum equipment capacities were determined, and future work will be directed toward the development and costing of the scrap recovery system chosen as reference

  11. Issues for Conceptual Design of AFCF and CFTC LWR Spent Fuel Separations Influencing Next-Generation Aqueous Fuel Reprocessing

    International Nuclear Information System (INIS)

    D. Hebditch; R. Henry; M. Goff; K. Pasamehmetoglu; D. Ostby

    2007-01-01

    In 2007, the U.S. Department of Energy (DOE) published the Global Nuclear Energy Partnership (GNEP) strategic plan, which aims to meet US and international energy, safeguards, fuel supply and environmental needs by harnessing national laboratory R and D, deployment by industry and use of international partnerships. Initially, two industry-led commercial scale facilities, an advanced burner reactor (ABR) and a consolidated fuel treatment center (CFTC), and one developmental facility, an advanced fuel cycle facility (AFCF) are proposed. The national laboratories will lead the AFCF to provide an internationally recognized R and D center of excellence for developing transmutation fuels and targets and advancing fuel cycle reprocessing technology using aqueous and pyrochemical methods. The design drivers for AFCF and the CFTC LWR spent fuel separations are expected to impact on and partly reflect those for industry, which is engaging with DOE in studies for CFTC and ABR through the recent GNEP funding opportunity announcement (FOA). The paper summarizes the state-of-the-art of aqueous reprocessing, gives an assessment of engineering drivers for U.S. aqueous processing facilities, examines historic plant capital costs and provides conclusions with a view to influencing design of next-generation fuel reprocessing plants

  12. Plant for retention of 14C in reprocessing plants for LWR fuel elements

    International Nuclear Information System (INIS)

    Braun, H.; Gutowski, H.; Bonka, H.; Gruendler, D.

    1983-01-01

    The 14 C produced from nuclear power plants is actually totally emitted from nuclear power plants and reprocessing plants. Using the radiation protection principles proposed in ICRP 26, 14 C should be retained at heavy water moderated reactors and reprocessing plants due to a cost-benefit analysis. In the frame of a research work to cost-benefit analysis, which was sponsored by the Federal Minister of the Interior, an industrial plant for 14 C retention at reprocessing plants for LWR fuel elements has been planned according to the double alkali process. The double alkali process has been chosen because of the sufficient operation experience in the conventional chemical technique. In order to verify some operational parameters and to gain experiences, a cold test plant was constructed. The experiment results showed that the double alkali process is a technically suitable method with high operation security. Solidifying CaCO 3 with cement gives a product fit for final disposal

  13. The situation of radioactive waste management in the fuel reprocessing facility (for fiscal 1979)

    International Nuclear Information System (INIS)

    1981-01-01

    In the fuel reprocessing facility of Power Reactor and Nuclear Fuel Development Corporation (PNC), the release of radioactive gaseous and liquid wastes was so controlled as not to exceed the set standards. Of the radioactive liquid wastes, concentrated wastes and sludge are stored in tanks. Radioactive solid wastes are suitably stored in containers. The situation of radioactive waste management in the fuel reprocessing facility in fiscal 1979 (from April, 1979, to March, 1980) is presented on the basis of the radiation control report made by PNC. The release of radioactive gaseous and liquid wastes was below the set standards. The following data are given in tables: the released quantity of radioactive gaseous and liquid wastes, the cumulative stored amount of radioactive liquid wastes, the produced quantity and cumulative stored amount of radioactive solid wastes; (for reference) the released quantity of radioactive gaseous and liquid wastes in fiscal 1977, 1978 and 1979. (J.P.N.)

  14. Radioactive waste management in a fuel reprocessing facility in fiscal 1982

    International Nuclear Information System (INIS)

    1984-01-01

    In the fuel reprocessing facility of the Power Reactor and Nuclear Fuel Development Corporation, radioactive gaseous and liquid waste are released not exceeding the respective permissible levels. Radioactive concentrated solutions are stored at the site. Radioactive solid waste are stored appropriately at the site. In fiscal 1982, the released quantities of radioactive gaseous and liquid waste were both below the permissible levels. The results of radioactive waste management in the fuel reprocessing facility in fiscal 1982 are given in the tables: the released quantities of radioactive gaseous and liquid waste, the produced quantities of radioactive solid waste, and the stored quantities of radioactive concentrated solutions and of radioactive solid waste as of the end of fiscal 1982. (Mori, K.)

  15. Radioactive wastes management in fiscal year 1983 in the fuel reprocessing plant

    International Nuclear Information System (INIS)

    1985-01-01

    In the nuclear fuel reprocessing plant of Power Reactor and Nuclear Fuel Development Corporation, the releases of radioactive gaseous and liquid wastes are so managed not to exceed the respective objective release levels. Of the radioactive liquid wastes, the high level concentrated wastes are stored in tanks and the low level wastes are stored in tanks or asphalt solidified. For radioactive solid wastes, high level solid wastes are stored in casks, low level solid wastes and asphalt solids in drums etc. The releases of radioactive gaseous and liquid wastes in the fiscal year 1983 were below the objective release levels. The radioactive wastes management in the fuel reprocessing plant in fiscal year 1983 is given in tables, the released quantities, the stored quantities, etc. (Mori, K.)

  16. Electrochemical Methods for Reprocessing Defective Fuel Elements and for Decontaminating Equipment

    International Nuclear Information System (INIS)

    Mikheykin, S. V.; Rybakov, K. A.; Simonov, V. P.

    2002-01-01

    Reprocessing of fuel elements receives much consideration in nuclear engineering. Chemical and electrochemical methods are used for the purpose. For difficultly soluble materials based on zirconium alloys chemical methods are not suitable. Chemical reprocessing of defective or irradiated fuel elements requires special methods for their decladding because the dissolution of the clad material in nitric acid is either impossible (stainless steel, Zr alloys) or quite slow (aluminium). Fuel elements are cut in air-tight glove-boxes equipped with a dust collector and a feeder for crushed material. Chemical treatment is not free from limitations. For this reason we started a study of the feasibility of electrochemical methods for reprocessing defective and irradiated fuel elements. A simplified electrochemical technology developed makes it possible to recover expensive materials which were earlier wasted or required multi-step treatment. The method and an electrochemical cell are suitable for essentially complete dissolution of any fuel elements, specifically those made of materials which are difficultly soluble by chemical methods

  17. Development of a surveillance robot for dimensional and visual inspection of fuel and reflector elements from the Fort St. Vrain HTGR

    International Nuclear Information System (INIS)

    Wallroth, C.F.; Marsh, N.I.; Miller, C.M.; Saurwein, J.J.; Smith, T.L.

    1979-11-01

    A robotic device has been developed for dimensional and visual inspection of irradiated HTGR core components. The robot consists of a rotary table and a two-finger probe, driven by stepping motors, and four remotely controlled television cameras. Automated operation is accomplished via minicomputer control. A total of 51 irradiated fuel and reflector elements were inspected at a fraction of the time and cost required for conventional methods

  18. Fractional release of short-lived noble gases and iodine from HTGR fuel compact containing a fraction of coated fuel particles with through-coating defects

    International Nuclear Information System (INIS)

    Ogawa, Toru; Fukuda, Kosaku; Kobayashi, Fumiaki; Kikuchi, Teruo; Tobita, Tsutomu; Kashimura, Satoru; Kikuchi, Hironobu; Yamamoto, Katsumune.

    1986-10-01

    Fractional release (R/B) data of short-lived noble gases and iodine from sweep-gas irradiated HTGR fuel compacts were analyzed. Empirical formulas to predict R/B of 88 Kr as a function of temperature and fraction through-coating defects, and to calculate ratios of R/B's of other shortlived gases to that of 88 Kr were proposed. A method to predict R/B of iodine was also proposed. As for 131 I, a fission product of major safety concern, (R/B) I 131 ≅ (R/B) Xe 133 was predicted. Applying those methods, R/B from OGL-1 fuel element (5th and 6th) was predicted to show a good agreement with observation. (author)

  19. NO/sub x/ emissions from Hanford nuclear fuels reprocessing plants

    International Nuclear Information System (INIS)

    Pajunen, A.L.; Dirkes, R.L.

    1978-01-01

    Operation of the existing Hanford nuclear fuel reprocessing facilities will increase the release of nitrogen oxides (NO/sub x/) to the atmosphere over present emission rates. Stack emissions from two reprocessing facilities, one waste storage facility and two coal burning power plants will contain increased concentrations of NO/sub x/. The opacity of the reprocessing facilities' emissions is predicted to periodically exceed the State and local opacity limit of twenty percent. Past measurements failed to detect differences in the ambient air NO/sub x/ concentration with and without reprocessing plant operations. Since the facilities are not presently operating, increases in the non-occupational ambient air NO/sub x/ concentration were predicted from theoretical diffusion models. Based on the calculations, the annual average ambient air NO/sub x/ concentration will increase from the present level of less than 0.004 ppM to less than 0.006 ppM at the Hanford site boundaries. The national standard for the annual mean ambient air NO 2 concentration is 0.05 ppM. Therefore, the non-occupational ambient air NO/sub x/ concentration will not be increased to significant levels by reprocessing operations in the Hanford 200 Areas

  20. Pyrochemical reprocessing of molten salt fast reactor fuel: focus on the reductive extraction step

    Directory of Open Access Journals (Sweden)

    Rodrigues Davide

    2015-12-01

    Full Text Available The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves several chemical steps based on redox and acido-basic properties of the various elements contained in the fuel salt. One challenge is to perform a selective extraction of actinides and lanthanides in spent liquid fuel. Extraction of actinides and lanthanides are successively performed by a reductive extraction in liquid bismuth pool containing metallic lithium as a reductive reagent. The objective of this paper is to give a description of the several steps of the reprocessing retained for the molten salt fast reactor (MSFR concept and to present the initial results obtained for the reductive extraction experiments realized in static conditions by contacting LiF-ThF4-UF4-NdF3 with a lab-made Bi-Li pool and for which extraction efficiencies of 0.7% for neodymium and 14.0% for uranium were measured. It was concluded that in static conditions, the extraction is governed by a kinetic limitation and not by the thermodynamic equilibrium.

  1. Krypton-85 health risk assessment for a nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Mellinger, P.J.; Brackenbush, L.W.; Tanner, J.E.; Gilbert, E.S.

    1984-08-01

    The risks involved in the routine release of 85 Kr from nuclear fuel reprocessing operations to the environment were compared to those resulting from the capture and storage of 85 Kr. Instead of releasing the 85 Kr to the environment when fuel is reprocessed, it can be captured, immobilized and stored. Two alternative methods of capturing 85 Kr (cryogenic distillation and fluorocarbon absorption) and one method of immobilizing the captured gas (ion implantation/sputtering) were theoretically incorporated into a representative fuel reprocessing plant, the Barnwell Nuclear Fuel Plant, even though there are no known plans to start up this facility. Given the uncertainties in the models used to generate lifetime risk numbers (0.02 to 0.027 radiation induced fatal cancers expected in the occupational workforce and 0.017 fatal cancers in the general population), the differences in total risks for the three situations, (i.e., no-capture and two-capture alternatives) cannot be considered meaningful. It is possible that no risks would occur from any of the three situations. There is certainly no reason to conclude that risks from 85 Kr routinely released to the environment are greater than those that would result from the other two situations considered. Present regulations mandate recovery and disposal of 85 Kr from the off gases of a facility reprocessing spent fuel from commercial sources. Because of the lack of a clear-cut indication that recovery woud be beneficial, it does not seem prudent to burden the facilities with a requirement for 85 Kr recovery, at least until operating experience demonstrates the incentive. The probable high aging of the early fuel to be processed and the higher dose resulting from the release of the unregulated 3 H and 14 C also encourage delaying implementation of the 85 Kr recovery in the early plants

  2. Use of fuel reprocessing plant instrumentation for international safeguards

    International Nuclear Information System (INIS)

    Ayers, A.L.

    1977-01-01

    The International Atomic Energy Agency has a program for developing instrumentation to be used by safeguards inspectors at reprocessing facilities. These instruments have generally been individual pieces of equipment for improving the accuracy of existing measurement instrumentation or equipment to perform nondestructive assay on a selected basis. It is proposed that greater use be made of redundant plant instrumentation and data recovery systems that could augment plant instrumentation to verify the validity of plant measurements. Use of these methods for verfication must be proven as part of an operating plant before they can be relied upon for safeguards surveillance. Inspectors must be qualified in plant operations, or have ready access to those so qualified, if the integrity of the operation is to be properly assessed. There is an immediate need for the development and in-plant proof testing of an integrated gamma, passive neutron, and active neutron measurement system for drum quantities of radioactive trash. The primary safeguards effort should be limited to plutonium and highly enriched uranium

  3. General Atomic Reprocessing Pilot Plant: engineering-scale dissolution system description

    International Nuclear Information System (INIS)

    Yip, H.H.

    1979-04-01

    In February 1978, a dissolver-centrifuge system was added to the cold reprocessing pilot plant at General Atomic Company, which completed the installation of an HTGR fuel head-end reprocessing pilot plant. This report describes the engineering-scale equipment in the pilot plant and summarizes the design features derived from development work performed in the last few years. The dissolver operating cycles for both thorium containing BISO and uranium containinng WAR fissile fuels are included. A continuous vertical centrifuge is used to clarify the resultant dissolver product solution. Process instrumentation and controls for the system reflect design philosophy suitable for remote operation

  4. Development of new decladding system in the reprocessing process for FBR fuel

    International Nuclear Information System (INIS)

    Yamada, Seiya; Washiya, Tadahiro; Takeuchi, Masayuki; Koizumi, Tsutomu; Aose, Shinichi

    2005-01-01

    As a part of the feasibility study on commercialized fast reactor cycle systems, Japan Nuclear Cycle Development Institute (JNC) has been developing the fuel decladding technology for the dry reprocessing process (oxide electrowinning process) and aqueous reprocessing process. Particularly, in the oxide electrowinning process, the spent fuel should be reduced to powder for quick dissolution in the molten salt at electrolyzer. Therefore, JNC proposes new decladding system with innovative mechanical decladding devices. The decladding system consists of fuel crushing stage, hull separation stage and hull rinsing stage. In the fuel crushing stage, disassembled spent fuel pins are crushed and powdered by mechanical decladding device, then the following stage, the hull and the fuel powder are separated by magnetic separator. Only the fuel powder is fed to the electrolyzer. On the other side, the separated hull is melted by induction heating method, and the small amount of oxide included in the hull fragments is recovered at the hull rinsing stage. The recovered oxide fuel is fed back to the electrolyzer. In this paper, the basic performance of the element equipment that composes this new decladding system will be described. (author)

  5. Workshop on instrumentation and analyses for a nuclear fuel reprocessing hot pilot plant

    International Nuclear Information System (INIS)

    Babcock, S.M.; Feldman, M.J.; Wymer, R.G.; Hoffman, D.

    1980-05-01

    In order to assist in the study of instrumentation and analytical needs for reprocessing plants, a workshop addressing these needs was held at Oak Ridge National Laboratory from May 5 to 7, 1980. The purpose of the workshop was to incorporate the knowledge of chemistry and of advanced measurement techniques held by the nuclear and radiochemical community into ideas for improved and new plant designs for both process control and inventory and safeguards measurements. The workshop was athended by experts in nuclear and radiochemistry, in fuel recycle plant design, and in instrumentation and analysis. ORNL was a particularly appropriate place to hold the workshop since the Consolidated Fuel Reprocessing Program (CFRP) is centered there. Requirements for safeguarding the special nuclear materials involved in reprocessing, and for their timely measurement within the process, within the reprocessing facility, and at the facility boundaries are being studied. Because these requirements are becoming more numerous and stringent, attention is also being paid to the analytical requirements for these special nuclear materials and to methods for measuring the physical parameters of the systems containing them. In order to provide a focus for the consideration of the workshop participants, the Hot Experimental Facility (HEF) being designed conceptually by the CFRP was used as a basis for consideration and discussions

  6. Nuclear fuel reprocessing deactivation plan for the Idaho Chemical Processing Plant, Revision 1

    International Nuclear Information System (INIS)

    Patterson, M.W.

    1994-10-01

    The decision was announced on April 28, 1992 to cease all United States Department of Energy (DOE) reprocessing of nuclear fuels. This decision leads to the deactivation of all fuels dissolution, solvent extraction, krypton gas recovery operations, and product denitration at the Idaho Chemical Processing Plant (ICPP). The reprocessing facilities will be converted to a safe and stable shutdown condition awaiting future alternate uses or decontamination and decommissioning (D ampersand D). This ICPP Deactivation Plan includes the scope of work, schedule, costs, and associated staffing levels necessary to achieve a safe and orderly deactivation of reprocessing activities and the Waste Calcining Facility (WCF). Deactivation activities primarily involve shutdown of operating systems and buildings, fissile and hazardous material removal, and related activities. A minimum required level of continued surveillance and maintenance is planned for each facility/process system to ensure necessary environmental, health, and safety margins are maintained and to support ongoing operations for ICPP facilities that are not being deactivated. Management of the ICPP was transferred from Westinghouse Idaho Nuclear Company, Inc. (WINCO) to Lockheed Idaho Technologies Company (LITCO) on October 1, 1994 as part of the INEL consolidated contract. This revision of the deactivation plan (formerly the Nuclear Fuel Reprocessing Phaseout Plan for the ICPP) is being published during the consolidation of the INEL site-wide contract and the information presented here is current as of October 31, 1994. LITCO has adopted the existing plans for the deactivation of ICPP reprocessing facilities and the plans developed under WINCO are still being actively pursued, although the change in management may result in changes which have not yet been identified. Accordingly, the contents of this plan are subject to revision

  7. Reprocessing of fast reactor fuels in the UP2 plant at La Hague

    International Nuclear Information System (INIS)

    Chenevier, F.; Grellard, J.; Wauquier, J.M.

    The installations of the UP2 plant and particularly the geometry of the HAO shop equipment were defined for reprocessing fuels from the ordinary water system. The high fissile substance level of fuels from the fast neutron system necessitated certain modifications to the installations and some operating restrictions so that they could be treated in the existing installation. After reviewing the characteristics of the reference fuel and describing the particular restrictions to be respected for safety-criticality, the choices made with respect to installation modifications and operating restrictions are presented. The observations made during a first treatment campaign confirm the validity of the options chosen [fr

  8. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    International Nuclear Information System (INIS)

    Park, Jee Won; Jeong, C. J.; Yang, M. S.

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs

  9. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Jeong, C. J.; Yang, M. S

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

  10. Evaluation of methods for seismic analysis of nuclear fuel reprocessing and fabrication facilities

    International Nuclear Information System (INIS)

    Arthur, D.F.; Dong, R.G.; Murray, R.C.; Nelson, T.A.; Smith, P.D.; Wight, L.H.

    1978-01-01

    Methods of seismic analysis for critical structures and equipment in nuclear fuel reprocessing plants (NFRPs) and mixed oxide fuel fabrication plants (MOFFPs) are evaluated. The purpose of this series of reports is to provide the NRC with a technical basis for assessing seismic analysis methods and for writing regulatory guides in which methods ensuring the safe design of nuclear fuel cycle facilities are recommended. The present report evaluates methods of analyzing buried pipes and wells, sloshing effects in large pools, earth dams, multiply supported equipment, pile foundations, and soil-structure interactions

  11. Reprocessing method of ceramic nuclear fuels in low-melting nitrate molten salts

    International Nuclear Information System (INIS)

    Brambilla, G.; Caporali, G.; Zambianchi, M.

    1976-01-01

    Ceramic nuclear fuel is reprocessed through a method wherein the fuel is dispersed in a molten eutectic mixture of at least two alkali metal nitrates and heated to a temperature in the range between 200 and 300 0 C. That heated mixture is then subjected to the action of a gaseous stream containing nitric acid vapors, preferably in the presence of a catalyst such as sodium fluoride. Dissolved fuel can then be precipitated out of solution in crystalline form by cooling the solution to a temperature only slightly above the melting point of the bath

  12. The reprocessing of irradiated fuels by halides and their compounds

    International Nuclear Information System (INIS)

    Bourgeois, M.; Faugeras, P.

    1964-01-01

    A brief description is given of the experiments leading to the choice of the process volatilization of fluorides by gas phase attack. The chemical process is described for certain current types of clad Fuels: the aluminium or the zirconium cladding is first volatilized as chloride by attack with gaseous hydrogen chloride. The uranium is then transformed into volatile hexafluoride by attack with fluorine. These reactions are carried out consecutively in the same reactor in the presence of a fluidized bed of alumina which facilitates heat exchange. The experiments have been carried out in quantities from 100 gms to several kilograms of fuel, first without activity, and then with tracers. A description is given of the laboratory research which was carried out simultaneously on the separation of uranium and plutonium fluorides. Finally, an apparatus is described which is intended to test the process on irradiated fuel at an activity level of several thousands of curies of fission products. (authors) [fr

  13. Determination of end-of-life-failure fractions of HTGR-fuel particles by postirradiation annealing and beta autoradiography

    International Nuclear Information System (INIS)

    Thiele, B.A.; Herren, M.

    1978-11-01

    Fission-product contamination of the helium coolant of High-Temperature Gas-Cooled Reactors (HTGR) is strongly influenced by the end-of-life (EOL) failed-particle fraction. Knowledge of the EOL-failure fraction is the basis for model calculations to predict the total fission product release from the reactor core. After disintegration of irradiation fuel rods, fuel particles are placed in individual holes of a graphite tray. During a 5-h heat treatment at 1000 0 C in a helium atmosphere failed particles leak fission products, especially the volatile cesium, into the graphite. After unloading a β-autoradiograph of the tray is made. Holes that housed defective particles are identified from black spots on the β-sensitive film. The EOL-failure fraction is the ratio of defective particles to the total number of particles tested. The technique is called PIAA, PostIrradiation Annealing and Autoradiography. The PIAA technique was applied to particles of a Trisocoated highly-enriched UO 2 fissile batch irradiated to a burnup of 35% FIMA at an irradiation temperature of 1250 0 C. Visual examination showed all particles to be intact. From 11 to 47% of the particles had failed, as determined by PIAA. Further, postirradiation examination showed that localized corrosion of the silicon carbide coating by fission-product rare-earth chlorides had occurred

  14. Estimation of gamma dose rate from hulls and shield design for the hull transport cask of Fuel Reprocessing Plant (FRP)

    International Nuclear Information System (INIS)

    Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.

    2012-01-01

    In Fuel Reprocessing Plant (FRP), un-dissolved clad of fuel pins known as hulls are the major sources of high level solid waste. Safe handling, transport and disposal require the estimation of radioactivity as a consequent of gamma dose rate from hulls in fast reactor fuel reprocessing plant in comparison with thermal reactor fuel. Due to long irradiation time and low cooling of spent fuel, the evolution of activation products 51 Cr, 58 Co, 54 Mn and 59 Fe present as impurities in the fuel clad are the major sources of gamma radiation. Gamma dose rate from hull container with hulls from Fuel Sub Assembly (FSA) and Radial Sub Assembly (RSA) of Fuel Reprocessing Plant (FRP) was estimated in order to design the hull transport cask. Shielding computations were done using point kernel code, IGSHIELD. This paper describes the details of source terms, estimation of dose rate and shielding design of hull transport cask in detail. (author)

  15. Nuclear fuel cycle reprocessing and waste management technology

    International Nuclear Information System (INIS)

    Allardice, R.H.

    1992-01-01

    In this address, the status of global and US nuclear fuel cycles is briefly reviewed. Projections for Europe and the Pacific basin include a transition towards mixed uranium and plutonium oxide (MOX) recycle in thermal and, eventually, fast reactors. Major environmental benefits could be expected by the development of fast reactor technology. Published estimates of the principal greenhouse gas emission from nuclear operations are reviewed. The final section notes the reduction in radiation dose uptake by operators and general public which can be anticipated when fast reactor and thermal reactor fuel cycles are compared. The major reduction follows elimination of the uranium mining/milling operation

  16. A view from the nuclear fuel reprocessing industry

    International Nuclear Information System (INIS)

    Smith, R.; Hartley, G.

    1982-01-01

    Radiological protection in UK nuclear industry is discussed, with special reference to British Nuclear Fuels Ltd. The following aspects are covered: historical introduction, relevant legislation and general principles; radioactive decay processes (fission, fission products, radio-isotopes, ionising radiations, neutrons); risk assessment (historical, biological radiation effects; ICRP recommendations, dose limits); cost effectiveness of protection; plant design principles; examples of containment (shielding, ventilation and contamination control required for various types of radioactive materials, e.g. fission products, plutonium, depleted uranium; fuel rod storage ponds and decanning caves; fission products at dissolution stage; glovebox handling of Pu operations; critical assembly of fissile materials; surface contamination control; monitoring radiation levels). (U.K.)

  17. Handling of spent nuclear fuel and final storage of nitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    The following stages of handling and transport of the fuel on its way to final storage are dealt with in the report. 1) The spent nuclear fuel is stored at the power station or in the central fuel storage facility awaiting reprocessing. 2) The fuel is reprocessed, i.e. uranium, plutonium and waste are separated from each other. Reprocessing does not take place in Sweden. The highlevel waste is vitrified and can be sent back to Sweden in the 1990s. 3) Vitrified waste is stored for about 30 years awaiting deposition in the final repository. 4) The waste is encapsulated in highly durable materials to prevent groundwater from coming into contact with the waste glass while the radioactivity of the waste is still high. 5) The canisters are emplaced in a final repository which is built at a depth of 500 m in rock of low permeability. 6) All tunnels and shafts are filled with a mixture of clay and sand of low permeability. A detailed analysis of possible harmful effects resulting from normal acitivties and from conceivable accidents is presented in a special section. (author)

  18. Reprocessing of nuclear fuels at the Savannah River Plant

    International Nuclear Information System (INIS)

    Gray, L.W.

    1986-01-01

    For more than 30 years, the Savannah River Plant (SRP) has been a major supplier of nuclear materials such as plutonium-239 and tritium-3 for nuclear and thermonuclear weapons, plutonium-238 for space exploration, and isotopes of americium, curium, and californium for use in the nuclear research community. SRP is a complete nuclear park, providing most of the processes in the nuclear fuel cycle. Key processes involve fabrication and cladding of the nuclear fuel, target, and control assemblies; rework of heavy water for use as reactor moderator; reactor loading, operation, and unloading; chemical recovery of the reactor transmutation products and spent fuels; and management of the gaseous, liquid, and solid nuclear and chemical wastes; plus a host of support operations. The site's history and the key processes from fabrication of reactor fuels and targets to finishing of virgin plutonium for use in the nuclear weapons complex are reviewed. Emphasis has been given to the chemistry of the recovery and purification of weapons grade plutonium from irradiated reactor targets

  19. Analysis of triso packing fraction and fissile material to DB-MHR using LWR reprocessed fuel

    International Nuclear Information System (INIS)

    Silva, Clarysson A.M. da; Pereira, Claubia; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Gual, Maritza R.

    2013-01-01

    Gas-cooled and graphite-moderated reactor is being considered the next generation of nuclear power plants because of its characteristic to operate with reprocessed fuel. The typical fuel element consists of a hexagonal block with coolant and fuel channels. The fuel pin is manufactured into compacted ceramic-coated particles (TRISO) which are used to achieve both a high burnup and a high degree of passive safety. This work uses the MCNPX 2.6.0 to simulate the active core of Deep Burn Modular Helium Reactor (DB-MHR) employing PWR (Pressurized Water Reactor) reprocessed fuel. However, before a complete study of DB-MHR fuel cycle and recharge, it is necessary to evaluate the neutronic parameters to some values of TRISO Packing Fractions (PF) and Fissile Material (FM). Each PF and FM combination would generate the best behaviour of neutronic parameters. Therefore, this study configures several PF and FM combinations considering the heterogeneity of TRISO layers and lattice. The results present the best combination of PF and FM values according with the more appropriated behaviour of the neutronic parameters during the burnup. In this way, the optimized combination can be used to future works of MHR fuel cycle and recharge. (author)

  20. Study of the chemical behaviour of technetium during irradiated fuels reprocessing

    International Nuclear Information System (INIS)

    Zelverte, A.

    1988-04-01

    This paper deals with the preparation of the lower oxidation states +III +IV and +V of technetium in nitric acid and its behaviour during the reprocessing of nuclear fuels (PUREX process). The first part of this work is a bibliographical study of this element in solution without any strong ligand. By chemical and electrochemical technics, pentavalent, tetravalent and trivalent technetium species, were prepared in nitric acid. The following chemical reactions are studied: - trivalent and tetravalent technetium oxidation by nitrate ion. - hydrazine and tetravalent uranium oxidation catalysed by technetium: in those reactions, we point out unequivocally the prominent part of trivalent and tetravalent technetium, - technetium behaviour towards hydroxylamine. Technetium should not cause any disturbance in the steps where hydroxylamine is employed to destroy nitrous acid and hydrazine replacement by hydroxylamine in uranium-plutonium partition could contribute to a best reprocessing of nuclear fuels [fr

  1. The use of curium neutrons to verify plutonium in spent fuel and reprocessing wastes

    International Nuclear Information System (INIS)

    Miura, N.

    1994-05-01

    For safeguards verification of spent fuel, leached hulls, and reprocessing wastes, it is necessary to determine the plutonium content in these items. We have evaluated the use of passive neutron multiplicity counting to determine the plutonium content directly and also to measure the 240 Pu/ 244 Cm ratio for the indirect verification of the plutonium. Neutron multiplicity counting of the singles, doubles, and triples neutrons has been evaluated for measuring 240 Pu, 244 Cm, and 252 Cf. We have proposed a method to establish the plutonium to curium ratio using the hybrid k-edge densitometer x-ray fluorescence instrument plus a neutron coincidence counter for the reprocessing dissolver solution. This report presents the concepts, experimental results, and error estimates for typical spent fuel applications

  2. Plutonium determination by spectrophotometry of plutonium (VI): control of the nuclear fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Grison, J [Compagnie Generale des Matieres Nucleaires (COGEMA), Centre de la Hague, 50 - Cherbourg (France)

    1980-10-01

    The plutonium (VI) spectrophotometric determination, after AgO oxidation in 3 M nitric acid medium, is used for the running-control of the nuclear fuel reprocessing plant at La Hague. Analytical device used in glove-box or shielded-cell is briefly described. This method is fast, sensitive, unfailing and gives simple effluents. It is applied by day and night shifts, during Light Water Reactor fuel reprocessing campaign, for 0.5 mg/l up to 20 g/l plutonium solutions. Reference solution measurements have a 0.8 to 1.4 % relative standard deviation; duplicate plutonium determinations give a 0.3% relative standard deviation for sample analysis. There is a discrepancy (- 0.3% to - 0.9%) between the spectrophotometric method results and the isotopic dilution analysis.

  3. Plutonium determination by spectrophotometry of plutonium (VI): control of the nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Grison, J.

    1980-01-01

    The plutonium (VI) spectrophotometric determination, after AgO oxidation in 3 M nitric acid medium, is used for the running-control of the nuclear fuel reprocessing plant at La Hague. Analytical device used in glove-box or shielded-cell is briefly described. This method is fast, sensitive, unfailing and gives simple effluents. It is applied by day and night shifts, during Light Water Reactor fuel reprocessing campaign, for 0.5 mg/l up to 20 g/l plutonium solutions. Reference solution measurements have a 0.8 to 1.4 % relative standard deviation; duplicate plutonium determinations give a 0.3% relative standard deviation for sample analysis. There is a discrepancy (- 0.3% to - 0.9%) between the spectrophotometric method results and the isotopic dilution analysis [fr

  4. Available reprocessing and recycling services for research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tozser, Sandor; Marshall, Frances M.; Adelfang, Pablo; Bradley, Edward [International Atomic Energy Agency, Vienna (Austria); Budu, Madalina Elena [SOSNY Research and Development Company, Moscow (Russian Federation); Chiguer, Mustapha [AREVA, Paris La Defense (France)

    2016-03-15

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. In the future inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. The IAEA, based on the experience gained during the decade of international cooperation in supporting the objectives of the HEU take-back programmes, drew up a report presenting available reprocessing and recycling services for RR SNF. This paper gives an overview of the report, which will address all aspects of reprocessing and recycling services for RR SNF.

  5. A numerical simulation of 129I in the atmosphere emitted from nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Nishizawa, Masato; Suzuki, Takashi; Nagai, Haruyasu; Togawa, Orihiko

    2010-01-01

    A global chemical transport model, MOZART-4, is applied to investigate the behavior of 129 I emitted from nuclear fuel reprocessing plants in Europe (Sellafield in the UK and La Hague in France). The result of numerical simulation for more than fifty-year period from the 1950s is validated by comparison with measurements of 129 I around the world and analyzed to clarify the characteristic of the distributions of concentration and deposition of 129 I. The modeled concentrations of 129 I in precipitation in Europe and the United States and inventories in the seawater around Japan and the Gulf of Mexico are in the same order as measurements. the emitted 129 I to the atmosphere is distributed all over the Northern Hemisphere due mainly to the prevailing westerlies and can be an important source of supply of artificial 129 I for the seawater remote from the point source such as a nuclear fuel reprocessing plant. (author)

  6. Report of the IAEA advisory group meeting on LMFBR fuel reprocessing

    International Nuclear Information System (INIS)

    1976-05-01

    A summary of the papers and discussions of the meeting is presented, reviewing the status of development in LMFBR fuel reprocessing and focusing attention on important problem areas. The following topics are discussed: Transport, storage and removal of sodium; decladding and shearing; dissolution; Purex process; fluoride volatility method; off-gas purification; waste disposal. Status reports of national programmes of Belgium, France, Federal Republic of Germany, Italy, Japan, United Kingdom, USSR and USA are included

  7. Performance of an accountability measurement system at an operating fuel reprocessing facility

    International Nuclear Information System (INIS)

    Wade, M.A.; Spraktes, F.W.; Hand, R.L.; Baldwin, J.M.; Filby, E.E.; Lewis, L.C.

    1978-01-01

    The ICPP has been engaged for 25 years in the recovery of uranium from spent reactor fuels. In concert with the reprocessing activity, an accountability measurements system has been operated throughout the history of the ICPP. The structure and functions of the accountability measurements system are presented. Its performance is evaluated in order to illustrate the relation of analytical methodology to the overall measurements system. 6 figures, 5 tables

  8. Method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions

    Science.gov (United States)

    Horwitz, E. Philip; Delphin, Walter H.

    1979-07-24

    A method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions containing these and other values by contacting the waste solution with an extractant of tricaprylmethylammonium nitrate in an inert hydrocarbon diluent which extracts the palladium and technetium values from the waste solution. The palladium and technetium values are recovered from the extractant and from any other coextracted values with a strong nitric acid strip solution.

  9. Savannah River Laboratory data banks for risk assessment of fuel reprocessing plants

    International Nuclear Information System (INIS)

    Durant, W.S.

    1981-10-01

    The Savannah River Laboratory maintains a series of computerized data banks primarily as an aid in probabilistic risk assessment studies in the fuel reprocessing facilities. These include component failure rates, generic incidents, and reports of specific deviations from normal operating conditions. In addition to providing data for probability studies, these banks, have served as a valuable aid in trend analysis, equipment histories, process hazards analysis, consequence assessments, incident audit, process problem solving, and training

  10. Conservatism in effective dose calculations for accident events involving fuel reprocessing waste tanks.

    Science.gov (United States)

    Bevelacqua, J J

    2011-07-01

    Conservatism in the calculation of the effective dose following an airborne release from an accident involving a fuel reprocessing waste tank is examined. Within the regulatory constraints at the Hanford Site, deterministic effective dose calculations are conservative by at least an order of magnitude. Deterministic calculations should be used with caution in reaching decisions associated with required safety systems and mitigation philosophy related to the accidental release of airborne radioactive material to the environment.

  11. The Planning of a Small Pilot Plant for Development Work on Aqueous Reprocessing of Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sjoeborg, T U; Haeffner, E; Hultgren, Aa

    1963-10-15

    A shielded volume (42 m{sup 3}) in the hot laboratory at Kjeller, Norway, has been used for the installation of a small pilot plant intended for studies on nuclear fuel reprocessing. During the first period of operation (1963) a plutonium separation method (the Silex process) developed at AB Atomenergi will be studied. This document is a description of the project during the stage of technical planning and chemical process development.

  12. A global-scale dispersion analysis of iodine-129 from nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Nishizawa, Masato; Suzuki, Takashi; Nagai, Haruyasu; Togawa, Orihiko

    2010-01-01

    A three-dimensional global chemical transport model, MOZART-2, is applied to investigate the global-sale dispersion of Iodine-129 from nuclear fuel reprocessing plants. The concentration and deposition of 129 I obtained by MOZART-2 are dispersed all over the Northern Hemisphere. The emission of 129 I to the atmosphere is thus important in considering the transport of 129 I to remote sites. (author)

  13. The use of neptunium-239 to assess neptunium distribution throughout a nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Mair, M.A.; Savage, D.J.; Prentice, P.C.

    1989-08-01

    A radiometric technique has been devised to use the gamma emission from the neptunium-239 daughter of americium-243 to estimate neptunium distribution in a plant reprocessing irradiated plutonium based fuels. Three trials were undertaken with samples from the Prototype Fast Reactor at Dounreay. The trials have confirmed the previous chemical measurements and the usefulness of this technique to highlight the effect of altered flowsheet conditions. (author)

  14. Back-end of the nuclear fuel cycle. A comparison of the direct disposal and reprocessing options

    International Nuclear Information System (INIS)

    Allan, C.J.; Baumgartner, P.

    1997-01-01

    Based on the need to address public concerns, the need to ensure long-term safety and an ethical concern for future generations, many countries are developing technology to dispose of nuclear fuel waste. The waste substances in used fuel can be disposed of either by directly disposing of the used fuel assemblies themselves, or by disposing of the long-lived waste from fuel reprocessing. The basic thesis of this paper is that the direct disposal of either used fuel or of the long-lived heat-generating and non-heat generating waste that arise from reprocessing is technically and economically feasible and that both options will meet the fundamental objectives of protecting human health and the environment. Decisions about whether, or when, to reprocess used fuel, or about whether to dispose of used fuel directly, are not fundamentally waste management issues. (author)

  15. Safety aspects of reprocessing and plutonium fuel facilities in power reactor and nuclear fuel development corporation

    International Nuclear Information System (INIS)

    Sato, S.; Akutsu, H.; Nakajima, K.; Kono, K.; Muto, T.

    1977-01-01

    PNC completed the construction of the first Japanese reprocessing plant in 1974, and the startup is now under way. The plant will have a capacity of 0.7 metric tons of spent fuel per day. Various safety measures for earthquake, radiation, criticality, fire, explosion and leakage of radioactive materials are provided in the plant. 8,000 Ci of Kr-85 and 50 Ci of H-3 per day will be released from the plant to enviroment. Skin dose is conservatively estimated to be about 30 mrem per year. Liquid waste containing 0.7 Ci per day will be discharged into the sea. Whole body dose is conservatively estimated to be 10 mrem per year. R and D for removal of Kr-85 and reducing radioactivity released into the sea are being carried out. Developmental works for solidification of radioactive liquid waste are also being conducted. Safety control in plutonium handling work for both R and D and fuel fabrication has been successfully conducted without significant abnormal occurrence in these ten years. By ''zero-contamination control policy'', surface contamination and airborne contamination in operation rooms are maintained at the background level in usual operation. The intake of plutonium was found at the maximum about one-hundredths of the MPB. External exposure has been generally controlled below three-tenths rem for three months, by shielding and mechanization of process. The radioactivity concentration of exhaust air and liquid effluent disposal is ensured far below the regulation level. Nuclear material control is maintained by a computer system, and no criticality problem has occurred. The safeguard system and installation has been improved, and is sufficient to satisfy the IAEA regulation

  16. Review of experience with plutonium exposure assessment methodologies at the nuclear fuel reprocessing site of British Nuclear Fuels plc

    International Nuclear Information System (INIS)

    Strong, R.

    1988-01-01

    British Nuclear Fuels plc and its predecessors have provided a complete range of nuclear fuel services to utilities in the UK and elsewhere for more than 30 years. Over 30,000 ton of Magnox and Oxide fuel have been reprocessed at Sellafield. During this time substantial experience has accumulated of methodologies for the assessment of exposure to actinides, mainly isotopes of plutonium. For most of the period monitoring of personnel included assessment of systemic uptake deduced from plutonium-in-urine results. The purpose of the paper is to present some conclusions of contemporary work in this area

  17. Conceptual design of the special nuclear material nondestructive assay and accountability system for the HTGR fuel refabrication pilot plant

    International Nuclear Information System (INIS)

    Jenkins, J.D.; McNeany, S.R.; Rushton, J.E.

    1975-07-01

    The conceptual design of the fissile material assay and accountability system for the HTGR refabrication pilot plant has been established. The primary feature affecting the design is the high, time varying, gamma activity of the process material due to the unavoidable presence of uranium-232. This imposes stringent requirements for remote operation and remote maintainability of system components. At the same time, the remote operation lends itself to implementation of an automated data collection and processing system for real-time accountability. The high time-varying gamma activity of the material also precludes application of a number of techniques presently employed for light-water reactor fuel assay. The techniques selected for application in the refabrication facility are (1) active thermal neutron interrogation with fast-fission or delayed-neutron counting for fuel-rod and small-sample assay, (2) calorimetry for high-level waste assay, and (3) passive gamma scanning for low-level waste assay, and rapid on-line relative rod-loading measurements. The principal nondestructive assay subsystems are identified as (1) on-line devices for 100 percent product fuel rod assay and quality control, (2) a multipurpose device in the sample inspection laboratory for small- sample assay and secondary standards calibration, and (3) equipment for assay of high- and low-uranium content scrap and waste materials. A data processing system, which coordinates data from these subsystems with information from other process control sensors, is included to provide real-time material balance information. (U.S.)

  18. Thoria-based nuclear fuels thermophysical and thermodynamic properties, fabrication, reprocessing, and waste management

    CERN Document Server

    Bharadwaj, S R

    2013-01-01

    This book presents the state of the art on thermophysical and thermochemical properties, fabrication methodologies, irradiation behaviours, fuel reprocessing procedures, and aspects of waste management for oxide fuels in general and for thoria-based fuels in particular. The book covers all the essential features involved in the development of and working with nuclear technology. With the help of key databases, many of which were created by the authors, information is presented in the form of tables, figures, schematic diagrams and flow sheets, and photographs. This information will be useful for scientists and engineers working in the nuclear field, particularly for design and simulation, and for establishing the technology. One special feature is the inclusion of the latest information on thoria-based fuels, especially on the use of thorium in power generation, as it has less proliferation potential for nuclear weapons. Given its natural abundance, thorium offers a future alternative to uranium fuels in nuc...

  19. Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

    International Nuclear Information System (INIS)

    Bond, W.D.; Mailen, J.C.; Michaels, G.E.

    1992-10-01

    The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO 2 fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed

  20. Surveillance system using the CCTV at the fuel transfer pond in the Tokai reprocessing plant

    International Nuclear Information System (INIS)

    Hayakawa, T.; Fukuhara, J.; Ochiai, K.; Ohnishi, T.; Ogata, Y.; Okamoto, H.

    1991-01-01

    The Fuel Transfer Pond (FTP) in the Tokai Reprocessing Plant (TRP) is a strategic point for safeguards. Spent fuels, therefore, in the FTP have been surveyed by the surveillance system using the underwater CCTV. This system was developed through the improvement of devices composed of cameras and VCRs and the provision of tamper resistance function as one of the JASPAS (Japan Support Program for Agency Safeguards) program. The purpose of this program is to realize the continuous surveillance of the slanted tunnel through which the spent fuel on the conveyor is moved from the FTP to the Mechanical Processing Cell (MPC). This paper reports that, when this surveillance system is applied to an inspection device, the following requirements are needed: To have the ability of continuous and unattended surveillance of the spent fuel on the conveyor path from the FTP to the MPC; To have the tamper resistance function for continuous and unattended surveillance of the spent fuel