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

International Nuclear Information System (INIS)

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

1996-12-01

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

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

Chemical reactor for a PUREX reprocessing plant of 200Kg U/day capacity

International Nuclear Information System (INIS)

Oliveria Lopes, M.J. de.

1974-03-01

Dissolution of spent reactor fuels in Purex process is studied. Design of a chemical reactor for PWR elements, 3% enriched uranium dioxide with zircaloy cladding, for a 200Kg/day uranium plant is the main objective. Chop-leach process is employed and 7.5M nitric acid is used. Non-criticality was obtained by safe geometry and checked by spectrum homogeneous calculus and diffusion codes. Fuel cycle is considered and decladding and dissolution are treated more accurately

2-22 Study of Oxidation/reduction Volatilization Technology

Institute of Scientific and Technical Information of China (English)

Tan; Cunmin[1; Cao; Shiwei[1; Tian; Yuan[1; Qin; Zhi[1

2015-01-01

As an advanced dry head-end processing of spent fuel reprocessing, the oxidation-reduction volatilization technology will use for pulverizing uranium oxide ceramic pellets, decladding, and removal of most of volatile and semi-volatile fission elements, 3H, 14C, Kr, Xe, I, Cs, Ru and Tc, from fuel prior to main treatment process. The AIROX and ORIOX process, including circulation of oxidation in oxygen atmosphere and reduction in hydrogen atmosphere, researched on international at present, is considered to be the first choice for head-end processing.

Pyrochemical head-end treatment for fast reactor fuel elements

International Nuclear Information System (INIS)

Avogadro, A.

1978-01-01

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

A Study on the Radioactivity Reduction Method for the Decladding Hull

International Nuclear Information System (INIS)

Kim, Jong Ho; Jung, In Ha; Park, Jang Jin; Shin, Jin Myeong; Lee, Ho Hee; Yang, Myung Seung

1994-01-01

The cladding materials remaining after reprocessing process of the nuclear fuel, generally called as hulls, are classified as a high-level radioactive waste. They are usually packaged in the container for disposal after being compacted, melted, or solidified into the matrix. The efforts to fabricated a better ingot for a more favorable disposal to the environment have failed due to the technical difficulties encountered in the chemical decontamination method. In the early 1990s, the accumulation of radio-chemical data on hulls and the advent of new technology such as a laser or plasma have made the pre-treatment of the hulls mere efficient. This paper summarizes the information regarding the radio-chemical analysis of the hull through a literature survey and determines the characteristics of the hull and depth profile of the radio-nuclides within the hull thickness. The feasibility study was carried out to evaluate the reduction of the radioactivity by peeling off the surface of the hull with the application of laser technology

The Thermox Process

International Nuclear Information System (INIS)

Tjaelldin, O.

1963-09-01

The Thermox process is a process developed by AB Atomenergi for the decladding and dissolution of irradiated Zircaloy-2 clad uranium dioxide fuel elements and consists of the following stages: 1. Decladding by means of thermal oxidation of the Zircaloy-2 with oxygen and water vapour at 825 C using nitrogen as a catalyst. 2. Oxidation of the uranium dioxide pellets with air and oxygen to U 3 O 8 at a temperature of 450 - 650 C. 3. Dissolving and leaching the uranium oxides with dilute nitric acid leaving the insoluble zirconium oxide as a residue. 4. Filtering the solution and washing the residues of the cladding. The work has included the following parts; The laboratory scale investigation of the conditions for the oxidation of Zircaloy-2 in various gas mixtures and of the conditions for oxidizing and dissolving sintered UO 2 pellets; The development on a pilot plant scale of suitable apparatus and process techniques for the safe and reproducible treatment of half length inactive fuel elements; Studies of some special operation and handling problems, which have to be solved before the method can be applied in full scale. Five half length fuel elements have been treated, and the results have been satisfactory. The pilot plant experiments have proved that inactive fuel elements can be decanned, oxidized and dissolved by means of the Thermox process. Solutions and canning residues are easy to filter, separate, and handle and are free from corroding agents. The uranium losses can be kept very low. The zirconium dioxide is obtained in a form suitable for permanent disposal

Analysis of F-Canyon Effluents During the Dissolution Cycle with a Fourier Transform Infrared Spectrometer/Multipath Cell

International Nuclear Information System (INIS)

Villa, E.

1999-01-01

Air samples from F-Canyon effluents were collected at the F-Canyon stack and transported to a laboratory at the Savannah River Technology Center (SRTC) for analysis using a Fourier transform infrared spectrometer in conjunction with a multipath cell. Air samples were collected during the decladding and acid cuts of the dissolution of the irradiated aluminum-cladded slugs. The FTIR analyses of the air samples show the presence of NO2, NO, HNO2, N2O, SF6, and 85Kr during the dissolution cycle. The concentration time profiles of these effluents corresponded with expected release rates from the F-Canyon operations

The Thermox Process

Energy Technology Data Exchange (ETDEWEB)

Tjaelldin, O

1963-09-15

The Thermox process is a process developed by AB Atomenergi for the decladding and dissolution of irradiated Zircaloy-2 clad uranium dioxide fuel elements and consists of the following stages: 1. Decladding by means of thermal oxidation of the Zircaloy-2 with oxygen and water vapour at 825 C using nitrogen as a catalyst. 2. Oxidation of the uranium dioxide pellets with air and oxygen to U{sub 3}O{sub 8} at a temperature of 450 - 650 C. 3. Dissolving and leaching the uranium oxides with dilute nitric acid leaving the insoluble zirconium oxide as a residue. 4. Filtering the solution and washing the residues of the cladding. The work has included the following parts; The laboratory scale investigation of the conditions for the oxidation of Zircaloy-2 in various gas mixtures and of the conditions for oxidizing and dissolving sintered UO{sub 2} pellets; The development on a pilot plant scale of suitable apparatus and process techniques for the safe and reproducible treatment of half length inactive fuel elements; Studies of some special operation and handling problems, which have to be solved before the method can be applied in full scale. Five half length fuel elements have been treated, and the results have been satisfactory. The pilot plant experiments have proved that inactive fuel elements can be decanned, oxidized and dissolved by means of the Thermox process. Solutions and canning residues are easy to filter, separate, and handle and are free from corroding agents. The uranium losses can be kept very low. The zirconium dioxide is obtained in a form suitable for permanent disposal.

Development of nuclear fuel cycle remote handling technology

International Nuclear Information System (INIS)

Kim, K. H.; Park, B. S.; Kim, S. H.

2010-04-01

This report presents the development of remote handling systems and remote equipment for use in the pyprocessing verification at the PRIDE (PyRoprocess Integrated inactive Demonstration facility). There are three areas conducted in this work. In first area, developed were the prototypes of an engineering-scale high-throughput decladding voloxidizer which is capable of separating spent fuel rod-cuts into hulls and powder and collecting them separately and an automatic equipment which is capable of collecting residual powder remaining on separated hulls. In second area, a servo-manipulator prototype was developed to operate and maintain pyroprocess equipment located at the argon cell of the PRIDE in a remote manner. A servo-manipulator with dual arm that is mounted on the lower part of a bridge transporter will be installed on the ceiling of the in-cell and can travel the length of the ceiling. In last area, a simulator was developed to simulate and evaluate the design developments of the pyroprocess equipment from the in-cell arrangements, remote operability and maintainability viewpoint in a virtual process environment in advance before they are constructed. The developed decladding voloxidizer and automatic equipment will be utilized in the development of a head-end process for pyroprocessing. In addition, the developed servo-manipulator will be installed in the PRIDE and used for remote operation and maintenance of the pyroprocess equipment. The developed simulator will be also used to verify and improve the design of the pyroprocess equipment for the PRIDE application. Moreover, these remote technologies described above can be directly used in the PRIDE and applied for the ESPF (Engineering Scale Pyroprocess Facility) and KAPF (Korea Advanced Pyroprocess Facility) development

Proceedings of DUPIC fuel workshop 97

International Nuclear Information System (INIS)

1997-07-01

The researchers discuss the technical aspects of DUPIC fuel fabrication in the workshop as follows; 1) The DUPIC fuel development program in KAERI 2) AECL's progress in developing the DUPIC fuel fabrication process 3) Mechanical decladding 4) Nonproliferation and safeguards aspects of the DUPIC fuel cycle concept 5) Assessment of DUPIC fuel compatibility with CANDU-6 6) The development of combination software for spent PWR fuel to fabricate the homogeneous DUPIC fuel 7) Thermodynamic properties of the DUPIC fuel and its performance 8) Captural properties of cesium and ruthenium 9) A secondary fuel removal process : Plasma processing 10) Technology development for DUPIC process safeguards

Proceedings of DUPIC fuel workshop 97

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-07-01

The researchers discuss the technical aspects of DUPIC fuel fabrication in the workshop as follows; (1) The DUPIC fuel development program in KAERI (2) AECL`s progress in developing the DUPIC fuel fabrication process (3) Mechanical decladding (4) Nonproliferation and safeguards aspects of the DUPIC fuel cycle concept (5) Assessment of DUPIC fuel compatibility with CANDU-6 (6) The development of combination software for spent PWR fuel to fabricate the homogeneous DUPIC fuel (7) Thermodynamic properties of the DUPIC fuel and its performance (8) Captural properties of cesium and ruthenium (9) A secondary fuel removal process : Plasma processing (10) Technology development for DUPIC process safeguards.

Estimation of penetration depth of fission products in cladding Hull

International Nuclear Information System (INIS)

Kim, Hee Moon; Jung, Yang Hong; Yoo, Byong Ok; Choo, Yong Sun; Hong, Kwon Pyo

2005-01-01

A disposal and a reprocessing for spent fuel rod with high burnup need de-cladding procedure. Pellet in this rod has been separated from a cladding hull to reduce a radioactivity of hull by chemical and mechanical methods. But fission products and actinides(U,Pu) still remain inside of cladding hull by chemical bonding and fission spike, which is called as 'contamination'. More specific removal of this contamination would have been considered. In this study, the sorts of fission products and penetration depth in hull were observed by EPMA test. To analyze this behavior, SRIM 2000 code was also used as energies of fission products and an oxide thickness of hull

Development of nuclear fuel cycle remote handling technology

International Nuclear Information System (INIS)

Kim, K. H.; Park, B. S.; Kim, S. H.

2012-04-01

This report presents the development of remote handling systems and remote equipment for use in the pyprocessing verification at the PRIDE (PyRoprocess Integrated inactive Demonstration facility). There are four areas conducted in this work. In first area, the prototypes of an engineering-scale high-throughput decladding voloxidizer which is capable of separating spent fuel rod-cuts into hulls and powder and collecting them separately, and an automatic equipment which is capable of collecting residual powder remaining on separated hulls were developed. In second area, a servo-manipulator system was developed to operate and maintain pyroprocess equipment located at the argon cell of the PRIDE in a remote manner. A servo-manipulator with dual arm that is mounted on the lower part of a bridge transporter will be installed on the ceiling of the in-cell and can travel the length of the ceiling. In third area, a digital mock-up and a remote handling evaluation mock-up were constructed to evaluate the pyroprocess equipments from the in-cell arrangements, remote operability and maintainability viewpoint before they are installed in the PRIDE. In last area, a base technology for remote automation of integrated pyroprocess was developed. The developed decladding voloxidizer and automatic equipment will be utilized in the development of a head-end process for pyroprocessing. In addition, the developed servo-manipulator will be used for remote operation and maintenance of the pyroprocess equipments in the PRIDE. The constructed digital mock-up and remote handling evaluation mock-up will be also used to verify and improve the pyroprocess equipments for the PRIDE application. Moreover, these remote technologies described above can be directly used in the PRIDE and applied for the KAPF (Korea Advanced Pyroprocess Facility) development

A study on manufacturing and quality control technology of DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Seung; Park, H. S.; Lee, Y. W. [and others

1997-09-01

A series of experiments are performed to verify the manufacturability of DUPIC fuel and its performance by use of HANARO test reactor. Major works performed during this research period are : analysis of manufacturing process of DUPIC fuel, fabrication technology development such as development of disassembly and decladding method of spent PWR fuel, study on the OREOX process using simulated high burnup fuel, weldability of end cap weld, and development of fabrication equipment including the conceptual and detailed design of DUPIC equipment mainly for the powder preparation, pelletization and fuel element fabrication. A study on the material properties of DUPIC fuel and performance analysis method using irradiation of test fuel was also performed. (author). 91 refs., 274 tabs., 254 figs.

Chemical dissolution of spent fuel and cladding using complexed fluoride species

International Nuclear Information System (INIS)

Rance, P.J.W.; Freeman, G.A.; Mishin, V.; Issoupov, V.

2001-01-01

The dissolution of LWR fuel cladding using two fluoride ion donors, HBF 4 and K 2 ZrF 6 , in combination with nitric acid has been investigated as a potential reprocessing head-end process suitable for chemical decladding and fuel dissolution in a single process step. Maximum zirconium concentrations in the order of 0,75 to 1 molar have been achieved and dissolution found to continue to low F:Zr ratios albeit at ever decreasing rates. Dissolution rates of un-oxidised zirconium based fuel claddings are fast, whereas oxidised materials exhibit an induction period prior to dissolution. Data is presented relating to the rates of dissolution of cladding and UO 2 fuels under various conditions. (author)

A study on manufacturing and quality control technology of DUPIC fuel

International Nuclear Information System (INIS)

Yang, Myung Seung; Park, H. S.; Lee, Y. W.

1997-09-01

A series of experiments are performed to verify the manufacturability of DUPIC fuel and its performance by use of HANARO test reactor. Major works performed during this research period are : analysis of manufacturing process of DUPIC fuel, fabrication technology development such as development of disassembly and decladding method of spent PWR fuel, study on the OREOX process using simulated high burnup fuel, weldability of end cap weld, and development of fabrication equipment including the conceptual and detailed design of DUPIC equipment mainly for the powder preparation, pelletization and fuel element fabrication. A study on the material properties of DUPIC fuel and performance analysis method using irradiation of test fuel was also performed. (author). 91 refs., 274 tabs., 254 figs

Research reactor spent fuel management in Argentina

International Nuclear Information System (INIS)

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

2002-01-01

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

Dry Refabrication Technology Development of Spent Nuclear Fuel

International Nuclear Information System (INIS)

Lee, Jung Won; Park, G. I.; Park, C. J.

2010-04-01

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

A process to remove ammonia from PUREX plant effluents

International Nuclear Information System (INIS)

Moore, J.D.

1990-01-01

Zirconium-clad nuclear fuel from the Hanford N-Reactor is reprocessed in the PUREX (Plutonium Uranium Extraction) Plant operated by the Westinghouse Hanford Comapny. Before dissolution, cladding is chemically removed from the fuel elements with a solution of ammonium fluoride-ammonium nitrate (AFAN). a solution batch with an ammonia equivalent of about 1,100 kg is added to each fuel batch of 10 metric tons. This paper reports on this decladding process, know as the 'Zirflex' process which produces waste streams containing ammonia and ammonium slats. Waste stream treatment, includes ammonia scrubbing, scrub solution evaporation, residual solids dissolution, and chemical neutralization. These processes produce secondary liquid and gaseous waste streams containing varying concentrations of ammonia and low-level concentrations of radionuclides. Until legislative restrictions were imposed in 1987, these secondary streams were released to the soil in a liquid disposal 'crib' and to the atmosphere

Development on the High-throughput Vol-oxidizer for Decladding and Voloxidation of Spent Fuel Rod-cuts

International Nuclear Information System (INIS)

Kim, Young Hwang; Jung, Jae Hoo; Kim, Ki Ho; Park, Byung Buk; Lee, Hyo Jik; Kim, Sung Hyun; Park, Hee Sung; Lee, Jong Kwang; Kim, Ho Dong

2009-12-01

A high-throughput vol-oxidizer which can handle a several ten kg HM/batch is being developed to supply U 3 O 8 powders to an electrolytic reduction reactor in pyro-processing. At the first year step(2007), for enhancement of oxidation and recovery rate, we analyzed the mechanical and chemical methods, and devised the main mechanism with ball drop methods and rotary kiln type. Also, the main devices for oxidation and recovery of rod-cuts were designed by using the Solid Works and COSMOS program tools, and manufactured after thermal/mechanical analysis. In order to verify the main devices, simulation fuels(W 90%+SiO 2 10%) were manufactured and the main devices were tested for the oxidation and recovery rate of its. Here the expansion ratio of simulation fuel is similar to U 3 O 8 (2.7). At the second year step(2008), with the constant ration of rod-cuts volume and expansion ratio of U 3 O 8 (2.7), we produced a theoretical equation that can estimate the volume of rod-cuts according to a variation of their weight and lengths. We considered various materials such as ceramics and Ni-Cr, finally, the APM material which can constantly maintain against high temperature(1,200 .deg. C) and vacuum(1 torr) was selected and a vol-oxidizer was designed. At the third year step(2009), in order to manufacture a high-throughput vol-oxidizer, we have analyzed the vol-oxidizer for remote operability and maintainability, also the remote assembling and disassembling possibilities of the selected modules have been analyzed in terms of visibility, interference, approach, weight, and so on. We have presented final modular design and manufactured a high-throughput vol-oxidizer. Also, we have conducted the blank, heating(over 500 .deg. C) and hull separation test(capacity : 50 kg HM/batch, hull length 50mm) on the high-throughput vol-oxidizer. Also, these design technologies for the high-throughput vol-oxidizer will be utilized in the development of a more efficient vol-oxidizer with higher capacity for PRIDE and ESPF facilities in the near future

Melting process to condition decladding hulls generated by the reprocessing of LWR and FBR spent fuels

International Nuclear Information System (INIS)

Bonniaud, R.; Jacquet-Francillon, N.; Jouan, A.; Sombret, C.

1981-01-01

The fusion compaction of metallic waste from spent fuel hulls is shown to be easily feasible for both Zircaloy and for stainless steel, and volume reduction factors in the region of 5 to 7, corresponding to the theoretical density of the alloy obtained, are arrived at quite easily. The Zircaloy copper alloy, put into use to lower the fusion point of the Zircaloy, appears extremely interesting both as to the ease with which it can be used and the possibility which it offers of working at temperatures always lower than 1250 0 C. The decreasing of fusion temperature is less spectacular with stainless steel; only the use of silicon enabling the lowering of the temperature to around 1200 0 C appears really feasible. The use of decontaminating agents either during or at the end of the fusion operation seems to be a promising technique, especially in the case of stainless steel where the use of a borosilicated glass is easy. The choice of decontaminating agent is more difficult for Zircaloy which reduces the principal oxide components of glasses and makes necessary the use of molten salts mixtures, the composition of which has not yet been defined. The decontamination factors obtained during the tests run on steel are encouraging although they were obtained using artificially contaminated hulls; they should therefore be considered with precaution and be confirmed by further tests in hot cells using real hulls

Development of the spent fuel rod cutting device using the blade cutters

International Nuclear Information System (INIS)

Jung, Jae Hoo; Yoon, Ji Sup; Hong, Dong Hee; Kim, Young Hwan; Park, Gee Yong; Kim, Do Woo

2000-11-01

A spent fuel rod cutting device is to cut a spent nuclear fuel rod to optimal size for consequent decladding operation. In this paper, various properties of fuel rod, such as a dimension and material of zircaloy tubes and fuel pellets, are investigated. Also, commercially available cutting method and tools is investigated in terms of its performance. As a result, the blade cutter is selected for the design. In order to fabricate the durable blade cutter, various materials are analyzed in terms of material properties, cutter shape, and heat treatment method, etc. Also, the durability of this tool is tested by cutting the SUS tubes and zircaloy tubes. In the device design, the remote maintainability is considered so that the modularized design is accomplished. Also, the other factors considered in the design are the round shape sustainability at the cut surface, the amount of debris generation, and the fire risk, etc. Considering these design consideration, the spent fuel rod cutting device is fabricated and tested

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

Corrosion of clean Mg-Zr alloys in various basic media for waste encapsulation

International Nuclear Information System (INIS)

Lambertin, D.; Frizon, F.; Blachere, A.; Bart, F.

2012-01-01

The dismantling of UNGG nuclear reactor generates a large volume of fuel decladding. These materials are based on Mg-Zr alloy. The strategy could be to encapsulate these wastes into an ordinary Portland cement (OPC) or geo-polymer (aluminosilicate material) in a form suitable for storage. Studies have been performed on Mg or Mg-Al alloy in basic media but no data are available on Mg-Zr behaviour. The influence of representative pore solution of OPC and geo-polymer with Mg-Zr alloy has been studied on corrosion behaviour. Electrochemical methods have been used to determine the corrosion densities at room temperature. Results show that the corrosion densities of Mg-Zr alloy in OPC solution is one order of magnitude more important than in geo-polymer solution environment and effect of inhibiting agent has been undertaken with Mg-Zr alloy. The evaluation of encapsulation of Mg-Zr alloy in OPC and geo-polymer has been done in term of corrosion hydrogen production. (authors)

Mechanistic model for Sr and Ba release from severely damaged fuel

International Nuclear Information System (INIS)

Rest, J.; Cronenberg, A.W.

1985-11-01

Among radionuclides associated with fission product release during severe accidents, the primary ones with health consequences are the volatile species of I, Te, and Cs, and the next most important are Sr, Ba, and Ru. Considerable progress has been made in the mechanistic understanding of I, Cs, Te, and noble gas release; however, no capability presently exists for estimating the release of Sr, Ba, and Ru. This paper presents a description of the primary physical/chemical models recently incorporated into the FASTGRASS-VFP (volatile fission product) code for the estimation of Sr and Ba release. FASTGRASS-VFP release predictions are compared with two data sets: (1) data from out-of-reactor induction-heating experiments on declad low-burnup (1000 and 4000 MWd/t) pellets, and (2) data from the more recent in-reactor PBF Severe Fuel Damage Tests, in which one-meter-long, trace-irradiated (89 MWd/t) and normally irradiated (approx.35,000 MWd/t) fuel rods were tested under accident conditions. 10 refs

The dangers of irradiate uranium in nuclear reactors; Les dangers de l'uranium irradie dans les reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Jammet, H; Joffre, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The danger of the uranium cans sur-activated by the use in the nuclear reactors is triple: - Irradiation from afar, during manipulations of the cans. - Contamination of air when decladding. - Contamination of air by fire of uranium in a reactor in function The first two dangers are usual and can be treated thanks to the rules of security in use in the atomic industry. The third has an accidental character and claimed for the use of special and exceptional rules, overflowing the industrial setting, to reach the surrounding populations. (authors) [French] Le danger des cartouches d'uranium suractive par utilisation dans les reacteurs nucleaires est triple: - Irradiation a distance, lors des manipulations des cartouches. - Contamination de l'air au moment de leur degainage. - Contamination de l'air par incendie d'uranium dans un reacteur en fonctionnement. Les deux premiers dangers sont habituels et peuvent etre traites grace aux regles de securite en usage dans l'industrie atomique. Le troisieme revet un caractere accidentel et reclame l'emploi de regles speciales et exceptionnelles, debordant le cadre industriel, pour atteindre celui des populations environnantes. (auteurs)

LARGE SCALE METHOD FOR THE PRODUCTION AND PURIFICATION OF CURIUM

Science.gov (United States)

Higgins, G.H.; Crane, W.W.T.

1959-05-19

A large-scale process for production and purification of Cm/sup 242/ is described. Aluminum slugs containing Am are irradiated and declad in a NaOH-- NaHO/sub 3/ solution at 85 to 100 deg C. The resulting slurry filtered and washed with NaOH, NH/sub 4/OH, and H/sub 2/O. Recovery of Cm from filtrate and washings is effected by an Fe(OH)/sub 3/ precipitation. The precipitates are then combined and dissolved ln HCl and refractory oxides centrifuged out. These oxides are then fused with Na/sub 2/CO/sub 3/ and dissolved in HCl. The solution is evaporated and LiCl solution added. The Cm, rare earths, and anionic impurities are adsorbed on a strong-base anfon exchange resin. Impurities are eluted with LiCl--HCl solution, rare earths and Cm are eluted by HCl. Other ion exchange steps further purify the Cm. The Cm is then precipitated as fluoride and used in this form or further purified and processed. (T.R.H.)

Design of the PRIDE Facility

International Nuclear Information System (INIS)

You, Gil Sung; Choung, Won Myung; Lee, Eun Pyo; Cho, Il Je; Kwon, Kie Chan; Hong, Dong Hee; Lee, Won Kyung; Ku, Jeong Hoe

2009-01-01

From 2007, KAERI is developing a PyRoprocess Integrated inactive DEmonstration facility (the PRIDE facility). The maximum annual treatment capacity of this facility will be a 10 ton-HM. The process will use a natural uranium feed material or a natural uranium mixed with some surrogate material for a simulation of a spent fuel. KAERI has also another plan to construct a demonstration facility which can treat a real spent fuel by pyroprocessing. This facility is called by ESPF, Engineering Scale Pyroprocess Facility. The ESPF will have the same treatment capability of spent fuel with the PRIDE facility. The only difference between the PRIDE and the ESPF is a radiation shielding capability. From the PRIDE facility designing works and demonstration with a simulated spent fuel after construction, it will be able to obtain the basic facility requirements, remote operability, interrelation properties between process equipment for designing of the ESPF. The flow sheet of the PRIDE processes is composed of five main processes, such as a decladding and voloxidation, an electro-reduction, an electrorefining, an electro-winning, and a salt waste treatment. The final products from the PRIDE facility are a simulated TRU metal and U metal ingot

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)

Development of spent fuel remote handling technology

International Nuclear Information System (INIS)

Yoon, J. S.; Hong, H. D.; Kim, S. H.

2004-02-01

In this research, the remote handling technology is developed for the advanced spent fuel conditioning process which gives a possible solution to deal with the rapidly increasing spent fuels. In detail, a fuel rod slitting device is developed for the decladding of the spent fuel. A series of experiments has been performed to find out the optimal condition of the spent fuel voloxidation which converts the UO 2 pellet into U 3 O 8 powder. The design requirements of the ACP equipment for hot test is established by analysing the modular requirement, radiation hardening and thermal protection of the process equipment, etc. The prototype of the servo manipulator is developed. The manipulator has an excellent performance in terms of the payload to weight ratio that is 30 % higher than that of existing manipulators. To provide reliability and safety of the ACP, the 3 dimensional graphic simulator is developed. Using the simulator the remote handling operation is simulated and as a result, the optimal layout of ACP is obtained. The supervisory control system is designed to control and monitor the several different unit processes. Also the failure monitoring system is developed to detect the possible accidents of the reduction reactor

Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R.

1994-11-01

The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl x , UAl x -Al and U 3 O 8 -Al cermets, U-5% fissium, UMo, UZrH x , UErZrH, UO 2 -stainless steel cermet, and U 3 O 8 -stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified

Melting process to condition decladding hulls generated by the reprocessing of L.W.R. and F.B.R. spent fuels

International Nuclear Information System (INIS)

Bonniaud, R.; Jacquet-Francillon, N.; Jouan, A.; Sombret, C.

1980-11-01

The fusion compaction of metallic waste from spent fuel hulls is shown to be easily feasible for both Zircaloy and for stainless steel with volume reduction factors of 5 to 7. The Zircaloy copper alloy, put into use to lower the fusion point of the Zircaloy appears extremely interesting both as to the ease with which it can be used and the possibility which it offers of working at temperatures always lower than 1250 0 C. With stainless steel, only the use of silicon enabling the lowering of the temperature to around 1200 0 C appears really feasible. The use of decontaminating agents either during or at the end of the fusion operation seems to be a promising technique, especially in the case of stainless steel where the use of a borosilicated glass is easy. The choice of decontaminating agent is more difficult for Zircaloy and makes necessary the use of molten salts mixtures, the composition of which has not yet been defined. The decontamination factors obtained during the tests run on steel are encouraging, they should be confirmed by further tests in hot cells using real hulls. This study has made it possible to determine the principal parameters necessitated by the setting up of an industrial furnace project. The realisation of fusion compaction units for waste from fuel hulls generated by future reprocessing plants seems to be a real short-term possibility

Development of spent fuel remote handling technology

International Nuclear Information System (INIS)

Yoon, J. S.; Hong, H. D.; Kim, Y. H.

2001-03-01

Since the amount of the spent fuel rapidly increases, the current R and D activities are focused on the technology development related with the storage and utilization of the spent fuel. In this research, to provide such a technology, the mechanical head-end process has been developed. In detail, the swing and shock-free crane and the RCGLUD(Remote Cask Grappling and Lid Unbolting Device) were developed for the safe transportation of the spent fuel assembly, the LLW drum and the transportation cask. Also, the disassembly devices required for the head-end process were developed. This process consists of an assembly downender, a rod extractor, a rod cutter, a fuel decladding device, a skeleton compactor, a force-rectifiable manipulator for the abnormal spent fuel disassembly, and the gantry type telescopic transporter, etc. To provide reliability and safety of these devices, the 3 dimensional graphic design system is developed. In this system, the mechanical devices are modelled and their operation is simulated in the virtual environment using the graphic simulation tools. So that the performance and the operational mal-function can be investigated prior to the fabrication of the devices. All the devices are tested and verified by using the fuel prototype at the mockup facility

Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

Energy Technology Data Exchange (ETDEWEB)

Guenther, R J; Johnson, Jr, A B; Lund, A L; Gilbert, E R [and others

1996-07-01

The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

Selective Trapping of Volatile Fission Products with an Off-Gas Treatment System

Energy Technology Data Exchange (ETDEWEB)

B.R. Westphal; J.J. Park; J.M. Shin; G.I. Park; K.J. Bateman; D.L. Wahlquist

2008-07-01

A head-end processing step, termed DEOX for its emphasis on decladding via oxidation, is being developed for the treatment of spent oxide fuel by pyroprocessing techniques. The head-end step employs high temperatures to oxidize UO2 to U3O8 resulting in the separation of fuel from cladding and the removal of volatile fission products. Development of the head-end step is being performed in collaboration with the Korean Atomic Energy Research Institute (KAERI) through an International Nuclear Energy Research Initiative. Following the initial experimentation for the removal of volatile fission products, an off-gas treatment system was designed in conjunction with KAERI to collect specific fission gases. The primary volatile species targeted for trapping were iodine, technetium, and cesium. Each species is intended to be collected in distinct zones of the off-gas system and within those zones, on individual filters. Separation of the volatile off-gases is achieved thermally as well as chemically given the composition of the filter media. A description of the filter media and a basis for its selection will be given along with the collection mechanisms and design considerations. In addition, results from testing with the off-gas treatment system will be presented.

Penetrate-leach dissolution of zirconium-clad uranium and uranium dioxide fuels

International Nuclear Information System (INIS)

Harmon, H.D.

1975-01-01

A new decladding-dissolution process was developed for zirconium-clad uranium metal and UO 2 fuels. The proposed penetrate-leach process consists of penetrating the zirconium cladding with Alniflex solution (2M HF--1M HNO 3 --1M Al(NO 3 ) 3 --0.1M K 2 Cr 2 O 7 ) and of leaching the exposed core with 10M HNO 3 . Undissolved cladding pieces are discarded as solid waste. Periodic HF and HNO 3 additions, efficient agitation, and in-line zirconium analyses are required for successful control of ZrF 4 and/or AlF 3 precipitation during the cladding-penetration step. Preliminary solvent extraction studies indicated complete recovery of uranium with 30 vol. percent tributyl phosphate (TBP) from both Alniflex solution and blended Alniflex-HNO 3 leach solutions. With 7.5 vol. percent TBP, high extractant/feed flow ratios and low scrub flows are required for satisfactory uranium recovery from Alniflex solution. Modified waste-handling procedures may be required for Alniflex waste, because it cannot be evaporated before neutralization and large quantities of solids are generated on neutralization. The effect of unstable UZr 3 (epsilon phase of uranium-zirconium system) on the safety of penetrate-leach dissolution was investigated

Removal of cesium from aluminum decladding wastes generated in irradiated target processing using a fixed-bed column of resorcinol-formaldehyde resin

International Nuclear Information System (INIS)

Brunson, R.R.; Williams, D.F.; Bond, W.D.; Benker, D.E.; Chattin, F.R.; Collins, E.D.

1994-09-01

The removal of cesium (Cs) from a low-level liquid waste (LLLW) with a cation-exchange column was demonstrated using a resorcinol-formaldehyde (RF) resin. The RF resin was developed at the Westinghouse Savannah River Laboratory (SRL) and is highly specific for the removal of Cs from an alkaline waste of high sodium content. It was determined that the RF resin would be suitable for removing Cs, the largest gamma radiation contributor, from the LLLW generated at the Radiochemical Engineering Development Center located at the Oak Ridge National Laboratory. Presently, the disposal of the LLLW is limited due to the amount of Cs contained in the waste. Cesium removal from the waste solution offers immediate benefits by conserving valuable tank space and would allow cask shipments of the treated waste should the present Laboratory pipelines become unavailable in the future. Preliminary laboratory tests of the RF resins, supplied from two different sources, were used to design a full-scale cation-exchange column for the removal of Cs from a Mark 42 SRL fuel element dejacketing waste solution. The in-cell tests reproduced the preliminary bench-scale test results. The initial Cs breakthrough range was 85--92 column volumes (CV). The resin capacity for Cs was found to be ∼0.35 meq per gram of resin. A 1.5-liter resin bed loaded a combined ∼1,300 Ci of 134 Cs and 137 Cs. A distribution coefficient of ∼110 CV was determined, based on a 50% Cs breakthrough point. The kinetics of the system was studied by examining the rate parameters; however, it was decided that several more tests would be necessary to define the mass transfer characteristics of the system

A respiratory model for uranium aluminide based on occupational data

International Nuclear Information System (INIS)

Leggett, R W; Eckerman, K F; Jr, J D Boice

2005-01-01

As part of an epidemiological study, doses from intake of radionuclides were estimated for workers employed during a 52-year period at the Rocketdyne/Atomics International facility in California. The facility was involved in a variety of research programmes, including nuclear fuel fabrication, spent nuclear fuel decladding, and reactor operation and disassembly. Most of the documented intakes involved inhalation of enriched uranium (U), fission products, or plutonium (Pu). Highest doses were estimated for a group of workers exposed to airborne uranium aluminide (UAl x ) during the fabrication of reactor fuel plates. Much of the exposure to UAl x occurred early in the fuel fabrication programme, before it was recognised that intake and lung retention were being underestimated from urinary data due to an unexpected delayed dissolution of the inhaled material. In workers who had been removed from exposure, the rate of urinary excretion of U increased for a few months, peaked, and then declined at a rate consistent with moderately soluble material. This pattern differs markedly from the monotonically decreasing absorption rates represented by the default absorption types in the Human Respiratory Tract Model (HRTM) of the International Commission on Radiological Protection (ICRP). This paper summarises the findings on the behaviour of UAl x in these workers and describes material-specific parameter values of the HRTM based on this information

Development of advanced spent fuel management process

International Nuclear Information System (INIS)

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

2001-04-01

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

Treatment Deployment Evaluation Tool

International Nuclear Information System (INIS)

M. A. Rynearson; M. M. Plum

1999-01-01

The U.S. Department of Energy (DOE) is responsible for the final disposition of legacy spent nuclear fuel (SNF). As a response, DOE's National Spent Nuclear Fuel Program (NSNFP) has been given the responsibility for the disposition of DOE-owned SNF. Many treatment technologies have been identified to treat some forms of SNF so that the resulting treated product is acceptable by the disposition site. One of these promising treatment processes is the electrometallurgical treatment (EMT) currently in development; a second is an Acid Wash Decladding process. The NSNFP has been tasked with identifying possible strategies for the deployment of these treatment processes in the event that a treatment path is deemed necessary. To support the siting studies of these strategies, economic evaluations are being performed to identify the least-cost deployment path. This model (tool) was developed to consider the full scope of costs, technical feasibility, process material disposition, and schedule attributes over the life of each deployment alternative. Using standard personal computer (PC) software, the model was developed as a comprehensive technology economic assessment tool using a Life-Cycle Cost (LCC) analysis methodology. Model development was planned as a systematic, iterative process of identifying and bounding the required activities to dispose of SNF. To support the evaluation process, activities are decomposed into lower level, easier to estimate activities. Sensitivity studies can then be performed on these activities, defining cost issues and testing results against the originally stated problem

The predicted impacts to the groundwater and Columbia River from ammoniated water discharges to the 216-A-36B crib

International Nuclear Information System (INIS)

Buelt, J.L.; Conbere, W.; Freshley, M.D.; Hicks, R.J.; Kuhn, W.L.; Lamar, D.A.; Serne, R.J.; Smoot, J.L.

1988-03-01

Impact from past and potential future discharges of ammoniated water to the 216-A-36B crib have on groundwater and river concentrations of hazardous chemical constitutents are studied. Until August 1987, the 216-A-36B crib, located in the 200-East Area of the Hanford Site, accepted ammoniated water discharges. Although this study addresses known hazardous chemical constituents associated with such discharges, the primary concern is the discharge of NH 4 OH because of its microbiological conversion to NO 2 /sup /minus// and NO 3 /sup /minus//. As a result of fuel decladding operations, material balance calculations indicate that NH 4 OH has been discharged to the 216-A-36B crib in amounts that exceed reportable quantities under the Comprehensive Environmental Response, Compensation and Liability Act of 1980. Although flow to the crib is relatively constant, the estimated NH 4 OH discharge varies from negligible to a maximum of 10,000 g-molesh. Because these discharges are intermittent, the concentration delivered to the groundwater is a function of soil sorption, microbiological conversion rates of NH 4 + to NO 2 /sup /minus// and NO 3 /sup /minus//, and groundwater dispersion. This report provides results based on the assumptions of maximum, nominal, and discountinued NH 4 OH discharges to the crib. Consequently, the results show maximum and realistic estimates of NH 4 + , NO 2 /sup /minus// and NO 3 /sup /minus// concentrations in the groundwater

Interim results from UO2 fuel oxidation tests in air

International Nuclear Information System (INIS)

Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j.

1987-08-01

An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO 2 , fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO 2 pellets in the temperature range of 135 to 250 0 C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10 5 R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10 5 R/h gamma field. 33 refs., 51 figs., 6 tabs

Management of radioactive waste from reprocessing plants

International Nuclear Information System (INIS)

Kanwar Raj

2010-01-01

Reprocessing and recycling of both fissile and fertile components back into appropriate reactor systems is an integral part of three stage nuclear energy programme of India. Different steps involved in processing of spent nuclear fuel (SNF) are decladding, dissolution and recovery of fissile and fertile materials. Reprocessing of SNF is a complex process involving handling of large quantity of radioactive materials and processing chemicals. There are three reprocessing plants in operation in the country at Trombay, Tarapur and Kalpakkam. Out of these plants, Trombay reprocessing plant is engaged in reprocessing of SNF from research reactors and other two plants are processing of SNF from PHWRs. A facility is being built for reprocessing of thorium based spent fuel at BARC, Trombay based on the experience of pilot plant scale. Like other industrial activities of nuclear fuel cycle, fuel reprocessing facilities too generate various types of radioactive waste streams. These are generated in all the three physical forms namely solid, liquid and gas. These waste streams are primarily categorized on the basis of concentration of radionuclides, their half lives and toxicity. Management of these wastes aims at (a) recovery and recycle of useful materials, (b) concentration and confinement of radioactivity in inert and stable matrices, (c) minimization of final waste volume for disposal, (d) decontamination of effluents following ALARA principle and (e) minimization of radioactive discharge to the environment. The present paper outlines the salient features of management of different types of radioactive waste generated in reprocessing plants handling SNF from research reactors and PHWR

Thoria-fuel irradiation. Program to irradiate 80% ThO2/20% UO2 ceramic pellets at the Savannah River Plant

International Nuclear Information System (INIS)

Pickett, J.B.

1982-02-01

This report describes the fabrication of proliferation-resistant thorium oxide/uranium oxide ceramic fuel pellets and preparations at the Savannah River Laboratory (SRL) to irradiate those materials. The materials were fabricated in order to study head end process steps (decladding, tritium removal, and dissolution) which would be required for an irradiated proliferation-resistant thorium based fuel. The thorium based materials were also to be studied to determine their ability to withstand average commercial light water reactor (LWR) irradiation conditions. This program was a portion of the Thorium Fuel Cycle Technology (TFCT) Program, and was coordinated by the Oak Ridge National Laboratory (ORNL) under the Consolidated Fuel Reprocessing Program (CFRP). The fuel materials were to be irradiated in a Savannah River Plant (SRP) reactor at conditions simulating the heat ratings and burnup of a commercial LWR. The program was terminated due to a de-emphasis of the TFCT Program, following completion of the fabrication of the fuel and the modified assemblies which were to be used in the SRP reactor. The reactor grade ceramic pellets were fabricated for SRL by Battelle, Pacific Northwest Laboratories. Five fuel types were prepared: 100% UO 2 pellets (control); 80% ThO 2 /20% UO 2 pellets; approximately 80% ThO 2 /20% UO 2 + 0.25 CaO (dissolution aid) pellets; 100% UO 2 hybrid pellets (prepared from sol-gel microspheres); and 100% ThO 2 pellets (control). All of the fuel materials were transferred to SRL from PNL and were stored pending a subsequent reactivation of the TFCT Programs

Development of the down-ender and the spent fuel rod cutting device

International Nuclear Information System (INIS)

Kim, S. H.; Yoon, Ji Sup; Kim, Young Hwan; Hoo, Jung Jae; Hong, Dong Hee; Kim, Do Woo

2000-07-01

It is necessary to disassemble the spent fuel assembly for the recycling of the PWR spent fuels. The spent fuel disassembling process includes transportation and handling of the spent fuel assembly, extraction and cutting of the spent fuel rods, and extraction of the spent fuel pellets(decladding). In this study, the downender of the spent fuel assembly and the spent fuel rod cutting device have been developed. The downender is used to change the posture of the spent fuel assembly from the vertical to the horizontal directions, prior to extracting the fuel rods. The concepts of the remote operation and maintenance has been introduced in the design of the downender. Also, the several design consideration has been given such as the reliable adaptation of the vertically accessing the assembly to the device, the minimization of the shock force when settling down the assembly, and the interface with the rod extraction device without intermittent operation. The spent fuel rod cutting device using a tube cutter is developed for cutting the fuel rods to the suitable size. In designing this device, the mechanical property of the spent fuel rod is examined such as the strength of the clad material and the optimal size of the rod for the extracting process. Also, several cutting methods, which are commercially available, are investigated and tested in terms of the durability, the deformation on the cutting surface of the rods, and the amount of the generated debris, and the fire risk. As like the downender, the design of this device accommodates the concepts of the remote operation and maintenance

Dry refabrication technology development of spent nuclear fuel

International Nuclear Information System (INIS)

Park, Geun Il; Lee, J. W.; Song, K. C.

2012-04-01

Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed

Dry refabrication technology development of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Park, Geun Il; Lee, J. W.; Song, K. C.; and others

2012-04-15

Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed.

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

International Nuclear Information System (INIS)

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

1998-10-01

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

Development of spent fuel remote handling technology

Energy Technology Data Exchange (ETDEWEB)

Park, B. S.; Yoon, J. S.; Hong, H. D. (and others)

2007-02-15

In this research, the remote handling technology was developed for the ACP application. The ACP gives a possible solution to reduce the rapidly cumulative amount of spent fuels generated from the nuclear power plants in Korea. The remote technologies developed in this work are a slitting device, a voloxidizer, a modified telescopic servo manipulator and a digital mock-up. A slitting device was developed to declad the spent fuel rod-cuts and collect the spent fuel UO{sub 2} pellets. A voloxidizer was developed to convert the spent fuel UO{sub 2} pellets obtained from the slitting process in to U{sub 3}O{sub 8} powder. Experiments were performed to test the capabilities and remote operation of the developed slitting device and voloxidizer by using simulated rod-cuts and fuel in the ACP hot cell. A telescopic servo manipulator was redesigned and manufactured improving the structure of the prototype. This servo manipulator was installed in the ACP hot cell, and the target module for maintenance of the process equipment was selected. The optimal procedures for remote operation were made through the maintenance tests by using the servo manipulator. The ACP digital mockup in a virtual environment was established to secure a reliability and safety of remote operation and maintenance. The simulation for the remote operation and maintenance was implemented and the operability was analyzed. A digital mockup about the preliminary conceptual design of an enginnering-scale ACP was established, and an analysis about a scale of facility and remote handling was accomplished. The real-time diagnostic technique was developed to detect the possible fault accidents of the slitting device. An assessment of radiation effect for various sensors was also conducted in the radiation environment.

Radioactive Waste Conditioning, Immobilisation, And Encapsulation Processes And Technologies: Overview And Advances (Chapter 7)

Energy Technology Data Exchange (ETDEWEB)

Jantzen, Carol M. [Savannah River National Lab., Aiken SC (United States); Lee, William E. [Imperial College, London (United Kingdom). Dept. of Materials; Ojovan, Michael I. [Univ. of Sheffield (United Kingdom). Dept. of Materials Science and Engineering

2012-10-19

The main immobilization technologies that are available commercially and have been demonstrated to be viable are cementation, bituminization, and vitrification. Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either alkali borosilicate glass or alkali aluminophosphate glass. The exact compositions of nuclear waste glasses are tailored for easy preparation and melting, avoidance of glass-in-glass phase separation, avoidance of uncontrolled crystallization, and acceptable chemical durability, e.g., leach resistance. Glass has also been used to stabilize a variety of low level wastes (LLW) and mixed (radioactive and hazardous) low level wastes (MLLW) from other sources such as fuel rod cladding/decladding processes, chemical separations, radioactive sources, radioactive mill tailings, contaminated soils, medical research applications, and other commercial processes. The sources of radioactive waste generation are captured in other chapters in this book regarding the individual practices in various countries (legacy wastes, currently generated wastes, and future waste generation). Future waste generation is primarily driven by interest in sources of clean energy and this has led to an increased interest in advanced nuclear power production. The development of advanced wasteforms is a necessary component of the new nuclear power plant (NPP) flowsheets. Therefore, advanced nuclear wasteforms are being designed for robust disposal strategies. A brief summary is given of existing and advanced wasteforms: glass, glass-ceramics, glass composite materials (GCM’s), and crystalline ceramic (mineral) wasteforms that chemically incorporate radionuclides and hazardous species atomically in their structure. Cementitious, geopolymer, bitumen, and other encapsulant wasteforms and composites that atomically bond and encapsulate

Chemical Engineering Division fuel cycle programs. Progress report, January--March 1978

Energy Technology Data Exchange (ETDEWEB)

Steindler, M.J.; Ader, M.; Barletta, R.E.

1979-04-01

Fuel cycle studies reported for this period include studies of advanced solvent extraction techniques focussed on the development of centrifugal contactors for use in Purex processes. Miniature single-stage and eight-stage centrifugal contactors are being employed in studies of contactor performance and the kinetics of extraction. A 9-cm-ID centrifugal contactor has been completed, and fabrication drawings are being prepared for a plant-scale contactor. In other work, tricaprylmethyl-ammonium nitrate and di-n-amyl n-amylphosphonate are being evaluated as extractants in the Thorex process. Literature on the dispersion of liquids by explosions is being reviewed. A process was developed for extracting TBP degradation products from TBP-Na/sub 2/CO/sub 3/ scrub solutions while the actinides remain with the raffinate. In the program on pyrochemical and dry processing of nuclear fuel, the literature is being reviewed for acceptable materials for containment vessels, decladding methods are being evaluated, salt transport processes are being studied, a candidate flow sheet (based upon the Dow Aluminum Pyrometallurgical process) for reprocessing spent uranium metal fuel was prepared, work was begun on the use of molten salts for reprocessing actinide oxides, and the reprocessing of (Th,U)O/sub 2/ solid solution in a KCl-LiCl salt containing ThCl/sub 4/ and thorium chips was studied. Work on the encapsulation of solidified radioactive waste in a metal matrix includes study of (1) chemical interactions between simulated waste forms and matrix metals, (2) the leach rates of simulated encapsulated waste forms, and (3) the corrosion of candidate matrix metals and canister materials in brine solutions.Work to establish criteria for the handling of waste cladding hulls is continuing. The transport properties of nuclear waste in geologic media are being studied to estimate leaching of radionuclides from deep repositories by groundwater.

Methodologies to determine the Pu content of spent fuel assemblies for input nuclear material accountancy of pyroporcessing

International Nuclear Information System (INIS)

Lee, Taehoon; Shin, Heesung; Kim, Youngsoo; Kim, Hodong; Kwon, Taeje

2011-01-01

This study shows two different non-destructive approaches to determine the Pu mass of spent fuel assemblies, and the analysis results on the errors in their Pu mass. For both methods, the Cm mass of the assembly is obtained based on the neutron measurement results. The Cm ratio of the assembly is determined from the Cm mass and the Pu mass obtained by using either of the two methods. In a comparison of two methods, the second method is simpler than the first and may not need a homogeneously-mixed sample of the spent fuel assembly. On the other hand, the second approach shows larger error in the estimated Pu mass than the first one for many different spent fuel cases of various burnup, initial enrichment, and cooling times. A member state support program for the development of the IAEA safeguards approach for an engineering-scale pyroprocessing facility, which is designated as the Reference Engineering-scale Pyroprocessing Facility(REPF), has been carried out by Korea Atomic Energy Research Institute since 2008. The nuclear material accountancy of the REPF is based on the 'Cm balance' technique. The Pu content of processing materials of pyroprocessing can be determined by measuring the Cm mass of the materials and multiplying it by the Cm ratio. The spent fuel assembly is de-cladded, and the irradiated UO 2 material of the assembly is homogeneously mixed in the homogenization process in order to obtain a representative sample of the spent fuel assembly for determining the mass of Pu, U and Cm elements, as well as the Cm ratio of the campaign. The shipper-receiver difference between the nuclear power plant and HPC of REPF is determined at this point. We found that the error for the Pu mass and Cm ratio determined from the homogenized uranium oxide powder is the most critical for the determination of the material unaccounted for throughout the whole processes. This paper presents two approaches to determine the Pu mass of spent fuel assemblies using non

Dismantlement and Radioactive Waste Management of DPRK Nuclear Facilities

Energy Technology Data Exchange (ETDEWEB)

Jooho, W.; Baldwin, G. T.

2005-04-01

One critical aspect of any denuclearization of the Democratic People’s Republic of Korea (DPRK) involves dismantlement of its nuclear facilities and management of their associated radioactive wastes. The decommissioning problem for its two principal operational plutonium facilities at Yongbyun, the 5MWe nuclear reactor and the Radiochemical Laboratory reprocessing facility, alone present a formidable challenge. Dismantling those facilities will create radioactive waste in addition to existing inventories of spent fuel and reprocessing wastes. Negotiations with the DPRK, such as the Six Party Talks, need to appreciate the enormous scale of the radioactive waste management problem resulting from dismantlement. The two operating plutonium facilities, along with their legacy wastes, will result in anywhere from 50 to 100 metric tons of uranium spent fuel, as much as 500,000 liters of liquid high-level waste, as well as miscellaneous high-level waste sources from the Radiochemical Laboratory. A substantial quantity of intermediate-level waste will result from disposing 600 metric tons of graphite from the reactor, an undetermined quantity of chemical decladding liquid waste from reprocessing, and hundreds of tons of contaminated concrete and metal from facility dismantlement. Various facilities for dismantlement, decontamination, waste treatment and packaging, and storage will be needed. The shipment of spent fuel and liquid high level waste out of the DPRK is also likely to be required. Nuclear facility dismantlement and radioactive waste management in the DPRK are all the more difficult because of nuclear nonproliferation constraints, including the call by the United States for “complete, verifiable and irreversible dismantlement,” or “CVID.” It is desirable to accomplish dismantlement quickly, but many aspects of the radioactive waste management cannot be achieved without careful assessment, planning and preparation, sustained commitment, and long

Dismantlement and Radioactive Waste Management of DPRK Nuclear Facilities

International Nuclear Information System (INIS)

Jooho, W.; Baldwin, G.T.

2005-01-01

One critical aspect of any denuclearization of the Democratic People's Republic of Korea (DPRK) involves dismantlement of its nuclear facilities and management of their associated radioactive wastes. The decommissioning problem for its two principal operational plutonium facilities at Yongbyun, the 5MWe nuclear reactor and the Radiochemical Laboratory reprocessing facility, alone present a formidable challenge. Dismantling those facilities will create radioactive waste in addition to existing inventories of spent fuel and reprocessing wastes. Negotiations with the DPRK, such as the Six Party Talks, need to appreciate the enormous scale of the radioactive waste management problem resulting from dismantlement. The two operating plutonium facilities, along with their legacy wastes, will result in anywhere from 50 to 100 metric tons of uranium spent fuel, as much as 500,000 liters of liquid high-level waste, as well as miscellaneous high-level waste sources from the Radiochemical Laboratory. A substantial quantity of intermediate-level waste will result from disposing 600 metric tons of graphite from the reactor, an undetermined quantity of chemical decladding liquid waste from reprocessing, and hundreds of tons of contaminated concrete and metal from facility dismantlement. Various facilities for dismantlement, decontamination, waste treatment and packaging, and storage will be needed. The shipment of spent fuel and liquid high level waste out of the DPRK is also likely to be required. Nuclear facility dismantlement and radioactive waste management in the DPRK are all the more difficult because of nuclear nonproliferation constraints, including the call by the United States for 'complete, verifiable and irreversible dismantlement,' or 'CVID.' It is desirable to accomplish dismantlement quickly, but many aspects of the radioactive waste management cannot be achieved without careful assessment, planning and preparation, sustained commitment, and long completion times

The MaNGA integral field unit fiber feed system for the Sloan 2.5 m telescope

Energy Technology Data Exchange (ETDEWEB)

Drory, N. [McDonald Observatory, The University of Texas at Austin, 1 University Station, Austin, TX 78712 (United States); MacDonald, N.; Byler, N. [Department of Astronomy, University of Washington, Box 351580 Seattle, WA 98195 (United States); Bershady, M. A.; Smith, M.; Tremonti, C. A.; Wake, D. A.; Eigenbrot, A.; Jaehnig, K. [Department of Astronomy, University of Wisconsin, 475 N. Charter St., Madison, WI 53706 (United States); Bundy, K. [Kavli Institute for the Physics and Mathematics of The Universe (Kavli IPMU, WPI), Todai Institutes for Advanced Study, The University of Tokyo, Kashiwa, Japan 277-8583 (Japan); Gunn, J. [Department of Astrophysical Sciences, Princeton University, Princeton, NJ 08544 (United States); Law, D. R.; Cherinka, B. [Dunlap Institute for Astronomy and Astrophysics, University of Toronto, 50 St. George St, Toronto, ON M5S 3H4 (Canada); Stoll, R. [C Technologies, Inc., 757 Route 202/206, Bridgewater, NJ 08807 (United States); Yan, R. [Department of Physics and Astronomy, University of Kentucky, Lexington, Kentucky, 40506-0055 (United States); Weijmans, A. M. [School of Physics and Astronomy, University of St Andrews, North Haugh, St Andrews, Fife KY16 9SS (United Kingdom); Cope, F.; Holder, D.; Huehnerhoff, J. [Apache Point Observatory, P.O. Box 59, Sunspot, NM 88349 (United States); Harding, P., E-mail: drory@astro.as.utexas.edu [Department of Astronomy, Case Western Reserve University, Cleveland, OH 44106 (United States); and others

2015-02-01

throughput and fiber metrology. Future applications include larger IFUs, higher fill factors with stripped buffer, de-cladding, and lenslet coupling.

Internal dose evaluation from actinide intakes during nuclear power reactor spent fuel reprocessing

International Nuclear Information System (INIS)

Pawar, S.K.; Kumar, Ranjeet; Gamre, Rupali; Purohit, R.G.

2011-01-01

Full text: Indian PHWR reactors are using natural uranium as fuel. After use they are discharged from the core and send for fuel reprocessing to extract the unused uranium and plutonium. Plutonium and other actinides are formed by activation of 238 U with neutrons and subsequent decay. During reprocessing of the spent fuel, major long lived actinides (Pu, Am and U) may become radiological safety hazard. Actinides intakes are more probable during declading and chopping of spent fuel. During routine plant operation in reprocessing, exposure to Pu is a major concern along with Am and U in working environment due to its higher radiological hazard and occupational workers are likely to get exposed to plutonium, Americium and Uranium mostly through inhalation. Internally deposited Pu-isotopes, Am-isotope and U-isotopes are estimated using techniques such as lung counting (in-vivo) and urine and faecal bioassay (in-vitro). Evaluation of internal dose of actinides is dependent upon urinary excreted activity. To estimate the internally deposited Pu, U and Am at an intake level of about one ALI (ICRP-78, 1997) of occupational workers, urine bioassay is the preferred technique due to high detection sensitivity, ease of sample handling and economical method. A small and measurable fraction of internally deposited Pu, Am and U are excreted through urine whose content is dependent on time of inhalation, quantity and type of chemical form of inhaled material (S and M class). A standardized radiochemical analysis method for separation and estimation of Pu, Am and U is used to evaluate the urinary excreted activity and internal dose. Several measurements techniques are employed for the estimation of plutonium, Americium and Uranium for example, Alpha Spectrometry, Gamma Spectrometry, Neutron Activation Analysis, Mass Spectrometry and Fission Track Analysis. The radiochemical separation followed by alpha counting and/or spectrometry is chosen due to its ease of handling and

The MaNGA integral field unit fiber feed system for the Sloan 2.5 m telescope

International Nuclear Information System (INIS)

Drory, N.; MacDonald, N.; Byler, N.; Bershady, M. A.; Smith, M.; Tremonti, C. A.; Wake, D. A.; Eigenbrot, A.; Jaehnig, K.; Bundy, K.; Gunn, J.; Law, D. R.; Cherinka, B.; Stoll, R.; Yan, R.; Weijmans, A. M.; Cope, F.; Holder, D.; Huehnerhoff, J.; Harding, P.

2015-01-01

and fiber metrology. Future applications include larger IFUs, higher fill factors with stripped buffer, de-cladding, and lenslet coupling.

Neutron activation analysis of maltenes recovered from EUROBITUM simulates

International Nuclear Information System (INIS)

Impens, N

2006-01-01

According to the present Belgian reference scenario, Eurobitum bituminised radioactive waste has to be disposed off in a deep underground repository in a stable geological formation such as Boom Clay. This waste originated mainly from mixtures of nuclear fuel decladding slurries and waste concentrates from the nuclear fuel cycle. Even though safety assessment studies up till present do not show that this waste is unacceptable for deep underground disposal, a final decision about the disposal of the bituminised waste has not been taken so far, and alternative solutions are still conceivable. To support the decision-making we investigate methods to recondition this bituminised waste. We continued studying a room temperature re-treatment method for Eurobitum. The aim of the method is the stabilisation and minimisation of final waste, and the free release of recovered materials. The method comprises the recovery of maltenes and water soluble salts. The recovery of maltenes is performed by dissolving the complete bitumen matrix with a 'solvent', followed by the precipitation of the asphaltenes by addition of a so-called 'nonsolvent'. The 'solvent' is a 50 percent aromatic blend of Shellsol A150 and Shellsol H, whereas the 'nonsolvent' is aliphatic Shellsol T. The recovered maltenes represent 40 % wt of the waste, as shown in the inner pie chart. Part of the maltenes could not be recovered and remain in the asphaltene matrix, as can be seen from the difference between the inner and outer pie chart, representing the real composition and the weight fractions after separation, respectively. A second step of the room temperature re-treatment method covers the complete removal of the water soluble (nitrate) salts, and is described in the main reference and references therein. Application of the room temperature re-treatment method results in a final waste that consists of water insoluble salts embedded in an asphaltene matrix which is less sensitive towards radiolysis than

Fast Track Characterization of Highly Radioactive Waste Pits Combining Off-the-Shelf Robotics with Innovative Investigation Protocols

International Nuclear Information System (INIS)

Chabeuf, Jean-Michel; Boya, Didier

2008-01-01

The investigation and characterization of radioactive waste pits and effluent storage tanks represents a substantial and challenging step in the overall decommissioning programme launched by AREVA NC in 1998 on the site of Marcoule on behalf of the French Atomic Energy commission. Physical ,radiological and regulatory constraints, combined with a tight schedule, have lead our teams to use proven conventional instrumentation and robotics in innovative configurations . One such investigation, conducted on a particularly challenging radioactive effluent storage pit, is described below. The 'H' pit is a stainless steel clad concrete cavity, located in the second basement of the de-cladding building of Marcoule site. It was used for forty years as buffer storage for high activity effluents and has a length of 5 meters, a width of 3 meters , a height of 2.5 meters, and is topped by lead plates over 5 cm thick and The bottom of the cavity is covered with a layer of mud containing mainly graphite, diatoms and resins. The mud level ranges from about 20 centimeters to over 50 centimeters. The overall mud volume is around 2.4 cubic meters. Ambient dose rates above the lead plates exceed 10 mSv/h. The main purpose of our investigation was to characterize the muds for future recovery and conditioning prior to decontaminating the pit. The history of the pit together with the varying mud altimetry lead us to believe that sedimentation had probably occurred throughout the years. We thus decided to combine dose rate measurements using IF104 probes, gamma spectroscopy with CdTe probes and sample collections at different depths to ensure the representativeness and full characterization of the muds. Poor access, ambient dose rates have lead us to conceive a robotic arm, mounted on an shaft which can be modified to fit a wide range of pits and tanks. Custom built robotic tools with maximum manoeuvrability generally involve costs and delays far exceeding our purposes. SIT, a French

Feasibility study on commercialization of fast breeder reactor cycle systems interim report of phase II. Technical study report for nuclear fuel cycle systems

International Nuclear Information System (INIS)

Sato, Koji; Amamoto, Ippei; Inoue, Akira

2004-06-01

combines the oxide electrowinning reprocessing and the vibropacking fuel fabrication process has the possibility of economical improvement if simplification of the process can be achieved. However this system has several primary problems which affect the technical feasibility of the MA recovery process etc. In the system which combines the metal electrorefining reprocessing and the metal casting fuel fabrication process for metal fuels, high economical efficiency is generally expected of small-scale facilities, although verification of the TRU recovery process by spent fuel is required. The development of the head-end process (reduction technology of oxide to metal) is required for introducing the metal fuel cycle. As for nitride fuels, specific technical developments of N-15 enrichment, recycling, and nitride conversion to the nitride fuel, etc. are required as the main technology, although the advanced aqueous reprocessing or the metal electrorefining reprocessing can be applied fundamentally. Besides technical developments of the fuel fabrication, assembly setting, dismantling, and decladding, etc. are required for coated particle fuels. For economical efficiency, less than 0.8 yen/kWh which is the demand value (total of the reprocessing expense and the fuel fabrication expense) of the fuel cycle expense of phase II satisfied each combination case at 200 tHM/y scale provisionally set for large-scale facilities. On the other hand, only the metal fuel cycle almost satisfied the demand value in the combination case with a high breeder reactor core at 50 tHM/y scale provisionally set for small-scale facilities. The combination case with a low breeder reactor core has satisfied the demand value, with improvement of the average burnup by the radial direction blanket fuel deletion contributing to the decrease of the fuel cycle expense. (author)

3. forum of radiation protection sciences - Book of abstracts

International Nuclear Information System (INIS)

2002-10-01

facility (Fabien Jeanne, Loic Lachevre, David Meyer, Moustapha Messaoudi, Assystem Services); P11 - Project of Marcoule's wastes management and conditioning: wastes sampling and characterization at the decladding and MAR 400 workshops (Ph. Chany, G. Cauquil, Ph. Spizz, H. Comole, R. Chomel, Cogema Marcoule); Demonstrations: D01 - Presentation of tele-operated demolition systems - Brokk-type (ATC company); D02 - Determination of the contamination depth inside a material (Benoit Marc, Charly Mahe, Flavien Brossard, Techman Industrie); D03 - A tele-operated system for dismantling operations in contaminated environment - SAMM on Brokk (Cybernetix company); D04 - Techclean surface decontamination process (Salvarem company)