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Sample records for fuel fabrication development

  1. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  2. Process development and fabrication for sphere-pac fuel rods

    International Nuclear Information System (INIS)

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted

  3. Development of Nuclear Fuel Remote Fabrication Technology

    International Nuclear Information System (INIS)

    Lee, Jung Won; Yang, M. S.; Kim, S. S. and others

    2005-04-01

    The aim of this study is to develop the essential technology of dry refabrication using spent fuel materials in a laboratory scale on the basis of proliferation resistance policy. The emphasis is placed on the assessment and the development of the essential technology of dry refabrication using spent fuel materials. In this study, the remote fuel fabrication technology to make a dry refabricated fuel with an enhanced quality was established. And the instrumented fuel pellets and mini-elements were manufactured for the irradiation testing in HANARO. The design and development technology of the remote fabrication equipment and the remote operating and maintenance technology of the equipment in hot cell were also achieved. These achievements will be used in and applied to the future back-end fuel cycle and GEN-IV fuel cycle and be a milestone for Korea to be an advanced nuclear country in the world

  4. Development of PHWR fuel fabrication in Korea

    International Nuclear Information System (INIS)

    Suh, K.S.; Yang, M.S.; Kim, D.H.; Rim, C.S.

    1988-01-01

    Korea Advanced Energy Research Institute (KAERI) started a research project to develop the PHWR (CANDU) nuclear fuel fabrication technology in 1981. Based on the results of the intensive developmental work, several prototype fuel bundles were fabricated and tested in the Hot Test Loop at KAERI continuously in 1983 and 1984. After that, irradiation test and post-irradiation examination were carried out for two KAERI-made fuel bundles at Chalk River Nuclear Laboratories in Canada in 1984. Since the results of in-pile and out-of-pile tests with prototype fuel bundles proved to be satisfactory, 48 additional fuel bundles were loaded in Wolsung reactor (CANDU) in 1984 and 1985, and all of them were discharged without a defect after excellent performance in the power reactor. In 1985, the Korean government decided that KAERI supplies all the fuel necessary for the Wolsung reactor. For the mass production of nuclear fuel bundle, several process equipment, facilities and automation methods have been improved making use of experience accumulated during research. A quality assurance program was also established, and quality inspection technology was reviewed and improved to fit the mass production. This paper deals with the development experience so far obtained with the design and fabrication of the Korean PHWR fuel

  5. Development of fabrication technology for CANDU advanced fuel -Development of the advanced CANDU technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Beom; Kim, Hyeong Soo; Kim, Sang Won; Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Jang, Ho Il; Kim, Sang Sik; Choi, Il Kwon; Cho, Dae Sik; Sheo, Seung Won; Lee, Soo Cheol; Kim, Yoon Hoi; Park, Choon Ho; Jeong, Seong Hoon; Kang, Myeong Soo; Park, Kwang Seok; Oh, Hee Kwan; Jang, Hong Seop; Kim, Yang Kon; Shin, Won Cheol; Lee, Do Yeon; Beon, Yeong Cheol; Lee, Sang Uh; Sho, Dal Yeong; Han, Eun Deok; Kim, Bong Soon; Park, Cheol Joo; Lee, Kyu Am; Yeon, Jin Yeong; Choi, Seok Mo; Shon, Jae Moon [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    The present study is to develop the advanced CANDU fuel fabrication technologies by means of applying the R and D results and experiences gained from localization of mass production technologies of CANDU fuels. The annual portion of this year study includes following: 1. manufacturing of demo-fuel bundles for out-of-pile testing 2. development of technologies for the fabrication and inspection of advanced fuels 3. design and munufacturing of fuel fabrication facilities 4. performance of fundamental studies related to the development of advanced fuel fabrication technology.

  6. Fuel Fabrication Capability Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Burkes, Douglas

    2013-06-28

    The purpose of this document is to provide a comprehensive review of the mission of the Fuel Fabrication Capability (FFC) within the Global Threat Reduction Initiative (GTRI) Convert Program, along with research and development (R&D) needs that have been identified as necessary to ensuring mission success. The design and fabrication of successful nuclear fuels must be closely linked endeavors.

  7. Current developments of fuel fabrication technologies at the plutonium fuel production facility, PFPF

    International Nuclear Information System (INIS)

    Asakura, K.; Aono, S.; Yamaguchi, T.; Deguchi, M.

    2000-01-01

    The Japan Nuclear Cycle Development Institute, JNC, designed, constructed and has operated the Plutonium Fuel Production Facility, PFPF, at the JNC Tokai Works to supply MOX fuels to the proto-type Fast Breeder Reactor, FBR, 'MONJU' and the experimental FBR 'JOYO' with 5 tonMOX/year of fabrication capability. Reduction of personal radiation exposure to a large amount of plutonium is one of the most important subjects in the development of MOX fabrication facility on a large scale. As the solution of this issue, the PFPF has introduced automated and/or remote controlled equipment in conjunction with computer controlled operation scheme. The PFPF started its operation in 1988 with JOYO reload fuel fabrication and has demonstrated MOX fuel fabrication on a large scale through JOYO and MONJU fuel fabrication for this decade. Through these operations, it has become obvious that several numbers of equipment initially installed in the PFPF need improvements in their performance and maintenance for commercial utilization of plutonium in the future. Furthermore, fuel fabrication of low density MOX pellets adopted in the MONJU fuel required a complete inspection because of difficulties in pellet fabrication compared with high density pellet for JOYO. This paper describes new pressing equipment with a powder recovery system, and pellet finishing and inspection equipment which has multiple functions, such as grinding measurements of outer diameter and density, and inspection of appearance to improve efficiency in the pellet finishing and inspection steps. Another development of technology concerning an annular pellet and an innovative process for MOX fuel fabrication are also described in this paper. (author)

  8. MOX fuel fabrication: Technical and industrial developments

    International Nuclear Information System (INIS)

    Lebastard, G.; Bairiot, H.

    1990-01-01

    The plutonium available in the near future is generally estimated rather precisely on the basis of the reprocessing contracts and the performance of the reprocessing plants. A few years ago, decision makers were convinced that a significant share of this fissile material would be used as the feed material for fast breeder reactors (FBRs) or other advanced reactors. The facts today are that large reprocessing plants are coming into commercial operations: UP3 and soon UP2-800 and THORP, but that FBR deployment is delayed worldwide. As a consequence, large quantities of plutonium will be recycled in light water reactors as mixed oxide (MOX) fuels. MOX fuel technology has been properly demonstrated in the past 25 years. All specific problems have been addressed, efficient fabrication processes and engineering background have been implemented to a level of maturity which makes MOX fuel behaving as well as Uranium fuel. The paper concentrates on todays MOX fabrication expertise and presents the technical and industrial developments prepared by the MOX fuel fabrication industry for this last decade of the century

  9. Development of technology of high density LEU dispersion fuel fabrication

    International Nuclear Information System (INIS)

    Wiencek, T.; Totev, T.

    2007-01-01

    Advanced Materials Fabrication Facilities at Argonne National Laboratory have been involved in development of LEU dispersion fuel for research and test reactors from the beginning of RERTR program. This paper presents development of technology of high density LEU dispersion fuel fabrication for full size plate type fuel elements. A brief description of Advanced Materials Fabrication Facilities where development of the technology was carried out is given. A flow diagram of the manufacturing process is presented. U-Mo powder was manufactured by the rotating electrode process. The atomization produced a U-Mo alloy powder with a relatively uniform size distribution and a nearly spherical shape. Test plates were fabricated using tungsten and depleted U-7 wt.% Mo alloy, 4043 Al and Al-2 wt% Si matrices with Al 6061 aluminum alloy for the cladding. During the development of the technology of manufacturing of full size high density LEU dispersion fuel plates special attention was paid to meet the required homogeneity, bonding, dimensions, fuel out of zone and other mechanical characteristics of the plates.

  10. Development of equipment for fabricating DUPIC fuel powder

    International Nuclear Information System (INIS)

    Kim, Ki Ho; Yang, M. S.; Park, J. J.; Lee, J. W.; Kim, J. H.; Cho, K. H.; Lee, D. Y.; Lee, Y. S.; Na, S. H.

    1999-06-01

    The powder fabrication processes, as the first stage of manufacturing DUPIC (Direct Use of PWR spent fuel In CANDU) fuel, consist of the slitting of spent PWR fuel rods, REOX (Oxidation and REduction of Oxide Fuels) processing to produce the powder feedstock, the milling of the produced powder, the granulation of the milled powder, and the mixing of the granulated powder with pressing lubricants. All these processes should be conducted by remote means in a hot-cell environment where the direct human access is limited to the strictest minimum due to the high radioactivity. This report describe the development of the equipment for fabricating DUPIC fuel powder. These equipment are Slitting Machine, Oxidation and Reduction (OREOX) Furnace, Mill, Roll Compactor, and Mixer. Remote design concept was applied to all the equipment for use in the M6 hot-cell of the IMEF. Mechanical design considerations and capabilities of the equipment for remote operation and maintenance are presented. First prototypes were developed and installed in the DUPIC full scale mock-up and tested using a master-slave manipulator. Redesign and reconstruction were made on each equipment based on mock-up test results. The remote technology acquired through this research was utilized in developing other equipment for DUPIC fuel fabrication, thereby improving safety and increasing productivity. This technology could also be extended to the area of remote handling equipment development for use in hazardous environments. (author). 14 refs., 9 tabs., 21 figs

  11. Development of equipment for fabricating DUPIC fuel powder

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Ho; Yang, M. S.; Park, J. J.; Lee, J. W.; Kim, J. H.; Cho, K. H.; Lee, D. Y.; Lee, Y. S.; Na, S. H

    1999-06-01

    The powder fabrication processes, as the first stage of manufacturing DUPIC (Direct Use of PWR spent fuel In CANDU) fuel, consist of the slitting of spent PWR fuel rods, REOX (Oxidation and REduction of Oxide Fuels) processing to produce the powder feedstock, the milling of the produced powder, the granulation of the milled powder, and the mixing of the granulated powder with pressing lubricants. All these processes should be conducted by remote means in a hot-cell environment where the direct human access is limited to the strictest minimum due to the high radioactivity. This report describe the development of the equipment for fabricating DUPIC fuel powder. These equipment are Slitting Machine, Oxidation and Reduction (OREOX) Furnace, Mill, Roll Compactor, and Mixer. Remote design concept was applied to all the equipment for use in the M6 hot-cell of the IMEF. Mechanical design considerations and capabilities of the equipment for remote operation and maintenance are presented. First prototypes were developed and installed in the DUPIC full scale mock-up and tested using a master-slave manipulator. Redesign and reconstruction were made on each equipment based on mock-up test results. The remote technology acquired through this research was utilized in developing other equipment for DUPIC fuel fabrication, thereby improving safety and increasing productivity. This technology could also be extended to the area of remote handling equipment development for use in hazardous environments. (author). 14 refs., 9 tabs., 21 figs.

  12. Development of CANFLEX fuel fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Kang, M. S.; Choi, C. B.; Park, C. H.; Kwon, W. J.; Kim, C. H.; Kim, B. J.; Koo, C. H.; Cho, D. S.; So, D. Y.; Suh, S. W.; Park, C. J.; Chang, D. H.; Yun, S. H. [KEPCO Nuclear Fuel Company, Taejeon (Korea)

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU(CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. One of the improvements in CANDU fuel fabrication technology, and advanced method of Zr-Be brazing was developed. For the formation of Zr-Be alloy, preheating and main heating temperature in the furnace is 700 deg C, 1200 deg C respectively. In order to find an appropriate material for the brazing joints in the CANDU fuel, the composition of Zr based amorphous metals were designed. And, the effect of hydrogen on the mechanical properties of cladding sheath and feasibility of the eddy current test to evaluate quality of end cap weld were also studied for the fundamental research purpose. As a preliminary study to suggest optimal way for the mass production of CANFLEX-NU fuel at KNFC the existing CANDU fuel facilities and fabrication/inspection processes were reviewed. The best way is that the current CANDU facility shall be modified to produce small diametrial CANFLEX elements and a new facility shall be constructed to produce large diametrial CANFLEX fuel elements. 46 refs., 99 figs., 10 tabs. (Author)

  13. Fuel Fabrication Capability Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Burkes, Douglas

    2014-04-17

    The purpose of this document is to provide a comprehensive review of the mission of the Fuel Fabrication Capability (FFC) within the Global Threat Reduction Initiative Convert Program, along with research and development (R&D) needs that have been identified as necessary to ensuring mission success. The design and fabrication of successful nuclear fuels must be closely linked endeavors. Therefore, the overriding motivation behind the FFC R&D program described in this plan is to foster closer integration between fuel design and fabrication to reduce programmatic risk. These motivating factors are all interrelated, and progress addressing one will aid understanding of the others. The FFC R&D needs fall into two principal categories, 1) baseline process optimization, to refine the existing fabrication technologies, and 2) manufacturing process alternatives, to evaluate new fabrication technologies that could provide improvements in quality, repeatability, material utilization, or cost. The FFC R&D Plan examines efforts currently under way in regard to coupon, foil, plate, and fuel element manufacturing, and provides recommendations for a number of R&D topics that are of high priority but not currently funded (i.e., knowledge gaps). The plan ties all FFC R&D efforts into a unified vision that supports the overall Convert Program schedule in general, and the fabrication schedule leading up to the MP-1 and FSP-1 irradiation experiments specifically. The fabrication technology decision gates and down-selection logic and schedules are tied to the schedule for fabricating the MP-1 fuel plates, which will provide the necessary data to make a final fuel fabrication process down-selection. Because of the short turnaround between MP-1 and the follow-on FSP-1 and MP-2 experiments, the suite of specimen types that will be available for MP-1 will be the same as those available for FSP-1 and MP-2. Therefore, the only opportunity to explore parameter space and alternative processing

  14. Metallic Reactor Fuel Fabrication for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.

  15. Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Bum; Park, Choon Hoh; Park, Chul Joo; Kwon, Woo Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO{sub 2}, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author).

  16. Nuclear fuel fabrication - developing indigenous capability

    International Nuclear Information System (INIS)

    Gupta, U.C.; Jayaraj, R.N.; Meena, R.; Sastry, V.S.; Radhakrishna, C.; Rao, S.M.; Sinha, K.K.

    1997-01-01

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

  17. A review on the development of the MOX fuel fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, See Hyung; Lee, Yung Woo; Sohn, Dong Sung; Yang, Myung Seung; Bae, Kee Kwang; Nah, Sang Hoh; Kim, Han Soo; Lee, Jung Won; Kim, Bong Koo; Song, Keun Woo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Development of the Mixed Oxide(MOX) fuel fabrication technology was reviewed in this study. Firstly, the feasibility of Pu utilization for nuclear fuel was analyzed by comparison of nuclear characteristics between U and Pu. Secondly, the feature and problem of processes developed so far was revealed and analyzed by reviewing each process in terms of technical difficulties and in connection with the pellet characteristics. Also, fabrication facilities currently existing were analyzed to understand particularities and circumstances in view of Pu handling, and finally, in-reactor behaviors of MOX fuel was compared with those of U fuel to understand how the Pu has an effect on fuel was compared with those of U fuel to understand how the Pu has an effect on fuel pellet structure and fuel rod. 73 figs., 15 tabs., 58 refs. (Author).

  18. Development, Fabrication and Characterization of Fuels for Indian Fast Reactor Programme

    International Nuclear Information System (INIS)

    Kumar, Arun

    2013-01-01

    Development of Fast Reactor fuels in India started in early Seventies. The successful development of Mixed Carbide fuels for FBTR and MOX fuel for PFBR have given confidence in manufacture of fuels for Fast Reactors. Effort is being put to develop high Breeding Ratio Metallic fuel (binary/ternary). Few fuel pins have been fabricated and is under test irradiation. However, this is only a beginning and complete fuel cycle activities are under development. Metal fuelled Fast Reactors will provide high growth rate in Indian Fast Reactor programme

  19. Fabrication of HTTR first loading fuel

    International Nuclear Information System (INIS)

    Kato, S.; Yoshimuta, S.; Hasumi, T.; Sato, K.; Sawa, K.; Suzuki, S.; Mogi, H.; Shiozawa, S.; Tanaka, T.

    2001-01-01

    This paper summarizes the fabrication of the first loading fuel for HTTR, High Temperature engineering Test Reactor constructed by JAERI, Japan Atomic Energy Research Institute. The fuel fabrication started at the HTR fuel facility of NFI, Nuclear Fuel Industries, Ltd., June 1995. 4,770 fuel rods were fabricated through the fuel kernel, coated fuel particle and fuel compaction process, then 150 fuel elements were assembled in the reactor building December 1997. Fabrication technology for the fuel was established through a lot of R and D activities and fabrication experience of irradiation examination samples spread over about 30 years. Most of all, very high quality and production efficiency of fuel were achieved by the development of the fuel kernel process using the vibration dropping technology, the continuous 4-layer coating process and the automatic compaction process. As for the inspection technology, the development of the automatic measurement equipment for coated layer thickness of a coated fuel particle and uranium content of a fuel compact contributed to the higher reliability and rationalization of the inspection process. The data processing system for the fabrication and quality control, which was originally developed by NFI, made possible not only quick feedback of statistical quality data to the fabrication processes, but also automatic document preparation, such as inspection certificates and accountability control reports. The quality of the first loading fuel fully satisfied the design specifications for the fuel. In particular, average bare uranium fraction and SiC defective fraction of fuel compacts were 2x10 -6 and 8x10 -5 , respectively. According to the preceding irradiation examinations being performed at JMTR, Japan Materials Testing Reactor of JAERI, the specimen sampled from the first loading fuel shows good irradiation performance. (author)

  20. Development of remote equipment for a DUPIC fuel fabrication at KAERI

    International Nuclear Information System (INIS)

    Lee, Jungwon; Kim, Kiho; Park, Geunil; Yang, Myungseung; Song, Keechan

    2007-01-01

    The DUPIC (Direct Use of spent PWR fuel In CANDU reactors) technology is to directly refabricate CANDU fuel from spent PWR fuel without any separation of the fissile materials and fission products. Thus, the DUPIC fuel material always remains in a highly radioactive state, which requires a remote fuel fabrication in a hot-cell. About 25 pieces of remote equipment including auxiliary systems such as a hot-cell shield plug were developed and installed in a hot cell. In order to supply a high electric current to a sintering furnace in-cell from an outside cell, a shield plug was developed. It consists of three components - a steel shield plug with an embedded spiral cooling line, stepped copper bus bars, and a shielding lead block. Experiments to evaluate the performance of the sintering furnace with the developed shield plug were carried out. It was concluded that, from the experimental results, the newly developed hot-cell shield plug satisfied all the requirements for a remote operation on a sintering furnace. DUPIC fuel pellets and elements were successfully fabricated with the developed remote equipment. (authors)

  1. Property-process relationships in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Tikare, V.

    2015-01-01

    Nuclear fuels are fabricated using many different techniques as they come in a large variety of shapes and compositions. The design and composition of nuclear fuels are predominantly dictated by the engineering requirements necessary for their function in reactors of various designs. Other engineering properties requirements originate from safety and security concerns, and the easy of handling, storing, transporting and disposing of the radioactive materials. In this chapter, the more common of these fuels will be briefly reviewed and the methods used to fabricate them will be presented. The fuels considered in this paper are oxide fuels used in LWRs and FRs, metal fuels in FRs and particulate fuels used in HTGRs. Fabrication of alternative fuel forms and use of standard fuels in alternative reactors will be discussed briefly. The primary motivation to advance fuel fabrication is to improve performance, reduce cost, reduce waste or enhance safety and security of the fuels. To achieve optimal performance, developing models to advance fuel fabrication has to be done in concert with developing fuel performance models. The specific properties and microstructures necessary for improved fuel performance must be identified using fuel performance models, while fuel fabrication models that can determine processing variables to give the desired microstructure and materials properties must be developed. (author)

  2. Development of fabrication technology for ceramic nuclear fuel

    International Nuclear Information System (INIS)

    Lee, Young Woo; Sohn, D. S.; Na, S. H.

    2003-05-01

    The purpose of the study is to develop the fabrication technology of MOX fuel. The researches carried out during the last stage(1997. 4.∼2003. 3.) mainly consisted of ; study of MOX pellet fabrication technology for application and development of characterization technology for the aim of confirming the development of powder treatment technology and sintering technology and of the optimization of the above technologies and fabrication of Pu-MOX pellet specimens through an international joint collaboration between KAERI and PSI based on the fundamental technologies developed in KAERI. Based on the studies carried out and the results obtained during the last stage, more extensive studies for the process technologies of the unit processes were performed, in this year, for the purpose of development of indigenous overall MOX pellet fabrication process technology, relating process parameters among the unit processes and integrating these unit process technologies. Furthermore, for the preparation of transfer of relevant technologies to the industries, a feasibility study was performed on the commercialization of the technology developed in KAERI with the relevant industry in close collaboration

  3. Development of CANDU high-burnup fuel fabrication technology

    International Nuclear Information System (INIS)

    Sim, Ki Seob; Suk, H. C.; Kwon, H. I.; Ji, C. G.; Cho, M. S.; Chang, H. I.

    1997-07-01

    This study is focused on the achievement of the fabrication process improvement of CANFLEX-NU and for this purpose, following two areas of basic research were executed this year. 1) development of amorphous alloy for use in brazing of nuclear materials. 2) development of ECT techniques for the end-cap weld inspection. Also, preliminary feasibility analyses on the characteristics and handling techniques of CANFLEX-RU fuel were executed this year. - Selection of optimum conversion process of RU power -Characterization of the composition of RU power - Radiological characterization of RU power and sintered pellets - Compaction and sintering characteristics of RU power - Required special process for the production of CANFLEX-RU fuel - Development of technical specification for RU powder and pellets. In addition, technical support activities were performed for in-pile and out-pile fuel performance tests such as precision measurement of out-pile test fuel dimensions, establishment of quality control technique on fuel bundle by providing bundle kits to AECL for use in-pile irradiation tests in the NRU research reactor. (author). 57 refs., 16 tabs.,40 figs

  4. Developments in MOX fuel pellet fabrication technology: Indian experience

    International Nuclear Information System (INIS)

    Kamath, H.S.; Majumdar, S.; Purusthotham, D.S.C.

    1998-01-01

    under glove box conditions. Pellets of different geometry, from simple cylindrical to chamfered, dished and annular pellets have been fabricated and irradiated in research reactors although plain cylindrical pellets with L/D less than 1.2 have been used for MOX fuel loading in power reactors. Fully automated wet centreless grinding of MOX pellets using composite diamond wheel and subsequent ultrasonic cleaning has been used in the fabrication flowsheet. The MOX pellets undergo vacuum degassing at 400 deg. C to ensure low hydrogen content prior to loading of pellets into zircaloy clad fuel tubes. A novel sol-gel microsphere pelletisation route (SGMP) combined with LTS has also been developed and is briefly discussed. (author)

  5. Nuclear fuel fabrication in India

    International Nuclear Information System (INIS)

    Kondal Rao, N.

    1975-01-01

    The important role of a nuclear power programme in meeting the growing needs of power in India is explained. The successful installation of Tarapur Atomic Power Station and Rajasthan Atomic Power Station as well as the work at Madras Atomic Power Station are described. The development of the Atomic Fuels Division and the Nuclear Fuel Complex, Hyderabad which is mainly concerned with the fabrication of fuel elements and the reprocessing of fuels are explained. The N.F.C. essentially has the following constituent units : Zirconium Plant (ZP) comprising of Zirconium Oxide Plant, Zirconium Sponge Plant and Zirconium Fabrication Plant; Natural Uranium Oxide Plant (UOP); Ceramic Fuel Fabrication Plant (CFFP); Enriched Uranium Oxide Plant (EUOP); Enriched Fuel Fabrication Plant (EEFP) and Quality Control Laboratory for meeting the quality control requirements of all plants. The capacities of various plants at the NFC are mentioned. The work done on mixed oxide fuels and FBTR core with blanket assemblies, nickel and steel assemblies, thermal research reactor of 100 MW capacity, etc. are briefly mentioned. (K.B.)

  6. Nuclear fuel fabrication in India

    Energy Technology Data Exchange (ETDEWEB)

    Kondal Rao, N

    1975-01-01

    The important role of a nuclear power program in meeting the growing needs of power in India is explained. The successful installation of Tarapur Atomic Power Station and Rajasthan Atomic Power Station as well as the work at Madras Atomic Power Station are described. The development of the Atomic Fuels Division and the Nuclear Fuel Complex, Hyderabad which is mainly concerned with the fabrication of fuel elements and the reprocessing of fuels are explained. The N.F.C. essentially has the following constituent units : Zirconium Plant (ZP) comprising of Zirconium Oxide Plant, Zirconium Sponge Plant and Zirconium Fabrication Plant; Natural Uranium Oxide Plant (UOP); Ceramic Fuel Fabrication Plant (CFFP); Enriched Uranium Oxide Plant (EUOP); Enriched Fuel Fabrication Plant (EEFP) and Quality Control Laboratory for meeting the quality control requirements of all plants. The capacities of various plants at the NFC are mentioned. The work done on mixed oxide fuels and FBTR core with blanket assemblies, nickel and steel assemblies, thermal research reactor of 100 MW capacity, etc. are briefly mentioned.

  7. Development of the fabrication technology of the simulated DUPIC fuel

    International Nuclear Information System (INIS)

    Kang, Kweon Ho; Yang, M. S.; Bae, K. K. and others

    2000-06-01

    It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties of the DUPIC fuel is different from the commercial UO 2 fuel. But what directly measures physical properties et al. of DUPIC fuel being resinterred simulated spent fuel through OREOX process is very difficult in laboratory owing to its high level radiation. Then fabrication of simulated DUPIC fuel is needed to measure its properties. In this study, processes on powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using simulated spent fuel are discribed. To fabricate simulated DUPIC fuel, the powder from 3 times OREOX and 5 times attrition milling simulated spent fuel is compacted with 1.3 ton/cm 2 . Pellets are sintered in 100% H 2 atmosphere over 10 h at 1800 deg C. Sintered densities of pellets are 10.2-10.5 g/cm 3

  8. Overview of MOX fuel fabrication achievements

    International Nuclear Information System (INIS)

    Bairiot, H.; Vliet, J. van; Chiarelli, G.; Edwards, J.; Nagai, S.H.; Reshetnikov, F.

    2000-01-01

    Such overview having been adequately covered in an OECD/NEA publication providing the situation as of end 1994, this paper is mainly devoted to an update as of end 1998. The Belgian plant, Belgonucleaire/Dessel, is now dedicated exclusively to the fabrication of MOX fuel and has operated consistently around its nameplate capacity (35tHM/a) through the 1990s involving a large variety of PWR and BWR fuels. The two French plants have also achieved routine operation during the 1990s. CFCa, historically the largest FBR MOX fuel manufacturer, is utilizing the genuine COCA process for that type of fuel and the MIMAS process for LWR fuel: a nominal capacity (40 tHM/a) has been gradually approached. MELOX has operated at 100 tHM/a, as defined in the operating licence granted originally. The British plant, MDF/Sellafield with 8tHM/a nameplate capacity is devoted to fuel and has manufactured several small fabrication campaigns. In Japan, JNC operates three facilities located at Tokai: PFDF, devoted to basic research and fabrication of test fuels, PFFF/ATR line, for the fabrication of Fugen fuel and of corresponding fuel for the critical facility DCA, and PFPF for the fabrication of FBR fuel. In Russia, fabrication techniques have been developed to fuel four BN-800 FBRs contemplated to be constructed and be fuelled with the civilian Pu stockpile. Two demonstration facilities Paket (Mayak) and RIAR (Dimitrovgrad) fabricated respectively pellet and vipac type FBR MOX fuel for BR-5, BOR-60, BN-350 and BN-600. The paper includes a brief description of each of the fabrication routes mentioned, as well as the production of respectively LWR and FBR MOX fuel in each fabrication facility, since the start-up of the plant, since 1 January 1993 and since 1 January 1998 up to 31 December 1998. (author)

  9. Oxide fuel fabrication technology development of the FaCT project (1). Overall review of fuel technology development of the FaCT project

    International Nuclear Information System (INIS)

    Abe, Tomoyuki; Namekawa, Takashi; Tanaka, Kenya

    2011-01-01

    The FaCT project is in progress in Japan for the commercialization of fast reactor cycle system. The development goal of the fuel in the FaCT project is a low-decontaminated TRU homo-recycling in a closed cycle and extension in average discharge burn-up to 150 GWd/t. Research and development on innovative technologies concerning the short process, remote maintenance and cooling system of automatic fuel production equipments, long life cladding material and control of oxygen potential have been conducted in phase I of the FaCT project. As the result of various test including 600 g batch MOX tests, it is concluded that the short process is available to fuel pellet fabrication of the FaCT project. Although cold mock-up tests on test model of some typical process equipments suggest possibilities of remote maintenance of automatic fuel fabrication equipment, it is concluded that it still needs further efforts to judge the operability of the completely remote fabrication for low-decontaminated TRU fuel. A cold mock-up test on fuel pin assembling equipment show that influence of decay heat of MA can be managed by cooling system. Irradiation tests in BOR-60 indicate that 9Cr-ODS possess the satisfactory in-reactor performance as the long life cladding material if homogeneity of alloy element is adequately controlled. Modification of cladding tube fabrication process to ensure homogeneity and further development of measures to control oxygen potential inside the fuel pin are necessary to reach the burn-up target of the FaCT project. (author)

  10. Development of ISA procedure for uranium fuel fabrication and enrichment facilities

    International Nuclear Information System (INIS)

    Yamate, Kazuki; Arakawa, Tomoyuki; Yamashita, Masahiro; Sasaki, Noriaki; Hirano, Mitsumasa

    2011-01-01

    The integrated safety analysis (ISA) procedure has been developed to apply risk-informed regulation to uranium fuel fabrication and enrichment facilities. The major development efforts are as follows: (a) preparing the risk level matrix as an index for items-relied-on-for-safety (IROFS) identification, (b) defining requirements of IROFS, and (c) determining methods of IROFS importance based on the results of risk- and scenario-based analyses. For the risk level matrix, the consequence and likelihood categories have been defined by taking into account the Japanese regulatory laws, rules, and safety standards. The trial analyses using the developed procedure have been performed for several representative processes of the reference uranium fuel fabrication and enrichment facilities. This paper presents the results of the ISA for the sintering process of the reference fabrication facility. The results of the trial analyses have demonstrated the applicability of the procedure to the risk-informed regulation of these facilities. (author)

  11. Full size U-10Mo monolithic fuel foil and fuel plate fabrication-technology development

    International Nuclear Information System (INIS)

    Moore, G.A.; Jue, J-F.; Rabin, B.H.; Nilles, M.J.

    2010-01-01

    Full-size U-10Mo foils are being developed for use in high density LEU monolithic fuel plates. The application of a zirconium barrier layer to the foil is performed using a hot co-rolling process. Aluminium clad fuel plates are fabricated using Hot Isostatic Pressing (HIP) or a Friction Bonding (FB) process. An overview is provided of ongoing technology development activities, including: the co-rolling process, foil shearing/slitting and polishing, cladding bonding processes, plate forming, plate-assembly swaging, and fuel plate characterization. Characterization techniques being employed include, Ultrasonic Testing (UT), radiography, and microscopy. (author)

  12. LEU fuel fabrication in Argentina

    International Nuclear Information System (INIS)

    Giorsetti, D.R.; Gomez, J.O.; Marajofsky, A.; Kohut, C.

    1985-01-01

    As an Institution, aiming to meet with its own needs, CNEA has been intensively developing reduced enriched fuel to use in its own research and test reactors. Development of the fabrication technology as well as the design, installation and operation of the manufacturing plant, have been carried out with its own funds. Irradiation and post-irradiation of test miniplates have been taking place within the framework of the RERTR program. During the last years, CNEA has developed three LEU fuel types. In the previous RERTR meetings, we presented the technological results obtained with these fuel types. This paper focuses on CNEA LEU fuel element manufacturing status and the trained personnel we can offer in design and manufacture fuel capability. CNEA has its own fuel manufacturing technology; the necessary facilities to start the fuel fabrication; qualified technicians and professionals for: fuel design and behaviour analysis; fuel manufacturing and QA; international recognition of its fuel development and manufacturing capability through its ORR miniplate irradiation; its own natural uranium and the future possibility to enrich up to 20% U 235 ; the probability to offer a competitive fuel manufacturing cost in the international market; the disposition to cooperate with all countries that wish to take part and aim to reach an self-sufficiency in their own fuel supply needs

  13. Development of IAEA safeguards at low enrichment uranium fuel fabrication plants

    International Nuclear Information System (INIS)

    Badawy, I.

    1988-01-01

    In this report the nuclear material at low enrichment uranium fuel fabrication plants under IAEA safeguards is studied. The current verification practices of the nuclear material and future improvements are also considered. The problems met during the implementation of the the verification measures of the nuclear material - particularly for the fuel assemblies are discussed. The additional verification activities as proposed for future improvements are also discussed including the physical inventory verification and the verification of receipts and shipments. It is concluded that the future development of the present IAEA verification practices at low enrichment uranium fuel fabrication plants would necessitate the application of quantitative measures of the nuclear material and the implementation of advanced measurement techniques and instruments. 2 fig., 4 tab

  14. Impact of fuel fabrication and fuel management technologies on uranium management

    International Nuclear Information System (INIS)

    Arnsberger, P.L.; Stucker, D.L.

    1994-01-01

    Uranium utilization in commercial pressurized water reactors is a complex function of original NSSS design, utility energy requirements, fuel assembly design, fuel fabrication materials and fuel fabrication materials and fuel management optimization. Fuel design and fabrication technologies have reacted to the resulting market forcing functions with a combination of design and material changes. The technologies employed have included ever-increasing fuel discharge burnup, non-parasitic structural materials, burnable absorbers, and fissile material core zoning schemes (both in the axial and radial direction). The result of these technological advances has improved uranium utilization by roughly sixty percent from the infancy days of nuclear power to present fuel management. Fuel management optimization technologies have also been developed in recent years which provide fuel utilization improvements due to core loading pattern optimization. This paper describes the development and impact of technology advances upon uranium utilization in modern pressurized water reactors. 10 refs., 3 tabs., 10 figs

  15. Development of DIPRES feed for the fabrication of mixed-oxide fuels for fast breeder reactors

    International Nuclear Information System (INIS)

    Griffin, C.W.; Rasmussen, D.E.; Lloyd, M.H.

    1983-01-01

    The DIrect PREss Spheroidized feed process combines the conversion of uranium-plutonium solutions into spheres by internal gelation with conventional pellet fabrication techniques. In this manner, gel spheres could replace conventional powders as the feed material for pellet fabrication of nuclear fuels. Objective of the DIPRES feed program is to develop and qualify a process to produce mixed-oxide fuel pellets from gel spheres for fast breeder reactors. This process development includes both conversion and fabrication activities

  16. Report on fabrication of pin components for fuel fabrication in FUJI project (Co-operation in the research and development of advanced sphere-pac fuel among PSI, JNC, and NRG)

    International Nuclear Information System (INIS)

    Suzuki, Masahiro; Hinai, Hiroshi; Shigetome, Yoshiaki; Kono, Shusaku; Matsuzaki, Masaaki

    2003-03-01

    Japan Nuclear Cycle Development Institute (JNC) has conducted the co-operation concerning vibro-packed fuels with Paul Scherrer Institut (PSI) in Switzerland and Nuclear Research and consultancy Group (NRG) in the Netherlands. The project 'Research and Development of advanced Sphere-pac Fuel' is called FUJI (FUel irradiations for JNC and PSI) Project. In this project, three types of fuels that are sphere-pac fuels, vipac fuels, and pellet fuels will be irradiated in the High Flux Reactor (HFR) to compare their performance. Based on the drawing which has been agreed among three parties, fabrication of the pin components and welding of the upper and lower connection end plugs were performed in accordance with ISO9001 in JNC. This report describes data of the fabricated pin components, results of welding qualification tests, and quality assurance of the welded components. The fabrication of pin components was successfully completed and they were delivered to PSI in October 2002. (author)

  17. A review on the development of the advanced fuel fabrication technology

    International Nuclear Information System (INIS)

    Lee, Jung Won; Lee, Yung Woo; Sohn, Dong Sung; Yang, Myung Seung; Bae, Kee Kwang; Nah, Sang Hoh; Kim, Han Soo; Kim, Bong Koo; Song, Keun Woo; Kim, See Hyung

    1995-07-01

    In this state-of art report, the development status of the advanced nuclear fuel was investigated. The current fabrication technology for coated particle fuel and non-oxide fuel such as sol-gel technology, coating technology, and carbothermic reduction reaction has also been examined. In the view point of inherent safety and efficiency in the operation of power plant, the coated particle fuel will keep going on its reputation as nuclear fuel for a high temperature gas cooled reactor, and the nitride fuel is very prospective for the next liquid metal fast breeder reactor. 43 figs., 17 tabs., 96 refs. (Author)

  18. A review on the development of the advanced fuel fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Lee, Yung Woo; Sohn, Dong Sung; Yang, Myung Seung; Bae, Kee Kwang; Nah, Sang Hoh; Kim, Han Soo; Kim, Bong Koo; Song, Keun Woo; Kim, See Hyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    In this state-of art report, the development status of the advanced nuclear fuel was investigated. The current fabrication technology for coated particle fuel and non-oxide fuel such as sol-gel technology, coating technology, and carbothermic reduction reaction has also been examined. In the view point of inherent safety and efficiency in the operation of power plant, the coated particle fuel will keep going on its reputation as nuclear fuel for a high temperature gas cooled reactor, and the nitride fuel is very prospective for the next liquid metal fast breeder reactor. 43 figs., 17 tabs., 96 refs. (Author).

  19. Alternative Fabrication of Recycling Fast Reactor Metal Fuel

    International Nuclear Information System (INIS)

    Kim, Ki-Hwan; Kim, Jong Hwan; Song, Hoon; Kim, Hyung-Tae; Lee, Chan-Bock

    2015-01-01

    Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. In order to develop innovative fabrication method of metal fuel for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, metal fuel slugs for SFR have been fabricated by modified casting method, and characterized to evaluate the feasibility of the alternative fabrication method. In order to prevent evaporation of volatile elements such as Am and improve quality of fuel slugs, alternative fabrication methods of metal fuel slugs have been studied in KAERI. U-10Zr-5Mn fuel slug containing volatile surrogate element Mn was soundly cast by modified injection casting under modest pressure. Evaporation of Mn during alternative casting could not be detected by chemical analysis. Mn element was most recovered with prevention of evaporation by alternative casting. Modified injection casting has been selected as an alternative fabrication method in KAERI, considering evaporation prevention, and proven benefits of high productivity, high yield, and good remote control

  20. Fabrication of preliminary fuel rods for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

    2012-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  1. Advanced high throughput MOX fuel fabrication technology and sustainable development

    International Nuclear Information System (INIS)

    Krellmann, Juergen

    2005-01-01

    The MELOX plant in the south of France together with the La Hague reprocessing plant, are part of the two industrial facilities in charge of closing the nuclear fuel cycle in France. Started up in 1995, MELOX has since accumulated a solid know-how in recycling plutonium recovered from spent uranium fuel into MOX: a fuel blend comprised of both uranium and plutonium oxides. Converting recovered Pu into a proliferation-resistant material that can readily be used to power a civil nuclear reactor, MOX fabrication offers a sustainable solution to safely take advantage of the plutonium's high energy content. Being the first large-capacity industrial facility dedicated to MOX fuel fabrication, MELOX distinguishes itself from the first generation MOX plants with high capacity (around 200 tHM versus around 40 tHM) and several unique operational features designed to improve productivity, reliability and flexibility while maintaining high safety standards. Providing an exemplary reference for high throughput MOX fabrication with 1,000 tHM produced since start-up, the unique process and technologies implemented at MELOX are currently inspiring other MOX plant construction projects (in Japan with the J-MOX plant, in the US and in Russia as part of the weapon-grade plutonium inventory reduction). Spurred by the growing international demand, MELOX has embarked upon an ambitious production development and diversification plan. Starting from an annual level of 100 tons of heavy metal (tHM), MELOX demonstrated production capacity is continuously increasing: MELOX is now aiming for a minimum of 140 tHM by the end of 2005, with the ultimate ambition of reaching the full capacity of the plant (around 200 tHM) in the near future. With regards to its activity, MELOX also remains deeply committed to sustainable development in a consolidated involvement within AREVA group. The French minister of Industry, on August 26th 2005, acknowledged the benefits of MOX fuel production at MELOX: 'In

  2. Fabrication of particulate metal fuel for fast burner reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Lee, Sun Yong; Kim, Jong Hwan; Woo, Yoon Myung; Ko, Young Mo; Kim, Ki Hwan; Park, Jong Man; Lee, Chan Bok

    2012-01-01

    U Zr metallic fuel for sodium cooled fast reactors is now being developed by KAERI as a national R and D program of Korea. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. Therefore, innovative fuel concepts should be developed to address the fabrication challenges pertaining to TRU while maintaining good performances of metallic fuel. Particulate fuel concepts have already been proposed and tested at several experimental fast reactor systems and vipac ceramic fuel of RIAR, Russia is one of the examples. However, much less work has been reported for particulate metallic fuel development. Spherical uranium alloy particles with various diameters can be easily produced by the centrifugal atomization technique developed by KAERI. Using the atomized uranium and uranium zirconium alloy particles, we fabricated various kinds of powder pack, powder compacts and sintered pellets. The microstructures and properties of the powder pack and pellets are presented

  3. Technical study report on fuel fabrication system

    International Nuclear Information System (INIS)

    Kono, Shusaku; Tanaka, Kenya; Ono, Kiyoshi; Iwasa, Katsuyoshi; Hoshino, Yasushi; Shinkai, Yasuo

    2000-07-01

    The feasibility study of FBR and related fuel cycle is performed for developing the FBR recycle system which ensures safety, economic competitiveness, efficient utilization of resources, reduction of environmental burden and enhancement of nuclear non-proliferation under consistency of FBR reactor and fuel cycle systems. In this study, a conceptual design study and system characteristics evaluation are conducted for fuel fabrication systems of pellet process, vibropack process for oxide and nitride fuel and casting process for metal fuel. Technical issues in each process are also extracted. In 1999 fiscal year, a conceptual design study were conducted for the fuel fabrication plants adopting (1) the short pellet process which simplifies the conventional MOX pellet fabrication processes, (2) vibropack processes of aqueous gelation process, improved RIAR process, improved ANL process and fluoride volatility process, (3) casting processes of injection process, centrifuging process. As a result, attainable perspective was obtained for each fuel fabrication system through the evaluation of apparatuses, layout and facility volume, etc. In each fuel fabrication system, technical issues for practical use were made clear. Hereafter, more detailed study will be performed for each system, and research programs for phase II study will be planned. (author)

  4. Development of advanced spent fuel management process. The fabrication and oxidation behavior of simulated metallized spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ro, Seung Gy; Shin, Y.J.; You, G.S.; Joo, J.S.; Min, D.K.; Chun, Y.B.; Lee, E.P.; Seo, H.S.; Ahn, S.B

    1999-03-01

    The simulated metallized spent fuel ingots were fabricated and evaluated the oxidation rates and the activation energies under several temperature conditions to develop an advanced spent fuel management process. It was also checked the alloying characteristics of the some elements with metal uranium. (Author). 3 refs., 1 tab., 36 figs.

  5. MOX fuel fabrication at AECL

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Jeffs, A.T.

    1995-01-01

    Atomic Energy of Canada Limited's mixed-oxide (MOX) fuel fabrication activities are conducted in the Recycle Fuel Fabrication Laboratories (RFFL) at the Chalk River Laboratories. The RFFL facility is designed to produce experimental quantities of CANDU MOX fuel for reactor physics tests or demonstration irradiations. From 1979 to 1987, several MOX fuel fabrication campaigns were run in the RFFL, producing various quantities of fuel with different compositions. About 150 bundles, containing over three tonnes of MOX, were fabricated in the RFFL before operations in the facility were suspended. In late 1987, the RFFL was placed in a state of active standby, a condition where no fuel fabrication activities are conducted, but the monitoring and ventilation systems in the facility are maintained. Currently, a project to rehabilitate the RFFL and resume MOX fuel fabrication is nearing completion. This project is funded by the CANDU Owners' Group (COG). The initial fabrication campaign will consist of the production of thirty-eight 37-element (U,Pu)O 2 bundles containing 0.2 wt% Pu in Heavy Element (H.E.) destined for physics tests in the zero-power ZED-2 reactor. An overview of the Rehabilitation Project will be given. (author)

  6. LOFT fuel modules design, characterization, and fabrication program

    International Nuclear Information System (INIS)

    Russell, M.L.

    1977-06-01

    The loss-of-fluid test [LOFT) fuel modules have evolved from a comprehensive five-year design, characterization, and fabrication program which has resulted in the accomplishment of many technical activities of interest in pressurized water reactor fuel design development and safety research. Information is presented concerning: determination of fundamental high-temperature reactor material properties; design invention related to in-core instrumentation attachment; implementation of advanced and/or unique fuel bundle characterization techniques; implementation of improved fuel bundle fabrication techniques; and planning and execution of a multimillion dollar design, characterization, and fabrication program for pressurized water reactor fuel

  7. Development of advanced fabrication technology for high-temperature gas-cooled reactor fuel. Reduction of coating failure fraction

    International Nuclear Information System (INIS)

    Minato, Kazuo; Kikuchi, Hironobu; Fukuda, Kousaku; Tobita, Tsutomu; Yoshimuta, Sigeharu; Suzuki, Nobuyuki; Tomimoto, Hiroshi; Nishimura, Kazuhisa; Oda, Takafumi

    1998-11-01

    The advanced fabrication technology for high-temperature gas-cooled reactor fuel has been developed to reduce the coating failure fraction of the fuel particles, which leads to an improvement of the reactor safety. The present report reviews the results of the relevant work. The mechanisms of the coating failure of the fuel particles during coating and compaction processes of the fuel fabrication were studied to determine a way to reduce the coating failure fraction of the fuel. The coating process was improved by optimizing the mode of the particle fluidization and by developing the process without unloading and loading of the particles at intermediate coating process. The compaction process was improved by optimizing the combination of the pressing temperature and the pressing speed of the overcoated particles. Through these modifications of the fabrication process, the quality of the fuel was improved outstandingly. (author)

  8. Automated breeder fuel fabrication

    International Nuclear Information System (INIS)

    Goldmann, L.H.; Frederickson, J.R.

    1983-01-01

    The objective of the Secure Automated Fabrication (SAF) Project is to develop remotely operated equipment for the processing and manufacturing of breeder reactor fuel pins. The SAF line will be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at the Department of Energy's (DOE) Hanford site near Richland, Washington, and is operated by the Westinghouse Hanford Company (WHC). The fabrication and support systems of the SAF line are designed for computer-controlled operation from a centralized control room. Remote and automated fuel fabriction operations will result in: reduced radiation exposure to workers; enhanced safeguards; improved product quality; near real-time accountability, and increased productivity. The present schedule calls for installation of SAF line equipment in the FMEF beginning in 1984, with qualifying runs starting in 1986 and production commencing in 1987. 5 figures

  9. Recent Progress on the DUPIC Fuel Fabrication Technology at KAERI

    International Nuclear Information System (INIS)

    Jung-Won Lee; Ho-Jin Ryu; Geun-Il Park; Kee-Chan Song

    2008-01-01

    Since 1991, KAERI has been developing the DUPIC fuel cycle technology. The concept of a direct use of spent PWR fuel in Candu reactors (DUPIC) is based on a dry processing method to re-fabricate Candu fuel from spent PWR fuel without any intentional separation of the fissile materials and fission products. A DUPIC fuel pellet was successfully fabricated and the DUPIC fuel element fabrication processes were qualified on the basis of a Quality Assurance program. Consequently, the DUPIC fuel fabrication technology was verified and demonstrated on a laboratory-scale. Recently, the fuel discharge burn-up of PWRs has been extended to reduce the amount of spent fuel and the fuel cycle costs. Considering this trend of extending the fuel burn-up in PWRs, the DUPIC fuel fabrication technology should be improved to process high burn-up spent fuels. Particularly the release behavior of cesium from the pellet prepared with a high burn-up spent fuel was assessed. an improved DUPIC fuel fabrication technology was experimentally established with a fuel burn-up of 65,000 MWd/tU. (authors)

  10. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

    1980-01-01

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  11. U.S. technology for mechanized/automated fabrication of fast reactor fuel

    International Nuclear Information System (INIS)

    Nyman, D.H.; Bennett, D.W.; Claudson, T.T.; Dahl, R.E.; Graham, R.A.; Keating, J.J.; Yatabe, J.M.

    1978-01-01

    The status of the U.S. fast reactor Fuel Fabrication Development Program is discussed. The objectives of the program are to develop and evaluate a high throughput pilot fuel fabrication line including close-coupled chemistry and wet scrap recycle operations. The goals of the program are to demonstrate by mechanized/automated and remote processes: reduced personnel exposure, enhanced safegurads/accountability, improved fuel performance, representative fabrication rates and reduced fuel costs

  12. Description of ECRI (CNEA'S MTR fuel fabrication plant)

    International Nuclear Information System (INIS)

    Echenique, P.; Fabro, J.; Podesta, D.; Restelli, M.; Rossi, G.; Alvarez, L.; Adelfang, P.

    2002-01-01

    The ECRI Plant is dedicated to the development and fabrication of high-density fuel elements and targets for 99 Mo. In this sector had been done the start up Fuel Elements for the Reactors of Peru, Iran, Algeria and Egypt. All of them were made with U 3 O 8 . The targets for 99 Mo using HEU were fabricated too in the last years. The new material of high-density for Fuel Elements as U 3 Si 2 were done in this sector, three prototypes were fabricated, two are still under irradiation. (P06 and P07). As new developments we are working with U-Mo (7%) Fuel Plates with both material Korean and HMD. This work is under the RERTR Program and two fuel elements, manufactured by us, with both powders, will be irradiated in Petten. For 99 Mo targets, we are fabricating miniplates of LEU with an AlUx powder by pulvi-metallurgy technique. And it is under development the foils targets under the RERTR Program. A general view of the fabrication facilities and control sector will be shown. The different operations that are done in each sector will be explained. All our activities will be certified under the ISO 9000 and we are working hard to get it in the middle of 2003. (author)

  13. International light water nuclear fuel fabrication supply. Are fabrication services assured?

    International Nuclear Information System (INIS)

    Rothwell, Geoffrey

    2010-01-01

    This paper examines the cost structure of fabricating light water reactor (LWR) fuel with low-enriched uranium (LEU, with less than 5% enrichment). The LWR-LEU fuel industry is decades old, and (except for the high entry cost of designing and licensing a fuel fabrication facility and its fuel), labor and additional fabrication lines can be added at Nth-of-a-Kind cost to the maximum capacity allowed by a site license. The industry appears to be competitive: nuclear fuel fabrication capacity is assured with many competitors and reasonable prices. However, nuclear fuel assurance has become an important issue for nations now to considering new nuclear power plants. To provide this assurance many proposals equate 'nuclear fuel banks' (which would require fuel for specific reactors) with 'LEU banks' (where LEU could be blended into nuclear fuel with the proper enrichment) with local fuel fabrication. The policy issues (which are presented, but not answered in this paper) become (1) whether the construction of new nuclear fuel fabrication facilities in new nuclear power nations could lead to the proliferation of nuclear weapons, and (2) whether nuclear fuel quality can be guaranteed under current industry arrangements, given that fuel failure at one reactor can lead to forced shutdowns at many others. (author)

  14. Fabrication of MOX fuel element clusters for irradiation in PWL, CIRUS

    International Nuclear Information System (INIS)

    Roy, P.R.; Purushotham, D.S.C.; Majumdar, S.

    1983-01-01

    Three clusters, each containing 6 zircaloy-2 clad short length fuel elements of either MOX or UO 2 fuel pellets were fabricated for irradiation in pressurized water loop of CIRUS. The major objectives of the programme were: (a) to optimize the various fabrication parameters for developing a flow sheet for MOX fuel element fabrication; (b) to study the performance of the MOX fuel elements at a peak heat flux of 110 W/cm 2 ; and (c) to study the effect of various fuel pellet design changes on the behaviour of the fuel element under irradiation. Two clusters, one each of UO 2 and MOX, have been successfully irradiated to the required burn-up level and are now awaiting post irradiation examinations. The third MOX cluster is still undergoing irradiation. Fabrication of these fuel elements involved considerable amount of developing work related to the fabrication of the MOX fuel pellets and the element welding technique and is reported in detail in this report. (author)

  15. Coated U(Mo) Fuel: As-Fabricated Microstructures

    Energy Technology Data Exchange (ETDEWEB)

    Emmanuel Perez; Dennis D. Keiser, Jr.; Ann Leenaers; Sven Van den Berghe; Tom Wiencek

    2014-04-01

    As part of the development of low-enriched uranium fuels, fuel plates have recently been tested in the BR-2 reactor as part of the SELENIUM experiment. These fuel plates contained fuel particles with either Si or ZrN thin film coating (up to 1 µm thickness) around the U-7Mo fuel particles. In order to best understand irradiation performance, it is important to determine the starting microstructure that can be observed in as-fabricated fuel plates. To this end, detailed microstructural characterization was performed on ZrN and Si-coated U-7Mo powder in samples taken from AA6061-clad fuel plates fabricated at 500°C. Of interest was the condition of the thin film coatings after fabrication at a relatively high temperature. Both scanning electron microscopy and transmission electron microscopy were employed. The ZrN thin film coating was observed to consist of columns comprised of very fine ZrN grains. Relatively large amounts of porosity could be found in some areas of the thin film, along with an enrichment of oxygen around each of the the ZrN columns. In the case of the pure Si thin film coating sample, a (U,Mo,Al,Si) interaction layer was observed around the U-7Mo particles. Apparently, the Si reacted with the U-7Mo and Al matrix during fuel plate fabrication at 500°C to form this layer. The microstructure of the formed layer is very similar to those that form in U-7Mo versus Al-Si alloy diffusion couples annealed at higher temperatures and as-fabricated U-7Mo dispersion fuel plates with Al-Si alloy matrix fabricated at 500°C.

  16. Improvements in fabrication of metallic fuels

    International Nuclear Information System (INIS)

    Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

    1989-12-01

    Argonne National Laboratory is currently developing a new liquid- metal cooled breeder reactor known as the Integral Fast Reactor (IFR). IFR fuels represent the state-of-the-art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, to be discussed below, will support the fully remote fuel cycle facility that as an integral part of the IFR concept will be demonstrated at the EBR-II site. 3 refs

  17. Fabrication of fuel elements interplay between typical SNR Mark Ia specifications and the fuel element fabrication

    International Nuclear Information System (INIS)

    Biermann, W.K.; Heuvel, H.J.; Pilate, S.; Vanderborck, Y.; Pelckmans, E.; Vanhellemont, G.; Roepenack, H.; Stoll, W.

    1987-01-01

    The core and fuel were designed for the SNR-300 first core by Interatom GmbH and Belgonucleaire. The fuel was fabricated by Alkem/RBU and Belgonucleaire. Based on the preparation of drawings and specifications and on the results of the prerun fabrication, an extensive interplay took place between design requirements, specifications, and fabrication processes at both fuel plants. During start-up of pellet and pin fabrication, this solved such technical questions as /sup 239/Pu equivalent linear weight, pellet density, stoichiometry of the pellets, and impurity content. Close cooperation of designers and manufacturers has allowed manufacture of 205 fuel assemblies without major problems

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

    International Nuclear Information System (INIS)

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

    1979-12-01

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

  19. Development and fabrication of seamless Aluminium finned clad tubes for metallic uranium fuel rods for research reactor

    International Nuclear Information System (INIS)

    Singh, A.K.; Hussain, M.M.; Jayachandran, N.K.; Abdulla, K.K.

    2012-01-01

    Natural uranium metal or its alloy is used as fuel in nuclear reactors. Usually fuel is clad with compatible material to prevent its direct contact with coolant which prevents spread of activity. One of the methods of producing fuel for nuclear reactor is by co-drawing finished uranium rods with aluminum clad tube to develop intimate contact for effective heat removal during reactor operation. Presently seam welded Aluminium tubes are used as clad for Research Reactor fuel. The paper will highlight entire fabrication process followed for the fabrication of seamless Aluminium finned tubes along with relevant characterisation results

  20. Development of inspection data collection and evaluation system for large scale MOX fuel fabrication plant safeguards (3)

    International Nuclear Information System (INIS)

    Kumakura, Shinichi; Masuda, Shoichiro; Iso, Shoko; Hisamatsu, Yoshinori; Kurobe, Hiroko; Nakajima, Shinji

    2015-01-01

    Inspection Data Collection and Evaluation System is the system to store inspection data and operator declaration data collected from various measurement equipment, which is installed in fuel fabrication processes of the large-scale MOX fuel fabrication plant, and to make safeguards evaluation based on Near Real Time Accountancy (NRTA) using these data. Nuclear Material Control Center developed the simulator to simulate fuel fabrication process, in-process material inventory/flow data and the measurement data and the adequacy/impact to the uncertainty of the material balance using the simulation results, such as the facility operation and the operational status, has been reviewed. Following the 34th INMM Japan chapter presentation, the model similar to the real nuclear material accountancy during the fuel fabrication process was simulated and the nuclear material accountancy and its uncertainty (Sigma MUF) have been reviewed. Some findings have been obtained, such as regarding evaluation related indicators for verification under a more realistic accountancy which could be applied by operator. (author)

  1. Remote fabrication of nuclear fuel: a secure automated fabrication overview

    International Nuclear Information System (INIS)

    Nyman, D.H.; Benson, E.M.; Yatabe, J.M.; Nagamoto, T.T.

    1981-01-01

    An automated line for the fabrication of breeder reactor fuel pins is being developed. The line will be installed in the Fuels and Materials Examination Facility (FMEF) presently under construction at the Hanford site near Richland, Washington. The application of automation and remote operations to fuel processing technology is needed to meet program requirements of reduced personnel exposure, enhanced safeguards, improved product quality, and increased productivity. Commercially available robots are being integrated into operations such as handling of radioactive material within a process operation. These and other automated equipment and chemistry analyses systems under development are described

  2. Fabrication of Fast Reactor Fuel Pins for Test Irradiations

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-09-15

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  4. Secure Automated Fabrication: an overview of remote breeder fuel fabrication

    International Nuclear Information System (INIS)

    Nyman, D.H.; Graham, R.A.

    1983-10-01

    The Secure Automated Fabrication (SAF) line is an automated, remotely controlled breeder fuel pin fabrication process which is to be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at Hanford and is scheduled for completion in 1984. The SAF line is scheduled for startup in 1987 and will produce mixed uranium-plutonium fuel pins for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor Plant (CRBRP). The fabrication line and support systems are described

  5. Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting

    International Nuclear Information System (INIS)

    Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B.

    2013-01-01

    The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs

  6. Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs.

  7. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

  8. Advanced Research Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-04-15

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

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

    International Nuclear Information System (INIS)

    1979-01-01

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

  10. Mixed U/Pu oxide fuel fabrication facility co-processed feed, pelletized fuel

    International Nuclear Information System (INIS)

    1978-09-01

    Two conceptual MOX fuel fabrication facilities are discussed in this study. The first facility in the main body of the report is for the fabrication of LWR uranium dioxide - plutonium dioxide (MOX) fuel using co-processed feed. The second facility in the addendum is for the fabrication of co-processed MOX fuel spiked with 60 Co. Both facilities produce pellet fuel. The spiked facility uses the same basic fabrication process as the conventional MOX plant but the fuel feed incorporates a high energy gamma emitter as a safeguard measure against diversion; additional shielding is added to protect personnel from radiation exposure, all operations are automated and remote, and normal maintenance is performed remotely. The report describes the fuel fabrication process and plant layout including scrap and waste processing; and maintenance, ventilation and safety measures

  11. Improvements in the fabrication of metallic fuels

    International Nuclear Information System (INIS)

    Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

    1989-01-01

    Argonne National Laboratory (ANL) is currently developing a new liquid-metal-cooled breeder reactor known as the Integral Fast Reactor (IFR). The IFR represents the state of the art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, discussed in this paper, will support ANL-West's (ANL-W) fully remote fuel cycle facility, which is an integral part of the IFR concept

  12. Development of an engineered safeguards system concept for a mixed-oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Chapman, L.D.; de Montmollin, J.M.; Deveney, J.E.; Fienning, W.C.; Hickman, J.W.; Watkins, L.D.; Winblad, A.E.

    1976-08-01

    An initial concept of an Engineered Safeguards System for a representative commercial mixed-oxide fuel fabrication facility is presented. Computer simulation techniques for evaluation and further development of the concept are described. An outline of future activity is included

  13. Fabrication, fabrication control and in-core follow up of 4 LEU leader fuel elements based on U3Si2 in RECH-1

    International Nuclear Information System (INIS)

    Chavez, J.C.; Barrera, M.; Olivares, L.; Lisboa, J.

    1999-01-01

    The RECH-1 MTR reactor has been converted from HEU to MEU (45% enrichment) and the decision to a LEU (20% enrichment) conversion was taken some years ago. This LEU conversion decision involved a local fuel development and fabrication based on U 3 Si 2 -Al dispersion fuel, and a fabrication qualification stage that resulted in four fuel elements fully complying with established fabrication standards for this type of fuel. This report-presents relevant points of these four leaders fuel elements fabrication, in particular a fuel plate core homogeneity control development. A summary of the intended in core follow-up studies for the leaders fuel elements is also presented here. (author)

  14. Role of ion chromatograph in nuclear fuel fabrication process at Nuclear Fuel Complex

    International Nuclear Information System (INIS)

    Balaji Rao, Y.; Prasada Rao, G.; Prahlad, B.; Saibaba, N.

    2012-01-01

    The present paper discusses the different applications of ion chromatography followed in nuclear fuel fabrication process at Nuclear Fuel Complex. Some more applications of IC for characterization of nuclear materials and which are at different stages of method development at Control Laboratory, Nuclear Fuel Complex are also highlighted

  15. Present state and problems of uranium fuel fabrication businesses

    International Nuclear Information System (INIS)

    Yuki, Akio

    1981-01-01

    The businesses of uranium fuel fabrication converting uranium hexafluoride to uranium dioxide powder and forming fuel assemblies are the field of most advanced industrialization among nuclear fuel cycle industries in Japan. At present, five plants of four companies engage in this business, and their yearly sales exceeded 20 billion yen. All companies are planning the augmentation of installation capacity to meet the growth of nuclear power generation. The companies of uranium fuel fabrication make the nuclear fuel of the specifications specified by reactor manufacturers as the subcontractors. In addition to initially loaded fuel, the fuel for replacement is required, therefore the demand of uranium fuel is relatively stable. As for the safety of enriched uranium flowing through the farbicating processes, the prevention of inhaling uranium powder by workers and the precaution against criticality are necessary. Also the safeguard measures are imposed so as not to convert enriched uranium to other purposes than peacefull ones. The strict quality control and many times of inspections are carried out to insure the soundness of nuclear fuel. The growth of the business of uranium fuel fabrication and the regulation of the businesses by laws are described. As the problems for the future, the reduction of fabrication cost, the promotion of research and development and others are pointed out. (Kako, I.)

  16. Nuclear fuel conversion and fabrication chemistry

    International Nuclear Information System (INIS)

    Lerch, R.E.; Norman, R.E.

    1984-01-01

    Following irradiation and reprocessing of nuclear fuel, two operations are performed to prepare the fuel for subsequent reuse as fuel: fuel conversion, and fuel fabrication. These operations complete the classical nuclear fuel cycle. Fuel conversion involves generating a solid form suitable for fabrication into nuclear fuel. For plutonium based fuels, either a pure PuO 2 material or a mixed PuO 2 -UO 2 fuel material is generated. Several methods are available for preparation of the pure PuO 2 including: oxalate or peroxide precipitation; or direct denitration. Once the pure PuO 2 is formed, it is fabricated into fuel by mechanically blending it with ceramic grade UO 2 . The UO 2 can be prepared by several methods which include direct denitration. ADU precipitation, AUC precipitation, and peroxide precipitation. Alternatively, UO 2 -PuO 2 can be generated directly using coprecipitation, direct co-denitration, or gel sphere processes. In coprecipitation, uranium and plutonium are either precipitated as ammonium diuranate and plutonium hydroxide or as a mixture of ammonium uranyl-plutonyl carbonate, filtered and dried. In direct thermal denitration, solutions of uranium and plutonium nitrates are heated causing concentration and, subsequently, direct denitration. In gel sphere conversion, solutions of uranium and plutonium nitrate containing additives are formed into spherical droplets, gelled, washed and dried. Refabrication of these UO 3 -PuO 2 starting materials is accomplished by calcination-reduction to UO 2 -PuO 2 followed by pellet fabrication. (orig.)

  17. The quality challenge for fuel fabrication

    International Nuclear Information System (INIS)

    Lannegrace, J.-P.

    1990-01-01

    Fuel fabrication is a key segment of the nuclear fuel cycle, since safe and economic operation of reactors is highly dependent on the quality of the fuel. Achieving and controlling quality is, therefore, of paramount importance to fuel fabricators dominating nearly every aspect of the business. The quality policy, concepts and assurance system at three French plants are outlined. The need for integrated inspection, process optimization and good employee motivation is stressed. (author)

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

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan; Jung, Sung Hoon

    1991-07-01

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

  19. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

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

  20. Minor Actinide Laboratory at JRC-ITU: Fuel fabrication facility

    International Nuclear Information System (INIS)

    Fernandez, A.; McGinley, J.; Somers, J.

    2008-01-01

    The Minor Actinide Laboratory (MA-lab) of the Institute for Transuranium Elements is a unique facility for the fabrication of fuels and targets containing minor actinides (MA). It is of key importance for research on Partitioning and Transmutation in Europe, as it is one of the only dedicated facilities for the fabrication of MA containing materials, either for property measurements or for the production of test pins for irradiation experiments. In this paper a detailed description of the MA-Lab facility and the fabrication processes developed to fabricate fuels and samples containing high content of minor actinides is given. In addition, experience gained and improvements are also outlined. (authors)

  1. Status of high-density fuel plate fabrication

    International Nuclear Information System (INIS)

    Wiencek, T.C.; Domagala, R.F.; Thresh, H.R.

    1991-01-01

    Progress has continued on the fabrication of fuel plates with equivalent fuel zone loadings approaching 9 gU/cm 3 . Through hot isostatic pressing (HIP), successful diffusion bonds have been made with 1100 Al and 6061 Al alloys. Although additional study is necessary to optimize the procedure, these bonds demonstrated the most critical processing step for proof-of-concept hardware. Two types of prototype highly loaded fuel plates have been fabricated. The first is a fuel plate in which 0.030-in. (0.76-mm) uranium compound wires are bonded within an aluminum cladding; the second, a dispersion fuel plate with uniform cladding and fuel zone thickness. The successful fabrication of these fuel plates derives from the unique ability of the HIP process to produce diffusion bonds with minimal deformation. (orig.)

  2. Interfacing robotics with plutonium fuel fabrication

    International Nuclear Information System (INIS)

    Bowen, W.W.; Moore, F.W.

    1986-01-01

    Interfacing robotic systems with nuclear fuel fabrication processes resulted in a number of interfacing challenges. The system not only interfaces with the fuel process, but must also interface with nuclear containment, radiation control boundaries, criticality control restrictions, and numerous other safety systems required in a fuel fabrication plant. The robotic system must be designed to allow operator interface during maintenance and recovery from an upset as well as normal operations

  3. Fabrication of Uranium Oxycarbide Kernels for HTR Fuel

    International Nuclear Information System (INIS)

    Barnes, Charles; Richardson, Clay; Nagley, Scott; Hunn, John; Shaber, Eric

    2010-01-01

    Babcock and Wilcox (B and W) has been producing high quality uranium oxycarbide (UCO) kernels for Advanced Gas Reactor (AGR) fuel tests at the Idaho National Laboratory. In 2005, 350-(micro)m, 19.7% 235U-enriched UCO kernels were produced for the AGR-1 test fuel. Following coating of these kernels and forming the coated-particles into compacts, this fuel was irradiated in the Advanced Test Reactor (ATR) from December 2006 until November 2009. B and W produced 425-(micro)m, 14% enriched UCO kernels in 2008, and these kernels were used to produce fuel for the AGR-2 experiment that was inserted in ATR in 2010. B and W also produced 500-(micro)m, 9.6% enriched UO2 kernels for the AGR-2 experiments. Kernels of the same size and enrichment as AGR-1 were also produced for the AGR-3/4 experiment. In addition to fabricating enriched UCO and UO2 kernels, B and W has produced more than 100 kg of natural uranium UCO kernels which are being used in coating development tests. Successive lots of kernels have demonstrated consistent high quality and also allowed for fabrication process improvements. Improvements in kernel forming were made subsequent to AGR-1 kernel production. Following fabrication of AGR-2 kernels, incremental increases in sintering furnace charge size have been demonstrated. Recently small scale sintering tests using a small development furnace equipped with a residual gas analyzer (RGA) has increased understanding of how kernel sintering parameters affect sintered kernel properties. The steps taken to increase throughput and process knowledge have reduced kernel production costs. Studies have been performed of additional modifications toward the goal of increasing capacity of the current fabrication line to use for production of first core fuel for the Next Generation Nuclear Plant (NGNP) and providing a basis for the design of a full scale fuel fabrication facility.

  4. USHPRR FUEL FABRICATION PILLAR: FABRICATION STATUS, PROCESS OPTIMIZATIONS, AND FUTURE PLANS

    Energy Technology Data Exchange (ETDEWEB)

    Wight, Jared M.; Joshi, Vineet V.; Lavender, Curt A.

    2018-03-12

    The Fuel Fabrication (FF) Pillar, a project within the U.S. High Performance Research Reactor Conversion program of the National Nuclear Security Administration’s Office of Material Management and Minimization, is tasked with the scale-up and commercialization of high-density monolithic U-Mo fuel for the conversion of appropriate research reactors to use of low-enriched fuel. The FF Pillar has made significant steps to demonstrate and optimize the baseline co-rolling process using commercial-scale equipment at both the Y-12 National Security Complex (Y-12) and BWX Technologies (BWXT). These demonstrations include the fabrication of the next irradiation experiment, Mini-Plate 1 (MP-1), and casting optimizations at Y-12. The FF Pillar uses a detailed process flow diagram to identify potential gaps in processing knowledge or demonstration, which helps direct the strategic research agenda of the FF Pillar. This paper describes the significant progress made toward understanding the fuel characteristics, and models developed to make informed decisions, increase process yield, and decrease lifecycle waste and costs.

  5. Quality assurance for breeder reactor fuel fabrication

    International Nuclear Information System (INIS)

    Marx, E.R.

    1978-01-01

    Fuel performance in the Fast Flux Test Facility (FFTF) depends on fabrication of fuel to rigorous quality standards. The quality program including Management, Procurement, Fabrication, Inspection, Records, and Audits is discussed as well as unique mixed oxide fuel inspections such as homogeneity inspection, analytical chemistry, and nondestructive fissile assay

  6. Development of the fabrication technology of the simulated fuel-I, 15,000MWd/tU

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kweon Ho; Kim, D. J.; Kim, H. S.; Lee, J. W.; Yang, M. S

    2001-04-01

    It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties, fission gas release, grain growth and et al. of the DUPIC fuel is different from the commercial UO2 fuel. But what directly measures physical properties et al. of DUPIC fuel being resinterred simulated spent fuel through OREOX process is very difficult in laboratory owing to its high level radiation. Then fabrication of simulated DUPIC fuel is needed to measure its properties. In this study, the sintering characterization of wet milled powder for 24 hours to fabricate 15,000MWd/tU equivalent burnup simulated fuel.

  7. Development of Automatic Quality Check Software in Mailbox Declaration For Nuclear Fuel Fabrication Plants

    International Nuclear Information System (INIS)

    Kim, Minsu; Shim, Hye Won; Jo, Seong Yeon; Lee, Kwang Yeol; Ban, Myoung Jin

    2014-01-01

    Short Notice Random Inspection (SNRI) is a new IAEA safeguards inspection regime for bulk handing facility, which utilities random inspection through a mailbox system. Its main objective is to verify 100% of the flow components of the safeguarded nuclear material at such a facility. To achieve the SNRI objective, it is required to provide daily mailbox declaration, by a facility's operator, to the IAEA with regard to information, such as the receipt and shipment of nuclear materials. Mailbox declarations are then later compared with accounting records so as to examine the accuracy and consistency of the facility operator's declaration at the time of the SNRI. The IAEA has emphasized the importance of accurate mailbox declarations and recommended that the ROK initiate its own independent quality control system in order to improve and maintain its mailbox declarations as a part of the SSAC activities. In an effort to improve the transparency of operational activities at fuel fabrication plants and to satisfy IAEA recommendation, an automatic quality check software application has been developed to improve mailbox declarations at fabrication plants in Korea. The ROK and the IAEA have recognized the importance of providing good quality mailbox declaration for an effective and efficient SNRI at fuel fabrication plants in Korea. The SRA developed an automatic quality check software program in order to provide an independent QC system of mailbox declaration, as well as to improve the quality of mailbox declaration. Once the automatic QC system is implemented, it will improve the quality of an operator's mailbox declaration by examining data before sending it to the IAEA. The QC system will be applied to fuel fabrication plants in the first half of 2014

  8. Advanced fuel fabrication

    International Nuclear Information System (INIS)

    Bernard, H.

    1989-01-01

    This paper deals with the fabrication of advanced fuels, such as mixed oxides for Pressurized Water Reactors or mixed nitrides for Fast Breeder Reactors. Although an extensive production experience exists for the mixed oxides used in the FBR, important work is still needed to improve the theoretical and technical knowledge of the production route which will be introduced in the future European facility, named Melox, at Marcoule. Recently, the feasibility of nitride fuel fabrication in existing commercial oxide facilities was demonstrated in France. The process, based on carbothermic reduction of oxides with subsequent comminution of the reaction product, cold pressing and sintering provides (U, Pu)N pellets with characteristics suitable for irradiation testing. Two experiments named NIMPHE 1 and 2 fabricated in collaboration with ITU, Karlsruhe, involve 16 nitride and 2 carbide pins, operating at a linear power of 45 and 73 kW/m with a smear density of 75-80% TD and a high burn-up target of 15 at%. These experiments are currently being irradiated in Phenix, at Marcoule. (orig.)

  9. An Overview of Current and Past W-UO[2] CERMET Fuel Fabrication Technology

    International Nuclear Information System (INIS)

    Douglas E. Burkes; Daniel M. Wachs; James E. Werner; Steven D. Howe

    2007-01-01

    Studies dating back to the late 1940s performed by a number of different organizations and laboratories have established the major advantages of Nuclear Thermal Propulsion (NTP) systems, particularly for manned missions. A number of NTP projects have been initiated since this time; none have had any sustained fuel development work that appreciably contributed to fuel fabrication or performance data from this era. As interest in these missions returns and previous space nuclear power researchers begin to retire, fuel fabrication technologies must be revisited, so that established technologies can be transferred to young researchers seamlessly and updated, more advanced processes can be employed to develop successful NTP fuels. CERMET fuels, specifically W-UO2, are of particular interest to the next generation NTP plans since these fuels have shown significant advantages over other fuel types, such as relatively high burnup, no significant failures under severe transient conditions, capability of accommodating a large fission product inventory during irradiation and compatibility with flowing hot hydrogen. Examples of previous fabrication routes involved with CERMET fuels include hot isostatic pressing (HIPing) and press and sinter, whereas newer technologies, such as spark plasma sintering, combustion synthesis and microsphere fabrication might be well suited to produce high quality, effective fuel elements. These advanced technologies may address common issues with CERMET fuels, such as grain growth, ductile to brittle transition temperature and UO2 stoichiometry, more effectively than the commonly accepted 'traditional' fabrication routes. Bonding of fuel elements, especially if the fabrication process demands production of smaller element segments, must be investigated. Advanced brazing techniques and compounds are now available that could produce a higher quality bond segment with increased ease in joining. This paper will briefly address the history of CERMET

  10. Application of vacuum technology during nuclear fuel fabrication, inspection and characterization

    International Nuclear Information System (INIS)

    Majumdar, S.

    2003-01-01

    Full text: Vacuum technology plays very important role during various stages of fabrication, inspection and characterization of U, Pu based nuclear fuels. Controlled vacuum is needed for melting and casting of U, Pu based alloys, picture framing of the fuel meat for plate type fuel fabrication, carbothermic reduction for synthesis of (U-Pu) mixed carbide powder, dewaxing of green ceramic fuel pellets, degassing of sintered pellets and encapsulation of fuel pellets inside clad tube. Application of vacuum technology is also important during inspection and characterization of fuel materials and fuel pins by way of XRF and XRD analysis, Mass spectrometer Helium leak detection etc. A novel method of low temperature sintering of UO 2 developed at BARC using controlled vacuum as sintering atmosphere has undergone successful irradiation testing in Cirus. The paper will describe various fuel fabrication flow sheets highlighting the stages where vacuum applications are needed

  11. Development of automation and remotisation systems for fabrication of (Th-233U)O2 MOX fuel for AHWR

    International Nuclear Information System (INIS)

    Saraswat, Anupam; Danny, K.M.; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun; Mittal, R.; Prasad, R.S.; Mahule, K.N.; Panda, S.; Jayarajan, K.

    2011-01-01

    To meet the ever increasing power requirement of India, country is planning to utilize its large thorium reserves for the third stage of nuclear power program based on Thorium-Uranium 233 fuel in A.H.W.R. Although there are many advantages of (Th- 233 U)O 2 fuel cycle, presence of radiological hazards due to the presence of 1000-2000 ppm level of 232 U in the 233 U fuel and inertness of ThO 2 makes handling and fabrication of fuel difficult. The associated high alpha and gamma activity demands high level of automation and remote handling in alpha tight hot cells. To demonstrate automation and remotisation in (Th- 233 U)O 2 fuel fabrication, a mock up facility is being set up at BARC. This facility shall develop automation systems required for remote fuel fabrication in a simulated hot cell environment. There are many innovative schemes and systems being developed like integrated powder pellet system, remote viewing system for hot cell application etc. Low visibility inside the hot cell has always been a problem for the operator. To overcome this problem a remote viewing system has been developed by which entire hot cell area can be scanned with the use of a joystick and the display can be seen on a LCD monitor. The viewing system is made up of radiation resistant optics which can work even in high gamma fields. It consists of objective end assembly which is used to scan the hot cell area with the help of prism doublets and drive mechanism for capturing full 360 deg solid angle view. There is a Galilean telescope and focusing system used for focusing images of distant objects. Drive mechanism can be controlled by the joystick available to the operator. System has a high resolution CCD display and camera which gives a clear display of objects lying inside the hot cell area. Integrated powder pellet system is being developed for fabrication of MOX pellets from feed powder. This will be automated system which will take input in the form of MOX powder and convert it

  12. Zirconia based inert matrix fuel: fabrication concepts and feasibility studies

    International Nuclear Information System (INIS)

    Ingold, F.; Burghartz, M.; Ledergerber, G.

    1999-01-01

    The internal gelation process has traditionally been applied to fabricate standard fuel based on uranium, typically UO2 and MOX. To meet the recent aim to destroy plutonium in the most effective way, a uranium free fuel was evaluated. The fuel development programme at PSI has been redirected toward a fuel based on zirconium oxide or a mixture of zirconia and a conducting material to form ceramic/metal (CERMET) or ceramic/ceramic (CERCER) combinations. A feasibility study was carried out to demonstrate that microspheres based on zirconia and spinel can be fabricated with the required properties. The gelation parameters were investigated to optimise compositions of the starting solutions. Studies to fabricate a composite material (from zirconia and spinel) are ongoing. If the zirconia/spinel ratio is chosen appropriately, the low thermal conductivity of pure zirconia can be compensated by the higher thermal conductivity of spinel. Another solution to offset the low thermal conductivity of zirconia is the development of a CERMET, which consists of fine particles bearing plutonium in a cubic zirconia lattice dispersed in a metallic matrix. The fabrication of such a CERMET is also being studied. (author)

  13. French development program on fuel cycle

    International Nuclear Information System (INIS)

    Viala, M.; Bourgeois, M.

    1991-01-01

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

  14. Develpment of quality assurance manual for fabrication of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Gun; Lee, J. W.; Kim, S. S. and others

    2001-09-01

    The Quality Assurance Manual for the fabrication of DUPIC fuel with high quality was developed. The Quality Assurance Policy established by this manual is to assure that the DUPIC fuel element supplied to customer conform to the specified requirements of customer, applicable codes and standards. The management of KAERI is committed to implementation and maintenance of the program described by this manual. This manual describes the quality assurance program for DUPIC fuel fabrication to comply with CAN3-Z299.2-85 to the extent as needed and appropriate. This manual describes the methods which DUPIC Fuel Development Team(DFDT) personnel must follow to achieve and assure high quality of our product. This manual also describes the quality management system applicable to the activities performed at DFDT.

  15. Develpment of quality assurance manual for fabrication of DUPIC fuel

    International Nuclear Information System (INIS)

    Lee, Young Gun; Lee, J. W.; Kim, S. S. and others

    2001-09-01

    The Quality Assurance Manual for the fabrication of DUPIC fuel with high quality was developed. The Quality Assurance Policy established by this manual is to assure that the DUPIC fuel element supplied to customer conform to the specified requirements of customer, applicable codes and standards. The management of KAERI is committed to implementation and maintenance of the program described by this manual. This manual describes the quality assurance program for DUPIC fuel fabrication to comply with CAN3-Z299.2-85 to the extent as needed and appropriate. This manual describes the methods which DUPIC Fuel Development Team(DFDT) personnel must follow to achieve and assure high quality of our product. This manual also describes the quality management system applicable to the activities performed at DFDT

  16. Assesment On The Possibility To Modify Fabrication Equipment For Fabrication Of HWR And LWR Fuel Elements

    International Nuclear Information System (INIS)

    Tri-Yulianto

    1996-01-01

    Based on TOR BATAN for PELITA VI. On of BATAN program in the fuel element production technology section is the acquisition of the fuel element fabrication technology for research reactor as well as power reactor. The acquisition can be achieved using different strategies, e.g. by utilizing the facility owned for research and development of the technology desired or by transferring the technology directly from the source. With regards to the above, PEBN through its facility in BEBE has started the acquisition of the fuel element fabrication technology for power reactor by developing the existing equipment initially designed to fabricate HWR Cinere fuel element. The development, by way of modifying the equipment, is intended for the production of HWR (Candu) and LWR (PWR and BWR) fuel elements. To achieve above objective, at the early stage of activity, an assesment on the fabrication equipment for pelletizing, component production and assembly. The assesment was made by comparing the shape and the size of the existing fuel element with those used in the operating reactors such as Candu reactors, PWR and BWR. Equipment having the potential to be modified for the production of HWR fuel elements are as followed: For the pelletizing equipment, the punch and dies can be used of the pressing machine for making green pellet can be modified so that different sizes of punch and dies can be used, depending upon the size of the HWR and LWR pellets. The equipment for component production has good potential for modification to produce the HWR Candu fuel element, which has similar shape and size with those of the existing fuel element, while the possibility of producing the LWR fuel element component is small because only a limited number of the required component can be made with the existing equipment. The assembly equipment has similar situation whit that of the component production, that is, to assemble the HWR fuel element modification of few assembly units very probable

  17. Typical IAEA operations at a fuel fabrication plant

    International Nuclear Information System (INIS)

    Morsy, S.

    1984-01-01

    The IAEA operations performed at a typical Fuel Fabrication Plant are explained. To make the analysis less general the case of Low Enriched Uranium (LEU) Fuel Fabrication Plants is considered. Many of the conclusions drawn from this analysis could be extended to other types of fabrication plants. The safeguards objectives and goals at LEU Fuel Fabrication Plants are defined followed by a brief description of the fabrication process. The basic philosophy behind nuclear material stratification and the concept of Material Balance Areas (MBA's) and Key Measurement Points (KMP's) is explained. The Agency operations and verification methods used during physical inventory verifications are illustrated

  18. 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

  19. Fabrication of simulated DUPIC fuel

    Science.gov (United States)

    Kang, Kweon Ho; Song, Ki Chan; Park, Hee Sung; Moon, Je Sun; Yang, Myung Seung

    2000-12-01

    Simulated DUPIC fuel provides a convenient way to investigate the DUPIC fuel properties and behavior such as thermal conductivity, thermal expansion, fission gas release, leaching, and so on without the complications of handling radioactive materials. Several pellets simulating the composition and microstructure of DUPIC fuel are fabricated by resintering the powder, which was treated through OREOX process of simulated spent PWR fuel pellets, which had been prepared from a mixture of UO2 and stable forms of constituent nuclides. The key issues for producing simulated pellets that replicate the phases and microstructure of irradiated fuel are to achieve a submicrometre dispersion during mixing and diffusional homogeneity during sintering. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent PWR fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent PWR fuel agrees well with the other studies. The leading structural features observed are as follows: rare earth and other oxides dissolved in the UO2 matrix, small metallic precipitates distributed throughout the matrix, and a perovskite phase finely dispersed on grain boundaries.

  20. Technical report: fabrication of PWR type rodlet fuel

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Uno, Hisao; Sasajima, Hideo

    1990-06-01

    With respect to the simulated reactivity initiated accident (RIA) experiments with pre-irradiated LWR type fuel rods at nuclear safety research reactor (NSRR), there were principally three technical difficulties which should be overcome: (1) Fabrication of the rodlet fuel; Fuel rods from the commercial power reactors had an active column length by 3.6m. To utilize this for NSRR pulse experiment, rodlet fuel having an active column length by 0.12m (reduced to one thirtieth) is requested to fabricate without changing the inside fuel conditions. (2) Development of in-core instrumentations: During pre-irradiation stages, a long-sized fuel rod had dimensional changes by waterside corrosion, bowing, creep down and so on. The fuel also had greater amount of radioactive fission products. This condition is significant to in-core instrumentations to be attached to the fuel rods. Well characterized data to be obtained from these, however, are quite necessary and important from research point of view. Remote handling techniques to attach the rod pressure sensor, the cladding extensometer, the fuel extensometer, and the cladding surface thermocouple to pre-irradiated fuel rods are, therefore, requested to develop. (3) Installation of PIE equipments for pulsed rodlet fuels: PIE on the pulsed rodlet fuels are necessary to better understanding the fuel performance detaily. Equipments which can easily detect the data related to PCMI type fuel failure are matter of concern. Since 1986, the technical difficulties have been tried to overcome by all staffs belonging to Reactivity Accident Laboratory, NSRR Operation Division, Department of Reactor Fuel Examination and Hot Laboratory. This report describes the technical achievements obtained through four years work. (author)

  1. The Role of Friction Stir Welding in Nuclear Fuel Plate Fabrication

    International Nuclear Information System (INIS)

    Burkes, D.; Medvedev, P.; Chapple, M.; Amritkar, A.; Wells, P.; Charit, I

    2009-01-01

    The friction bonding process combines desirable attributes of both friction stir welding and friction stir processing. The development of the process is spurred on by the need to fabricate thin, high density, reduced enrichment fuel plates for nuclear research reactors. The work seeks to convert research and test reactors currently operating on highly enriched uranium fuel to operate on low enriched uranium fuel without significant loss in reactor performance, safety characteristics, or significant increase in cost. In doing so, the threat of global nuclear material proliferation will be reduced. Feasibility studies performed on the process show that this is a viable option for mass production of plate-type nuclear fuel. Adapting the friction stir weld process for nuclear fuel fabrication has resulted in the development of several unique ideas and observations. Preliminary results of this adaptation and process model development are discussed

  2. Reproduction of the RA reactor fuel element fabrication

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

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

  3. Development of nuclear fuel cycle technologies

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Matsumoto, Takashi; Suzuki, Kazumichi; Kawamura, Fumio

    1995-01-01

    In the long term plan for atomic energy that the Atomic Energy Commission decided the other day, the necessity of the technical development for establishing full scale fuel cycle for future was emphasized. Hitachi Ltd. has engaged in technical development and facility construction in the fields of uranium enrichment, MOX fuel fabrication, spent fuel reprocessing and so on. In uranium enrichment, it took part in the development of centrifuge process centering around Power Reactor and Nuclear Fuel Development Corporation (PNC), and took its share in the construction of the Rokkasho uranium enrichment plant of Japan Nuclear Fuel Service Co., Ltd. Also it cooperates with Laser Enrichment Technology Research Association. In Mox fuel fabrication, it took part in the construction of the facilities for Monju plutonium fuel production of PNC, for pellet production, fabrication and assembling processes. In spent fuel reprocessing, it cooperated with the technical development of maintenance and repair of Tokai reprocessing plant of PNC, and the construction of spent fuel stores in Rokkasho reprocessing plant is advanced. The centrifuge process and the atomic laser process of uranium enrichment are explained. The high reliability of spent fuel reprocessing plants and the advancement of spent fuel reprocessing process are reported. Hitachi Ltd. Intends to exert efforts for the technical development to establish nuclear fuel cycle which increases the importance hereafter. (K.I.)

  4. Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

    Energy Technology Data Exchange (ETDEWEB)

    Wiencek, T.C. [Argonne National Lab., IL (United States). Energy Technology Div.

    1995-08-01

    This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched ({approx}93% {sup 235}U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm{sup 3}) HEU fuel elements to highly loaded (up to 7 g U/cm{sup 3}) low-enrichment (<20% {sup 235}U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing.

  5. Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

    International Nuclear Information System (INIS)

    Wiencek, T.C.

    1995-08-01

    This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched (∼93% 235 U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm 3 ) HEU fuel elements to highly loaded (up to 7 g U/cm 3 ) low-enrichment ( 235 U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  7. New fabrication techniques for the nuclear fuels of tomorrow

    International Nuclear Information System (INIS)

    Babelot, J.F.; Bokelund, H.; Gerontopoulos, P.; Gueugnon, J.F.; Richter, K.

    1995-01-01

    The shift of the emphasis of the work at the Institute for Transuranium Elements (ITU) from the development of fuels based on uranium and plutonium to safety aspects concerning the use of plutonium and other of actinides, necessitates the production of targets containing appreciable amounts of minor actinides for irradiation experiments. The handling of minor actinides requires additional protective measures, combined with improved fuel fabrication techniques. The boundary conditions for a suitable process are flexibility, adaptability to remote control, and minimization of dust formation. A method based on the sol-gel fabrication technique meets these criteria, and was selected for the present developments at ITU. (author)

  8. Development of likelihood estimation method for criticality accidents of mixed oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tamaki, Hitoshi; Yoshida, Kazuo; Kimoto, Tatsuya; Hamaguchi, Yoshikane

    2010-01-01

    A criticality accident in a MOX fuel fabrication facility may occur depending on several parameters, such as mass inventory and plutonium enrichment. MOX handling units in the facility are designed and operated based on the double contingency principle to prevent criticality accidents. Control failures of at least two parameters are needed for the occurrence of criticality accident. To evaluate the probability of such control failures, the criticality conditions of each parameter for a specific handling unit are necessary for accident scenario analysis to be clarified quantitatively with a criticality analysis computer code. In addition to this issue, a computer-based control system for mass inventory is planned to be installed into MOX handling equipment in a commercial MOX fuel fabrication plant. The reliability analysis is another important issue in evaluating the likelihood of control failure caused by software malfunction. A likelihood estimation method for criticality accident has been developed with these issues been taken into consideration. In this paper, an example of analysis with the proposed method and the applicability of the method are also shown through a trial application to a model MOX fabrication facility. (author)

  9. DUPIC nuclear fuel manufacturing and process technology development

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Park, J. J.; Lee, J. W.

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated

  10. DUPIC nuclear fuel manufacturing and process technology development

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, J. J.; Lee, J. W. [and others

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated.

  11. Advances in AGR fuel fabrication - now and the future

    International Nuclear Information System (INIS)

    Bleasdale, P.A.

    1995-01-01

    To date, over 3 million AGR fuel pins have been manufactured at Springfields for the UK AGR programme. During this time, AGR fuel design and manufacture has developed and evolved in response to the needs of the reactor operators to enhance fuel reliability and performance. More recently, major advances have been made in the systems and organisational culture which support fuel manufacture at Fuel Division. The introduction of MRP II in 1989 into Fuel Division enabled significant reductions in stock and work-in-progress, together with reductions in manufacturing lead times. Other successful initiatives introduced into Fuel Division have been Just-in-Time (JIT) and AST (Additional Skills Training) which have built on the success of MRP II. All of these initiatives are evidence of Fuel Division's ''Total Quality'' approach to fabricating fuel. Fuel Division is currently in the final stages of commissioning the New Oxide Fuels Complex (NOFC) where both AGR and PWR fuel will be manufactured to the highest standards of quality, safety and environmental protection. NOFC is a totally integrated plant which represents a Pound 200M investment, demonstrating Fuel Division's commitment to building on its 40+ years of fuel fabrication experience and ensuring secure supply of fuel to its customers for years to come. (author)

  12. Fabrication development for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Pace, B.W.; Copeland, G.L.

    1995-08-01

    This report presents the fuel fabrication development for the Advanced Neutron Source (ANS) reactor. The fuel element is similar to that successfully fabricated and used in the High Flux Isotope Reactor (HFIR) for many years, but there are two significant differences that require some development. The fuel compound is U 3 Si 2 rather than U 3 O 8 , and the fuel is graded in the axial as well as the radial direction. Both of these changes can be accomplished with a straightforward extension of the HFIR technology. The ANS also requires some improvements in inspection technology and somewhat more stringent acceptance criteria. Early indications were that the fuel fabrication and inspection technology would produce a reactor core meeting the requirements of the ANS for the low volume fraction loadings needed for the highly enriched uranium design (up to 1.7 Mg U/m 3 ). Near the end of the development work, higher volume fractions were fabricated that would be required for a lower- enrichment uranium core. Again, results look encouraging for loadings up to ∼3.5 Mg U/m 3 ; however, much less evaluation was done for the higher loadings

  13. Design and development on automated control system of coated fuel particle fabrication process

    International Nuclear Information System (INIS)

    Liu Malin; Shao Youlin; Liu Bing

    2013-01-01

    With the development trend of the large-scale production of the HTR coated fuel particles, the original manual control system can not meet the requirement and the automation control system of coated fuel particle fabrication in modern industrial grade is needed to develop. The comprehensive analysis aiming at successive 4-layer coating process of TRISO type coated fuel particles was carried out. It was found that the coating process could be divided into five subsystems and nine operating states. The establishment of DCS-type (distributed control system) of automation control system was proposed. According to the rigorous requirements of preparation process for coated particles, the design considerations of DCS were proposed, including the principle of coordinated control, safety and reliability, integration specification, practical and easy to use, and open and easy to update. A complete set of automation control system for coated fuel particle preparation process was manufactured based on fulfilling the requirements of these principles in manufacture practice. The automated control system was put into operation in the production of irradiated samples for HTRPM demonstration project. The experimental results prove that the system can achieve better control of coated fuel particle preparation process and meet the requirements of factory-scale production. (authors)

  14. 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

  15. 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.

  16. Estimates of Canadian fuel fabrication costs for alternative fuel cycles and systems

    International Nuclear Information System (INIS)

    Blahnik, C.

    1979-04-01

    Unit fuel fabrication costs are estimated for alternate fuel cycles and systems that may be of interest in Ontario Hydro's strategy analyses. A method is proposed for deriving the unit fuel fabrication price to be paid by a Canadian utility as a function of time (i.e. the price that reflects the changing demand/supply situation in the particular scenario considered). (auth)

  17. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2014-01-01

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

  18. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

  19. Shield requirement estimation for pin storage room in fuel fabrication plant

    International Nuclear Information System (INIS)

    Shanthi, M.M.; Keshavamurthy, R.S.; Sivashankaran, G.

    2012-01-01

    Fast Reactor Fuel Cycle Facility (FRFCF) is an upcoming project in Kalpakkam. It has the facility to recycle the fuel from PFBR. It is an integrated facility, consists of fuel reprocessing plant, fuel fabrication plant (FFP), core subassembly plant, Reprocessed Uranium plant (RUP) and waste management plant. The spent fuel from PFBR would be reprocessed in fuel reprocessing plant. The reprocessed fuel material would be sent to fuel fabrication plant. The main activity of fuel fabrication plant is the production of MOX fuel pins. The fuel fabrication plant has a fuel pin storage room. The shield requirement for the pin storage room has been estimated by Monte Carlo method. (author)

  20. Safety analysis of DUPIC fuel development facility

    International Nuclear Information System (INIS)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Baek, S. Y.; Ahn, J. Y.

    2001-01-01

    Various experimental facilities are necessary in order to perform experimental verification for development of DUPIC fuel fabrication technology. In special, since highly radioactive material such as spent PWR fuel is used for this experiment, DUPIC fuel fabrication has to be performed in hot cell by remote handling. Therefore, it should be provided with proper engineering requirement and safety. M6 hot cell of IMEF which is to used for DUPIC fuel fabrication experiment was constructed as an α-γ hot cell for material examination of small amount of high-burnup fuel. The characteristics and amount of spent fuel for DUPIC fuel fabrication experiment will be different from the original design criteria. Therefore, the increased amount of spent fuel and different characteristics of experiment result in not only change of shielding and enviornmental evaluation results but new requirement of nuclear criticality evaluation. Therefore, this study includes evaluation of shielding, environmental effect and nuclear criticality in case that IMEF M6 hot cell is used for DUPIC fuel fabrication

  1. AECL's progress in DUPIC fuel development

    International Nuclear Information System (INIS)

    Sullivan, J.D.; Ryz, M.A.; Lee, J.W.

    1997-01-01

    Previous papers described progress in choosing a fabrication route for the DUPIC (Direct Use of Spent PWR Fuel in CANDU) fuel cycle [1], details of the OREOX (Oxidation Reduction of Oxide fuel) process, and preliminary results of out-cell and small-scale in-cell experiments [2]. AECL's project to develop the DUPIC fuel cycle has now progressed to the stage of fabricating DUPIC fuel elements for irradiation testing in a research reactor. Because of the high radiation fields around the spent PWR fuel, all work is being done in hot cells. The equipment used for fabrication of the DUPIC fuel elements is described in this paper. The commissioning, in-cell installation and current status of the fabrication process are also described and plans for the completion of this phase of the DUPIC project are outlined. The goal of this phase of the project is demonstration of the technical feasibility of the DUPIC fuel cycle. (author)

  2. Unified fuel elements development for research reactors

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  3. MOX fuel fabrication, in reactor performance and improvement

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  4. Development and implementation of computational geometric model for simulation of plate type fuel fabrication process with microspheres dispersed in metallic matrix

    International Nuclear Information System (INIS)

    Lage, Aldo M.F.; Reis, Sergio C.; Braga, Daniel M.; Santos, Armindo; Ferraz, Wilmar B.

    2005-01-01

    In this report it is presented the development of a geometric model to simulate the plate type fuel fabrication process with fuels microspheres dispersed in metallic matrix, as well as its software implementation. The developed geometric model encloses the steps of pellets pressing and sintering, as well as the plate rolling passes. The model permits the simulation of structures, where the values of the various variables of the fabrication processes can be studied and modified. The following variables were analyzed: microspheres diameters, density of the powder/microspheres mixing, microspheres density, fuel volume fraction, sintering densification, and rolling passes number. In the model implementation, which was codified in DELPHI programming language, systems of structured analysis techniques were utilized. The structures simulated were visualized utilizing the AutoCAD applicative, what permitted to obtain planes sections in diverse directions. The objective of this model is to enable the analysis of the simulated structures and supply information that can help in the improvement of the dispersion microspheres fuel plates fabrication process, now in development at CDTN (Centro de Desenvolvimento da Tecnologia Nuclear) in cooperation with the CTMSP (Centro Tecnologico da Marinha em Sao Paulo). (author)

  5. Fuel-pellet-fabrication experience using direct-denitration-recycle-PuO2-coprecipitated mixed oxide

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Schaus, P.S.

    1980-01-01

    The fuel pellet fabrication experience described in this paper involved three different feed powders: coprecipitated PuO 2 -UO 2 which was flash calcined in a fluidized bed; co-direct denitrated PuO 2 -UO 2 ; and direct denitrated LWR recycle PuO 2 which was mechanically blended with natural UO 2 . The objectives of this paper are twofold; first, to demonstrate that acceptable quality fuel pellets were fabricated using feed powders manufactured by processes other than the conventional oxalate process; and second, to highlight some pellet fabrication difficulties experienced with the direct denitration LWR recycle PuO 2 feed material, which did not produce acceptable pellets. The direct denitration LWR recycle PuO 2 was available as a by-product and was not specifically produced for use in fuel pellet fabrication. Nevertheless, its characteristics and pellet fabrication behavior serve to re-emphasize the importance of continued process development involving both powder suppliers and fuel fabricators to close the fuel cycle in the future

  6. Informal presentations by fuel fabricators and others [contributed by A. Nishiyama, Nuclear Fuel Industries, Ltd.

    International Nuclear Information System (INIS)

    Nishiyama, A.

    1993-01-01

    This paper contains a brief summary of activities in the field of research reactor fuel fabrication in Nuclear Fuel Industries Sumitomo and Furukawa Industries. Since 1956 2 million dollars were spent for development of nuclear fuels and plant facilities including complete manufacturing and testing capabilities. Now this company is the only fuel supplier for the research reactors in Japan. The fabrication process starts with the melting, alloying, and casting of U-Al. The uranium billets are prepared by foreign fabricators. The uranium content varies from 13 to 22 wt % according to the purchaser's specifications. In making fuel plates, the picture frame method is applied. In this case, the original procedure is sufficiently effective in avoiding dogboning. The plates are finished by hot and cold roll milling and inspected dimensionally, metallurgically, and mechanically, and at the same time the blister test and X-ray radiographic tests are performed. Fuel elements are assembled by rolling flat or curved plates into side plate grooves and end-fit welding. Finished elements are tested dimensionally and hydraulically. Nominal losses during operation are less than 1% of the uranium metal. Our present capacity licensed by the Japanese Government is approximately 950 fuel elements a year. About 35 employees including engineers are engaged in development and manufacturing of fuels. Owing to the small limited demand of the research reactor fuels in Japan during the past 20 years (mostly in last 10 years), we processed only about 350 kg of highly enriched uranium and supplied approximately 1000 fuel elements to JAERI, Kyoto University, and others, and we have been suffering red-ink balance of budget every year. Some of trials in development are briefly discussed. In case of UO 2 -Al metal fuel plates, the vibratory compacting method was very popular among many researchers about 10 years ago. A lot of time and money was spent to study the economic fabrication process of

  7. Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

    International Nuclear Information System (INIS)

    Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.; Wood, Thomas W.; Perkins, Casey J.

    2011-01-01

    For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may be constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.

  8. Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pasqualini, E.E. [Laboratorio de Nanotecnología Nuclear, Centro Atómico Constituyentes, Comisión Nacional de Energía Atómica, Av. General Paz 1499, B1650KNA, San Martín, Prov. Buenos Aires (Argentina); Robinson, A.B. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Wachs, D.M. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organisation, PMB 1, Menai, NSW, 2234 (Australia)

    2016-10-15

    Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm{sup 3}, 3.8E+21 (peak).

  9. Fabrication experience with mixed-oxide LWR fuels at the BELGONUCLEAIRE plant

    International Nuclear Information System (INIS)

    Vanhellemont, G.

    1979-01-01

    For nearly 20 years BELGONUCLEAIRE has been involved in a steadily growing effort to increase its production of mixed oxides. This programme has ranged from basic research and process development through a pilot-scale unit to today's mixed-oxide fuel fabrication plant at Dessel, which has been in operation for just over 5 years. The reference fabrication flow sheet includes UO 2 , PuO 2 and a scraped powder preparation, sintered ground pellets as well as rod fabrication and assembling. With regard to quality, attention is especially paid to the process monitoring and quality controls at the qualification step and during the routine production. Entirely different types of thermal UO 2 -PuO 2 fuel pellets, rods and assemblies have been manufactured for PWR and BWR operation. For these fabrications, some diagrams of the results with regard to the required technical specifications are presented. Special emphasis is placed on the occasional deviations of some finished products from the specifications and on the solutions applied to avoid such problems. Concerning the actual capacity of the mixed-oxide fuel fabrication plant, several limiting factors due to the nature of plutonium itself are discussed. Taking into account all these ambient limitations, a reference PWR mixed-oxide fuel output of nominally 18 t/a is obtained. The industrial feasibility of UO 2 -PuO 2 fuel fabrication has been thoroughly demonstrated by the present BELGONUCLEAIRE plant. The experience obtained has led to progressive improvements of the fabrication process and adaptation of the product controls in order to ensure the requested quality levels. (author)

  10. Design of the MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Johnson, J.V.; Brabazon, E.J.

    2001-01-01

    A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)

  11. Design of the MOX fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.V. [MFFF Technical Manager, U.S. dept. of Energy, Washington, DC (United States); Brabazon, E.J. [MFFF Engineering Manager, Duke Cogema Stone and Webster, Charlotte, NC (United States)

    2001-07-01

    A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)

  12. TRIGA International, a new TRIGA fuel fabrication facility at CERCA

    International Nuclear Information System (INIS)

    Harbonnier, G.

    1997-01-01

    At the time when General Atomics expressed its intention to cease fuel fabrication on its site of San Diego, CERCA has been chosen to carry on the fabrication of TRIGA fuel. After negotiations in 1994 and 1995, a partnership 50%/50% was decided and on July 1995, a new company was founded, with the name TRIGA INTERNATIONAL SAS, head office in Paris and fuel fabrication facility at CERCA in Romans. The intent of this presentation is, after a short reminder about TRIGA fuel design and fabrication to describe the new facility with special emphasis on the safety features associated with the modification of existing fabrication buildings. (author)

  13. Hybrid pellets: an improved concept for fabrication of nuclear fuel

    International Nuclear Information System (INIS)

    Matthews, R.B.; Hart, P.E.

    1979-09-01

    The feasibility of fabricating fuel pellets using gel-derived microspheres as press feed was evaluated. By using gel-derived microspheres as press feed, the potential exists for eliminating dusty operations like milling, slugging, and granulation, from the pelleting process. The free-flowing character of the spheres should also result in limited dust generation during powder transport and pressing operations. The results of this study clearly demonstrate that fuel pellets can be successfully fabricated on a laboratory scale using UO 2 gel microspheres as press feed. Under moderate sintering conditions, 1,500 0 C for 4 h in Ar-4% H 2 , UO 2 pellets with densities up to 96% TD were fabricated. A range of pellet microstructures and densities were achieved depending on sphere forming and calcining conditions. Based on these results, a set of necessary sphere properties are suggested: O/U less than 2.20, crystallite size less than 500 A, specific surface area greater than 8 m 2 /g, and sphere size 200 and 400 μm. Preliminary attempts to fabricate ThO 2 and ThO 2 -UO 2 pellets using microspheres were unsuccessful because the requisite sphere properties were not achieved. Areas requiring additional development include: demonstration of the process on scaled-up equipment suitable for use in a remote fuel fabrication facility and evaluation of the irradiation performance of pellet fuels from gel-spheres

  14. Prediction of dose and field mapping around a shielded plutonium fuel fabrication glovebox

    International Nuclear Information System (INIS)

    Strode, J.N.; Soldat, K.L.; Brackenbush, L.W.

    1984-01-01

    Westinghouse Hanford Company, as the Department of Energy's (DOE) prime contractor for the operation of the Hanford Engineering Development Laboratory (HEDL), is responsible for the development of the Secure Automated Fabrication (SAF) Line which is to be installed in the recently constructed Fuels and Materials Examination Facility (FMEF). The SAF Line will fabricate mixed-oxide (MOX) fuel pins for the Fast Flux Test Facility (FFTF) at an annual throughput rate of six (6) metric tons (MT) of MOX. The SAF Line will also demonstrate the automated manufacture of fuel pins on a production-scale. This paper describes some of the techniques used to reduce personnel exposure on the SAF Line, as well as the prediction and field mapping of doses from a shielded fuel fabrication glovebox. Tables are also presented from which exposure rate estimates can be made for plutonium recovered from fuels having different isotopic compositions as a result of varied burnup

  15. Criticality safety studies for plutonium–uranium metal fuel pin fabrication facility

    International Nuclear Information System (INIS)

    Stephen, Neethu Hanna; Reddy, C.P.

    2013-01-01

    Highlights: ► Criticality safety limits for PUMP-F facility is identified. ► The fissile mass which can be handled safely during alloy preparation is 10.5 kg. ► The number of fuel slugs which can be handled safely during injection casting is 53. ► The number of fuel slugs which can be handled safely after fuel fabrication is 71. - Abstract: This study focuses on the criticality safety during the fabrication of fast reactor metal fuel pins comprising of the fuel type U–15Pu, U–19Pu and U–19Pu–6Zr in the Plutonium–Uranium Metal fuel Pin fabrication Facility (PUMP-F). Maximum amount of fissile mass which can be handled safely during master alloy preparation, Injection casting and fuel slug preparation following fuel pin fabrication were identified and fixed based on this study. In the induction melting furnace, the fissile mass can be limited to 10.5 kg. During fuel slug preparation and fuel pin fabrication, fuel slugs and pins were arranged in hexagonal and square lattices to identify the most reactive configuration. The number of fuel slugs which can be handled safely after injection casting can be fixed to be 53, whereas after fuel fabrication it is 71

  16. Fabrication of metallic channel-containing UO2 fuels

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho

    2004-01-01

    The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet

  17. Molybdenum-base cermet fuel development

    International Nuclear Information System (INIS)

    Gurwell, W.E.; Moss, R.W.; Pilger, J.P.; White, G.D.

    1987-07-01

    Development of a multimegawatt (MMW) space nuclear power system requires identification and resolution of several technical feasibility issues before selecting one or more promising system concepts. Demonstration of reactor fuel fabrication technology is required for cermet-fueled reactor concepts. MMW reactor fuel development activity at Pacific Northwest Laboratory (PNL) is focused on producing a molybdenum-matrix uranium-nitride (UN) fueled cermet. This cermet is to have a high matrix density (≥95%) for high strength and high thermal conductance coupled with a high particle (UN) porosity (∼25%) for retention of released fission gas at high burnup. Fabrication process development involves the use of porous TiN microspheres as surrogate fuel material until porous UN microspheres become available. Process development has been conducted in the areas of microsphere synthesis, particle sealing/coating, and high-energy-rate forming (HERF) and vacuum hot press consolidation techniques. This paper summarizes the status of these activities

  18. Fabrication of inert matrix fuel for the incineration of plutonium - a feasibility study

    International Nuclear Information System (INIS)

    Burghartz, M.; Ledergerber, G.; Ingold, F.; Xie, T.; Botta, F.; Idemitsu, K.

    1998-01-01

    The internal gelation process has been applied to fabricate classical fuel based on uranium like UO 2 and MOX. For recent aims to destroy plutonium in the most effective way, a uranium free fuel was evaluated. The fuel development at PSI has been redirected to a fuel based on zirconium oxide or a mixture of zirconia and a conducting material leading to ceramic/metal (CERMET) or ceramic/ceramic (CERCER) combinations. A feasibility study was carried out to demonstrate that microspheres based on zirconia and spinel can be fabricated. The gelation parameters were investigated leading to optimised compositions for the starting solutions. Studies to fabricate a composite material (from zirconia and spinel) are ongoing. If the zirconia/spinel ratio is chosen appropriately, the low thermal conductivity of pure zirconia could be compensated by the higher thermal conductivity of spinel. Another solution to improve the low thermal conductivity of zirconia is the development of a CERMET, which consists of fine particles bearing plutonium in a cubic zirconia dispersed in a metallic matrix. The fabrication of such a CERMET is also being studied. (author)

  19. Highlights of 50 years of nuclear fuels developments

    International Nuclear Information System (INIS)

    Simnad, M.T.

    1989-01-01

    The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel cycle costs and reliable performance in the fuel elements are discussed in the historical context

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  1. GCRA review and appraisal of fuel material development programs

    International Nuclear Information System (INIS)

    1980-09-01

    The Fuel material Development Program has as its principal objective and responsibility the development of a fuel that is both economical and licensable and that, at the same time, will fulfill the required performance criteria. To accomplish this, the program is broken down into the following major fuel development task areas: development of the experimental and analytical data base for selecting, qualifying, and verifying the reference fuel design; providing the data base and developing models for evaluating fuel performance under upset and accident conditions; and developing and justifying fuel fabrication specifications which are consistent with the overall fuel performance criteria and with the fuel fabrication process capabilities

  2. Application of robotics in remote fuel fabrication operations

    International Nuclear Information System (INIS)

    Nyman, D.H.; Nagamoto, T.T.

    1984-01-01

    The Secure Automated Fabrication (SAF) line, an automated and remotely controlled manufacturing process, is scheduled for startup in 1987 and will produce mixed uranium/plutonium oxide fuel pins for the Fast Flux Test Facility (FFTF). The application of robotics in the fuel fabrication and supporting operations is described

  3. Fabrication and testing of uranium nitride fuel for space power reactors

    Science.gov (United States)

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

    1988-02-01

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

  4. Fuel fabrication instrumentation and control system overview

    International Nuclear Information System (INIS)

    Bennett, D.W.; Fritz, R.L.

    1980-10-01

    A process instrumentation and control system is being developed for automated fabrication of breeder reactor fuel at the Hanford Engineering Development Laboratory (HEDL) in Richland, Washington. The basic elements of the control system are a centralized computer system linked to distributed local computers, which direct individual process applications. The control philosophy developed for the equipment automation program stresses system flexibility and inherent levels of redundant control capabilities. Four different control points have been developed for each unit process operation

  5. Report of the collaboration project for research and development of sphere-pac fuel among JNC-PSI-NRG (1). Planning, fuel design, pin fabrication

    International Nuclear Information System (INIS)

    Morihira, Masayuki; Ozawa, Takayuki; Tomita, Yutaka; Suzuki, Masahiro; Kihara, Yoshiyuki; Shigetome, Yoshiaki; Kohno, Shusaku

    2004-07-01

    The collaboration project concerning sphere-pac fuel among JNC, Swiss PSI (Paul Scherrer Institut) and Dutch NRG (Nuclear Research and Consultancy Group) is in progress. Final target of the project is comparative irradiation tests of sphere-pac fuel in the HFR (High Flux Reactor) in Petten in the Netherlands with pellet type fuel and vipack fuel. Total 16 fuel segments (8 pins) of these three types of fuel are planned to be irradiated. Two sphere-pac fuel segments contain 5%Np in addition to 20%Pu-MOX. Other segments contain no Np. The objective of the irradiation tests is to obtain the restructuring data in the early beginning of life for SPF as well as power-to-melt test data for the potential study of SPF. At the same time introduction of modeling technique for irradiation performance analysis, fuel design, fuel fabrication is also important objective for JNC. Fabrication of irradiation test pins was completed till May 2003 in PSI. After transportation of the fuel pins to Petten, two times of irradiation were performed in January to March in 2004 and now post irradiation tests are in progress. Later two irradiations will be done till the autumn in 2004. This report summarized the basic plan, fuel design, and fabrication of irradiation test pins concerning this collaboration project. (author)

  6. Highlights of 50 years of nuclear fuel development

    International Nuclear Information System (INIS)

    Simnad, M.T.

    1989-01-01

    The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel-cycle costs and reliable performance in the fuel elements are discussed in the historical context. 10 refs

  7. Fabrication and characterization of fully ceramic microencapsulated fuels

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, K.A., E-mail: kurt.terrani@gmail.com [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kiggans, J.O.; Katoh, Y. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Shimoda, K. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Montgomery, F.C.; Armstrong, B.L.; Parish, C.M. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Hinoki, T. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hunn, J.D. [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, L.L. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2012-07-15

    The current generation of fully ceramic microencapsulated fuels, consisting of Tristructural Isotropic fuel particles embedded in a silicon carbide matrix, is fabricated by hot pressing. Matrix powder feedstock is comprised of alumina-yttria additives thoroughly mixed with silicon carbide nanopowder using polyethyleneimine as a dispersing agent. Fuel compacts are fabricated by hot pressing the powder-fuel particle mixture at a temperature of 1800-1900 Degree-Sign C using compaction pressures of 10-20 MPa. Detailed microstructural characterization of the final fuel compacts shows that oxide additives are limited in extent and are distributed uniformly at silicon carbide grain boundaries, at triple joints between silicon carbide grains, and at the fuel particle-matrix interface.

  8. MOX fuel fabrication technology in J-MOX

    International Nuclear Information System (INIS)

    Osaka, Shuichi; Yoshida, Ryouichi; Yamazaki, Yukiko; Ikeda, Hiroyuki

    2014-01-01

    Japan Nuclear Fuel Ltd. (JNFL) has constructed JNFL MOX Fuel Fabrication Plant (J-MOX) since 2010. The MIMAS process has been introduced in the powder mixing process from AREVA NC considering a lot of MOX fuel fabrication experiences at MELOX plant in France. The feed material of Pu for J-MOX is MH-MOX powder from Rokkasho Reprocessing Plant (RRP) in Japan. The compatibility of the MH-MOX powder with the MIMAS process was positively evaluated and confirmed in our previous study. This paper describes the influences of the UO2 powder and the recycled scrap powder on the MOX pellet density. (author)

  9. Chilean fuel elements fabrication progress report

    International Nuclear Information System (INIS)

    Baeza, J.; Contreras, H.; Chavez, J.; Klein, J.; Mansilla, R.; Marin, J.; Medina, R.

    1993-01-01

    Due to HEU-LEU core conversion necessity for the Chilean MTR reactors, the Fuel Elements Plant is being implemented to LEU nuclear fuel elements fabrication. A glove box line for powder-compact processing designed at CCHEN, which supposed to operate under an automatic control system, is at present under initial tests. Results of first natural uranium fuel plates manufacturing runs are shown

  10. Development of Melting Crucible Materials of Metallic Fuel Slug for SFR

    International Nuclear Information System (INIS)

    Kim, K. H.; Lee, C. T.; Oh, S. J.; Kim, S. K.; Lee, C. B.; Ko, Y. M.; Woo, W. M.

    2010-01-01

    The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr-Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized

  11. Mixed U/Pu oxide fabrication facility for gel-sphere-pac fuel

    International Nuclear Information System (INIS)

    1978-09-01

    This paper describes a conceptual plant which uses the gel-sphere-pac process to fabricate mixed oxide (MOX) fuel and covers (1) fabrication of co-processed MOX fuel and (2) fabrication of co-processed spiked MOX fuel, using 60 Co. The report describes: the fuel fabrication process and plant layout, including scrap and waste processing; and maintenance safety and ventilation measures. A description of the conversion of U and Pu nitrate using a gel sphere process is given in Appendix A

  12. Design study and evaluation of fuel fabrication systems for FR fuel cycle

    International Nuclear Information System (INIS)

    Namekawa, Takashi; Tanaka, Kenya; Kawaguchi, Koichi; Koike, Kazuhiro; Shimuta, Hiroshi; Suzuki, Yoshihiro

    2004-01-01

    The plant concept for each FBR fuel fabrication system has been constructed and evaluated, which achieves economical improvement, decrease in the environmental burden, better resource utilization, and proliferation resistance by the various innovative techniques employed. The results are as follows: (1) For oxide fuels, the simplified pelletizing method has a high technical feasibility, and it is possible to apply this method to practical process at early stage, because this method is based on wealth results of a conventional method. (2) For oxide fuels, the sphere packing fuel fabrication system by gelation and vibro-compaction processes has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the process. However this system shoulders additional cost for the liquid waste treatment process to dispose a large bulk of process liquid waste. (3) For the metal fuel, the casting system is generally expected to have high economical efficiency even for small-scale facilities, although verification for fabrication of the TRU alloy slug is required. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

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

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

    International Nuclear Information System (INIS)

    1985-01-01

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

  15. Fabrication details for wire wrapped fuel assembly components

    International Nuclear Information System (INIS)

    Bosy, B.J.

    1978-09-01

    Extensive hydraulic testing of simulated LMFBR blanket and fuel assemblies is being carried out under this MIT program. The fabrication of these test assemblies has involved development of manufacturing procedures involving the wire wrapped pins and the flow housing. The procedures are described in detail in the report

  16. Fabrication of oxide dispersion strengthened ferritic clad fuel pins

    International Nuclear Information System (INIS)

    Zirker, L.R.; Bottcher, J.H.; Shikakura, S.; Tsai, C.L.

    1991-01-01

    A resistance butt welding procedure was developed and qualified for joining ferritic fuel pin cladding to end caps. The cladding are INCO MA957 and PNC ODS lots 63DSA and 1DK1, ferritic stainless steels strengthened by oxide dispersion, while the end caps are HT9 a martensitic stainless steel. With adequate parameter control the weld is formed without a residual melt phase and its strength approaches that of the cladding. This welding process required a new design for fuel pin end cap and weld joint. Summaries of the development, characterization, and fabrication processes are given for these fuel pins. 13 refs., 6 figs., 1 tab

  17. Natural uranium metallic fuel elements: fabrication and operating experience

    International Nuclear Information System (INIS)

    Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

    1980-01-01

    The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

  18. Fabrication of FFTF fuel pin wire wrap

    International Nuclear Information System (INIS)

    Epperson, E.M.

    1980-06-01

    Lateral spacing between FFTF fuel pins is required to provide a passageway for the sodium coolant to flow over each pin to remove heat generated by the fission process. This spacing is provided by wrapping each fuel pin with type 316 stainless steel wire. This wire has a 1.435mm (0.0565 in.) to 1.448mm (0.0570 in.) diameter, contains 17 +- 2% cold work and was fabricated and tested to exacting RDT Standards. About 500 kg (1100 lbs) or 39 Km (24 miles) of fuel pin wrap wire is used in each core loading. Fabrication procedures and quality assurance tests are described

  19. Fabrication of internally instrumented reactor fuel rods

    International Nuclear Information System (INIS)

    Schmutz, J.D.; Meservey, R.H.

    1975-01-01

    Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

  20. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G. [Oak Ridge National Lab., TN (United States); Carrell, R.D. [Technical Resources International, Inc., Richland, WA (United States); Jaeger, C.D. [Sandia National Labs., Albuquerque, NM (United States); Thompson, M.L.; Strasser, A.A. [Delta-21 Resources, Inc., Oak Ridge, TN (United States)

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET.

  1. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    International Nuclear Information System (INIS)

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O'Connor, D.G.; Carrell, R.D.; Jaeger, C.D.; Thompson, M.L.; Strasser, A.A.

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET

  2. The DUPIC fuel development program in KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yang, M S; Park, H S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    This study describes the DUPIC fuel development program in KAERI as follows; Burning spent PWR fuel again in CANDU by DUPIC, Compatibility with existing CANDU system, Feasibility of DUPIC fuel fabrication, Waste reduction, Safeguard ability, Economics of DUPIC fuel cycle, The DUPIC fuel development program, and International prospective. 5 refs., 10 figs.

  3. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period

  4. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period.

  5. Fuel fabrication and post-irradiation examination

    Energy Technology Data Exchange (ETDEWEB)

    Venter, P J; Aspeling, J C [Atomic Energy Corporation of South Africa Ltd., Pretoria (South Africa)

    1990-06-01

    This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)

  6. Fuel fabrication and post-irradiation examination

    International Nuclear Information System (INIS)

    Venter, P.J.; Aspeling, J.C.

    1990-01-01

    This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  8. Zero risk fuel fabrication: a systems analysis

    International Nuclear Information System (INIS)

    1979-01-01

    Zero risk is a concept used to ensure that system requirements are developed through a systems approach such that the choice(s) among alternatives represents the balanced viewpoints of performance, achievability and risk. Requirements to ensure characteristics such as stringent accountability, low personnel exposure and etc. are needed to guide the development of component and subsystems for future LMFBR fuel supply systems. To establish a consistent and objective set of requirements, RF and M-TMC has initiated a systems requirements analysis activity. This activity pivots on judgement and experience provided by a Task Force representing industrial companies engaged in fuel fabrication in licensed facilities. The Task Force members are listed in Appendix A. Input developed by this group is presented as a starting point for the systems requirements analysis

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  10. Control in fabrication of PWR and BWR type reactor fuel elements

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1981-01-01

    Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

  11. Manufacturing experience and perspectives of WWER nuclear fuel development

    International Nuclear Information System (INIS)

    Aksenov, P.; Kolosovskiy, V.

    2011-01-01

    The purposes of new shroudless working fuel assembly (PK-3) development, basic design peculiarities of working fuel assembly (PK-3) and the results of PK-3 implementation are presented in this paper. Values of 440.19.000-02 working fuel assembly with debris filter Burnup at Kola NPP unit 2 are given. The main issues settled in the course of TVSA-T implementation like: The development of the design and fabrication method of mixing grids; The development of the design and fabrication method of basic assemblies and components of TVSA-T, including fuel rods of new generation; and The obtainment of specified pellet microstructure with average grain size more than 25μm are listed. The development of the design and fabrication method of removable uprated headpiece of shortened length as well as the development of the design and fabrication method of a tailpiece equipped with a debris filter are also illustrated

  12. Remote fabrication of breeder reactor fuel

    International Nuclear Information System (INIS)

    Gerber, E.W.; Hoitink, N.C.; Graham, R.A.

    1984-06-01

    The Secure Automated Fabrication (SAF) Line, a remotely operable plutonium fuel fabrication facility, incorporates advanced automation techniques. Included in the plant are 24 robots used to perform complex operations, and to enhance equipment standardization and ease of maintenance. Automated equipment is controlled remotely from centrally located supervisory computer control consoles or alternatively from control consoles dedicated to individual systems

  13. Quality control in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Abdelhalim, A.S.; Elsayed, A.A.; Shaaban, H.I.

    1988-01-01

    The department of metallurgy, NRC Inchass is embarking on a programme of on a laboratory scale, fuel pins containing uranium dioxide pellets are going to be produced. The department is making use of the expertise and equipment at present available and is going to utilize the new fuel pin fabrication unit which would be shortly in operation. The fabrication and testing of uranium dioxide pellets then gradually adapt them and develop, a national know how in this field. This would also involve building up of indigenous experience through proper training of qualified personnel. That are applied to ensure quality of U o 2 pellets, the techniques implemented, the equipment used and the specifications of the equipment presently available. The following parameters are subject to quality control tests: density. O/U ration, hydrogen content, microstructure, each property will be discussed, measurements related to U o 2 powders, including flow ability, bulk density, O/U ratio, bet surface area and water content will be critically discussed. Relevant tests to ensure Q C of pellets are reviewed. These include surface integrity, density, dimensions, microstructure.4 fig., 1 tab

  14. High performance fuel technology development

    Energy Technology Data Exchange (ETDEWEB)

    Koon, Yang Hyun; Kim, Keon Sik; Park, Jeong Yong; Yang, Yong Sik; In, Wang Kee; Kim, Hyung Kyu [KAERI, Daejeon (Korea, Republic of)

    2012-01-15

    {omicron} Development of High Plasticity and Annular Pellet - Development of strong candidates of ultra high burn-up fuel pellets for a PCI remedy - Development of fabrication technology of annular fuel pellet {omicron} Development of High Performance Cladding Materials - Irradiation test of HANA claddings in Halden research reactor and the evaluation of the in-pile performance - Development of the final candidates for the next generation cladding materials. - Development of the manufacturing technology for the dual-cooled fuel cladding tubes. {omicron} Irradiated Fuel Performance Evaluation Technology Development - Development of performance analysis code system for the dual-cooled fuel - Development of fuel performance-proving technology {omicron} Feasibility Studies on Dual-Cooled Annular Fuel Core - Analysis on the property of a reactor core with dual-cooled fuel - Feasibility evaluation on the dual-cooled fuel core {omicron} Development of Design Technology for Dual-Cooled Fuel Structure - Definition of technical issues and invention of concept for dual-cooled fuel structure - Basic design and development of main structure components for dual- cooled fuel - Basic design of a dual-cooled fuel rod.

  15. Re-qualification of MTR-type fuel plates fabrication process

    International Nuclear Information System (INIS)

    Elseaidy, I.M.; Ghoneim, M.M.

    2010-01-01

    The fabricability issues with increased uranium loading due to use low enrichment of uranium (LEU), i.e. less than 20 % of U 235 , increase the problems which occur during compact manufacturing, roll bonding of the fuel plates, potential difficulty in forming during rolling process, mechanical integrity of the core during fabrication, potential difficulty in meat homogeneity, and the ability to fabricate plates with thicker core as a means of increasing total uranium loading. To produce MTR- type fuel plates with high uranium loading (HUL) and keep the required quality of these plates, many of qualification process must be done in the commissioning step of fuel fabrication plant. After that any changing of the fabrication parameters, for example changing of any of the raw materials, devises, operators, and etc., a re- qualification process should be done in order to keep the quality of produced plates. Objective of the present work is the general description of the activities to be accomplished for re-qualification of manufacturing MTR- type nuclear fuel plates. For each process to be re-qualified, a detailed of re-qualification process were established. (author)

  16. Developments in fuel manufacturing

    International Nuclear Information System (INIS)

    Ion, S.E.; Harrop, G.; Maricalva Gonzalez, J.

    1995-01-01

    The status of the investment and R and D programmes in the UK and Spanish fuel fabrication facilities is outlined. Due to a number of circumstances, BNFL and ENUSA have been in the forefront of capital investment, with associated commitment to engineering and scientific research and development. Carrying through this investment has allowed the embodiment of proven state of the art technologies in the design of fuel fabrication plants, with particular emphasis on meeting the future challenge of health and safety, and product quality, at an acceptable cost. ENUSA and BNFL currently supply fuel, not only to their respective 'home' markets but also to France, Belgium, Sweden, and Germany. Both organisations employ an International Business outlook and partake in focused and speculative R and D projects for the design and manufacture of nuclear fuel. (orig./HP)

  17. Experiences in transferring of AFA 3G fuel assembly fabrication

    International Nuclear Information System (INIS)

    Yang Xiaodong; Wu Zhiming; Luo Jiankang

    2002-01-01

    Implementation program is developed for the transferring of AFA 3G technology, together with the project management experts designated by Framatome Company, to facilitate the technology import under the guidance of strict program. Technical documents and quality insurance management documents are developed based on the full understanding of the information provided by Framatome to guide the fabrication of AFA 3G fuel elements. Technical requirement suggested by Framatome is adopted as much as possible, considering the practical process capability of YFP. The focus is the technology about fabrication difficulties in the AFA 3G technology, to insure the successful transfer of the AFA 3G fabrication technology

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

    International Nuclear Information System (INIS)

    He Jun; Zou Yanwen; Liang Tongxiang; Qiu Xueliang

    2002-01-01

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

  19. Advanced accountability techniques for breeder fuel fabrication facilities

    International Nuclear Information System (INIS)

    Bennion, S.I.; Carlson, R.L.; DeMerschman, A.W.; Sheely, W.F.

    1978-01-01

    The United States Department of Energy (DOE) has assigned the Hanford Engineering Development Laboratory (HEDL), operated by the Westinghouse Hanford Company, the project lead in developing a uniform nuclear materials reporting system for all contractors on the Hanford Reservation. The Hanford Nuclear Inventory System (HANISY) is based upon HEDL's real-time accountability system, originally developed in 1968. The HANISY system will receive accountability data either from entry by process operators at remote terminals or from nondestructive assay instruments connected to the computer network. Nuclear materials will be traced from entry, through processing to final shipment through the use of minicomputer technology. Reports to DOE will be formed directly from the realtime files. In addition, HEDL has established a measurement program that will complement the HANISY system, providing direct interface to the computer files with a minimum of operator intervention. This technology is being developed to support the High Performance Fuels Laboratory (HPFL) which is being designed to assess fuel fabrication techniques for proliferation-resistant fuels

  20. Advanced LWR Nuclear Fuel Cladding Development

    International Nuclear Information System (INIS)

    Bragg-Sitton, S.; Griffith, G.

    2012-01-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

  1. Fabrication of metallic fuel for fast breeder reactor

    International Nuclear Information System (INIS)

    Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arbind; Mittal, R.K.; Prasad, R.S.; Mahule, N.; Kumar, Arun; Prasad, G.J.

    2012-01-01

    Natural uranium oxide fuelled PHWRs comprises of first stage of Indian nuclear power programme. Liquid metal fast breeder reactors fuelled by Pu (from PHWR's) form the second stage. A shorter reactor doubling time is essential in order to accelerate the nuclear power growth in India. Metallic fuels are known to provide shorter doubling times, necessitating to be used as driver fuel for fast breeder reactors. One of the fabrication routes for metallic fuels having random grain orientation, is injection casting technique. The technique finds its basis in an elementary physical concept - the possibility of supporting a liquid column within a tube, by the application of a pressure difference across the liquid interface inside and outside the tube. At AFD, BARC a facility has been set-up for injection casting of uranium rods in quartz tube moulds, demoulding of cast rods, end-shearing of rods and an automated inspection system for inspection of fuel rods with respect to mass, length, diameter and diameter variation along the length and internal and external porosities/voids. All the above facilities have been set-up in glove boxes and have successfully been used for fabrication of uranium bearing fuel rods. The facility has been designed for fabrication and inspection of Pu-bearing metallic fuels also, if required

  2. Fabrication of cermet fuel for fast reactor

    International Nuclear Information System (INIS)

    Mishra, Sudhir; Kumar, Arun; Kutty, T.R.G.; Kamath, H.S.

    2011-01-01

    Mixed oxide (MOX) (U,Pu)O 2 , and metallic (U,Pu ,Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity , low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion. The higher coefficient of linear expansion is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burnup, fuel cladding interaction and lower margin between operating and melting temperature. The optimal solution may lie in cermet fuel (U, PuO 2 ), where PuO 2 is dispersed in U metal matrix and combines the favorable features of both the fuel types. The advantages of this fuel include high thermal conductivity, larger margin between melting and operating temperature, ability to retain fission product etc. The matrix being of high density metal the advantage of high breeding ratio is also maintained. In this report some results of fabrication of cermet pellet comprising of UO 2 /PuO 2 dispersed in U metal powder through classical powder metallurgy route and characterization are presented. (author)

  3. Prototypic fabrication of TRIGA irradiated fuel shipping casks

    International Nuclear Information System (INIS)

    Kim, B.K.; Lee, Y.W.; Whang, C.K.; Lee, J.B.

    1980-01-01

    This is the safety analysis report on the prototypic fabrication of ''TRIGA Irradiated Fuel Shipping Cask'' conducted by KAERI in 1980. The results of the evaluation show that the shipping cask is in compliance with the applicable regulation for the normal conditions of transport as well as hypothetical accident conditions. The prototypic fabrication of the shipping cask (type B) was carried out for the first time in Korea after getting technical experience from fabrication of the ''TRIGA Spent Fuel Shipping Cask'' and ''the KO-RI Unit 1 surveillance capsule shipping cask'' in 1979. This report contains structural evaluation, thermal evaluation, shielding, criticality, quality assurance, and handling procedures of the shipping cask

  4. Product Conversion: The Link between Separations and Fuel Fabrication

    International Nuclear Information System (INIS)

    Felker, L.K.; Vedder, R.J.; Walker, E.A.; Collins, E.D.

    2008-01-01

    Several chemical processing flowsheets are under development for the separation and isolation of the actinide, lanthanide, and fission product streams in spent nuclear fuel. The conversion of these product streams to solid forms, typically oxides, is desired for waste disposition and recycle of product fractions back into transmutation fuels or targets. The modified direct denitration (MDD) process developed at Oak Ridge National Laboratory (ORNL) in the 1980's offers significant advantages for the conversion of the spent fuel products to powder form suitable for direct fabrication into recycle fuels. A glove-box-contained MDD system and a fume-hood-contained system have been assembled at ORNL for the purposes of testing the co-conversion of uranium and mixed-actinide products. The current activities are focused on the conversion of the first products from the processing of spent nuclear fuel in the Coupled End-to-End Demonstration currently being conducted at ORNL. (authors)

  5. Fabrication routes for Thorium and Uranium233 based AHWR fuel

    International Nuclear Information System (INIS)

    Danny, K.M.; Saraswat, Anupam; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun

    2011-01-01

    India's economic growth is on a fast growth track. The growth in population and economy is creating huge demand for energy which has to be met with environmentally benign technologies. Nuclear Energy is best suited to meet this demand without causing undue environmental impact. Considering the large thorium reserves in India, the future nuclear power program will be based on Thorium- Uranium 233 fuel cycle. The major characteristic of thorium as the fuel of future comes from its superior fuel utilization. 233 U produced in a reactor is always contaminated with 232 U. This 232 U undergoes a decay to produce 228 Th and it is followed by decay chain including 212 Bi and 208 Tl. Both 212 Bi and 208 Tl are hard gamma emitters ranging from 0.6 MeV-1.6 MeV and 2.6 MeV respectively, which necessitates its handling in hot cell. The average concentration of 232 U is expected to exceed 1000 ppm after a burn-up of 24,000 MWD/t. Work related to developing the fuel fabrication technology including automation and remotization needed for 233 U based fuels is in progress. Various process for fuel fabrication have been developed i.e. Coated Agglomerate Pelletisation (CAP), impregnation technique (Pellet/Gel), Sol Gel Micro-sphere Pelletisation (SGMP) apart from Powder to Pellet (POP) route. This paper describes each process with respect to its advantages, disadvantages and its amenability to automation and remotisation. (author)

  6. Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

    2008-09-01

    The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

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

    Science.gov (United States)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar

    2012-01-01

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

  8. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1985-01-01

    In the Law for the Regulations of Nuclear Source Material, Nuclear Fuel Material and Reactors, the regulations have all been revised on the fabrication business of nuclear fuel materials. The revised regulations are given : application for permission of the fabrication business, application for permission of the alteration, application for approval of the design and the construction methods, application for approval of the alteration, application for the facilities inspection, facilities inspection, recordings, entry limitations etc. for controlled areas, measures concerning exposure radiation doses etc., operation of the fabrication facilities, transport within the site of the business, storage, disposal within the site of the business, security regulations, designation etc. of the licensed engineer of nuclear fuels, collection of reports, etc. (Mori, K.)

  9. Gas Tungsten Arc Welding for Fabrication of SFR Fuel Rodlet

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Woo, Yoon Myeng; Kim, Bong Goo; Park, Jeong Yong; Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    To evaluate the PGSFR fuel performance, the irradiation test in HANARO research reactor was planned and the fuel rodlet to be used for irradiation test should be fabricated under the appropriate Quality Assurance (QA) program. For the fabrication of PGSFR metallic fuel rodlets, the end plug welding is a crucial process. The sealing of end plug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the end plug welding of fuel rodlet for irradiation test in HANARO was carried out based on the qualified welding technique as reported in the previous paper. The end plug welding of fuel rodlets for irradiation test in HANARO was successfully carried out under the appropriate QA program. The results of the quality inspections on the end plug weld satisfied well the quality criteria on the weld. Consequently the fabricated fuel rodlets are ready for irradiation test in HANARO.

  10. Research reactor fuel development at AECL

    International Nuclear Information System (INIS)

    Sears, D.F.; Wang, N.

    2000-09-01

    This paper reviews recent U 3 Si 2 and U-Mo dispersion fuel development activities at AECL. The scope of work includes fabrication development, irradiation testing, post-irradiation examination and performance qualification. U-Mo alloys with a variety of compositions, ranging from 6 to 10 wt % Mo, have been fabricated with high purity and homogeneity in the product. The alloys and powders were characterized using optical and scanning electron microscopy, chemical analysis, and X-ray diffraction and neutron diffraction analysis. U-Mo powder samples have been supplied to the Argonne National Laboratory for irradiation testing in the ATR reactor. Low-enriched uranium fuel elements containing U-7 wt % Mo and U-10 wt % Mo with loadings up to 4.5 gU/cm 3 have been fabricated at CRL for irradiation testing in the NRU reactor. The U-Mo fuel elements will be tested in NRU at linear powers up to 145 kW/m, and to 85 atom % 235 U burnup. (author)

  11. Research reactor fuel development at AECL

    International Nuclear Information System (INIS)

    Sears, D.F.; Wang, N.

    2000-01-01

    This paper reviews recent U 3 Si 2 and U-Mo dispersion fuel development activities at AECL. The scope of work includes fabrication development, irradiation testing, postirradiation examination and performance qualification. U-Mo alloys with a variety of compositions, ranging from 6 to 10 wt % Mo, have been fabricated with high purity and homogeneity in the product. The alloys and powders were characterized using optical and scanning electron microscopy, chemical analysis, and X-ray diffraction and neutron diffraction analysis. U-Mo powder samples have been supplied to the Argonne National Laboratory for irradiation testing in the ATR reactor. Low-enriched uranium fuel elements containing U-7 wt % Mo and U-10 wt % Mo with loadings up to 4.5 gU/cm 3 have been fabricated at CRL for irradiation testing in the NRU reactor. The U-Mo fuel elements will be tested in NRU at linear powers up to 145 kW/m, and to 85 atom % 235 U burnup. (author)

  12. Research on plant of metal fuel fabrication using casting process

    International Nuclear Information System (INIS)

    Senda, Yasuhide; Mori, Yukihide

    2003-12-01

    This document presents the plant concept of metal fuel fabrication system (38tHM/y) using casting process in electrolytic recycle, which based on recent studies of its equipment design and quality control system. And we estimate the cost of its construction and operation, including costs of maintenance, consumed hardware and management of waste. The content of this work is as follows. (1) Designing of fuel fabrication equipment: We make material flow diagrams of the fuel fabrication plant and rough designs of the injection casting furnace, demolder and inspection equipment. (2) Designing of resolution system of liquid waste, which comes from analytical process facility. Increased analytical items, we rearrange analytical process facility, estimate its chemicals and amount of waste. (3) Arrangement of equipments: We made a arrangement diagram of the metal fuel fabrication equipments in cells. (4) Estimation of cost data: We estimated cost to construct the facility and to operate it. (author)

  13. Competition still fierce in the US fuel fabrication market

    International Nuclear Information System (INIS)

    Schwartz, M.H.

    1990-01-01

    The US market for nuclear fuel fabrication services is characterized by an annual production capacity significantly in excess of both current and anticipated demand. The trends toward longer operating cycle lengths and higher burnup fuel continue in the United States. This, together with the lack of any prospects for new light water reactors coming on line in the US during the next ten years, is expected to hold the annual demand for fuel fabrication services from US LWRs at around 2000t of uranium into the next century. (author)

  14. Development of metallic fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Ho; Lee, Chong Yak; Lee, Myung Ho and others

    1999-03-01

    With the vacuum melting and casting of the U-10wt%Zr alloy which is metallic fuel for liquid metal fast breeder reactor, we studied the microstructure of the alloy and the parameters of the melting and casting for the fuel rods. Internal defects of the U-10wt%Zr fuel by gravity casting, were inspected by non-destructive test. U-10wt%Zr alloy has been prepared for the thermal stability test in order to estimate the decomposition of the lamellar structure with relation to swelling under irradiation condition. (author)

  15. Results of fuel elements fabrication on the basis of increased concentration dioxide fuel for research reactors

    International Nuclear Information System (INIS)

    Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

    1996-01-01

    According to the Russian Reduced Enrichment for Research and Test Reactors (RERTR) program, that were constructed under the Russian projects, at the Novosibirsk Chemical Concentrates Plant the pilot series of different configuration (WR-M2, MR, IRT-4M) fuel elements, based on increased concentration uranium dioxide fuel, have been fabricated for reactor tests. Comprehensive fabricated fuel elements quality estimation has been carried out. (author)

  16. Fabrication drawings of fuel pins for FUJI project among PSI, JNC and NRG. Revised version

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Nagayama, Masahiro

    2002-02-01

    Irradiation tests and post-irradiation examinations in the framework of JNC-PSI-NRG collaboration project will be performed in 2003-2005. Irradiation fuel pins will be fabricated by the middle of 2003. The fabrication procedure for irradiation fuel pins has been started in 2001. Several fabrication tests and qualification tests in JNC and PSI (Paul Scherrer Institute, Switzerland) have been performed before the fuel pin fabrication. According to the design assignment between PSI and JNC in the frame of this project, PSI should make a specification document for the fuel pellet, the sphere-pac fuel particles, the vipac fuel particles, and the fuel pin. JNC should make a fabrication drawing for irradiation pins. JNC has been performed the fuel design in cooperation with PSI and NRG (Nuclear Research and Consultancy Group, Netherlands). In this project, the pelletized fuel, the sphere-pac fuel, and the vipac fuel will be simultaneously irradiated on HFR (High Flux Reactor, Netherlands). This fabrication drawing has been made under the design assignment with PSI, and consists of the drawing of MOX pellet, thermal insulator pellet, pin components, fuel segments, and the constructed pin. The fabrication drawings were approved in October 2001, but after that, the optimization of specifications has been discussed and agreed among all partners. In this report, the revised fabrication drawings will be shown. Based on the commission of Plutonium Fuel Technology Group, Advanced Fuel Recycle Technology Division, this design work has been performed in Fuel Design and Evaluation Group, Plutonium Fuel Fabrication Division, Plutonium Fuel Center. (author)

  17. Greenfield Alternative Study LEU-Mo Fuel Fabrication Facility

    Energy Technology Data Exchange (ETDEWEB)

    Washington Division of URS

    2008-07-01

    This report provides the initial “first look” of the design of the Greenfield Alternative of the Fuel Fabrication Capability (FFC); a facility to be built at a Greenfield DOE National Laboratory site. The FFC is designed to fabricate LEU-Mo monolithic fuel for the 5 US High Performance Research Reactors (HPRRs). This report provides a pre-conceptual design of the site, facility, process and equipment systems of the FFC; along with a preliminary hazards evaluation, risk assessment as well as the ROM cost and schedule estimate.

  18. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P.O.1236909. Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-12-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design

  19. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P. O. 1236909. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-12-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design.

  20. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    International Nuclear Information System (INIS)

    Sample, C.R.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL

  1. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  2. Oxide fuel fabrication technology development of the FaCT project (5). Current status on 9Cr-ODS steel cladding development for high burn-up fast reactor fuel

    International Nuclear Information System (INIS)

    Ohtsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya

    2011-01-01

    This paper describes evaluation results of in-reactor integrity of 9Cr and 12Cr-ODS steel cladding tubes and the plan for reliability improvement in homogeneous tube production, both of which are key points for the commercialized use of ODS steels as long-life fuel cladding tubes. A fuel assembly in the BOR-60 irradiation test including 9Cr and 12Cr-ODS fuel pins has achieved the highest burn-up, i.e. peak burn-up of 11.9at% and peak neutron dose of 51dpa, without any fuel pin rupture and microstructure instability. In another fuel assembly containing 9Cr and 12Cr-ODS steel fuel pins whose peak burn-up was 10.5at%, one 9Cr-ODS steel fuel pin failed near the upper end of the fuel column. A peculiar microstructure change occurred in the vicinity of the ruptured area. The primary cause of this fuel pin rupture and microstructure change was shown to be the presence of metallic Cr inclusions in the 9Cr-ODS steel tube, which had passed an ultrasonic inspection test for defects. In the next stage from 2011 to 2013, the fabrication technology of full pre-alloy 9Cr-ODS steel cladding tube will be developed, where the handling of elemental powder is prohibited in the process. (author)

  3. Coated fuel particles: requirements and status of fabrication technology

    International Nuclear Information System (INIS)

    Huschka, H.; Vygen, P.

    1977-01-01

    Fuel cycle, design, and irradiation performance requirements impose restraints on the fabrication processes. Both kernel and coating fabrication processes are flexible enough to adapt to the needs of the various existing and proposed high-temperature gas-cooled reactors. Extensive experience has demonstrated that fuel kernels with excellent sphericity and uniformity can be produced by wet chemical processes. Similarly experience has shown that the various multilayer coatings can be produced to fully meet design and specification requirements. Quality reliability of coated fuel particles is ensured by quality control and quality assurance programs operated by an aduiting system that includes licensing officials and the customer

  4. Babcock and Wilcox plate fabrication experience with uranium silicide spherical fuel

    International Nuclear Information System (INIS)

    Todd, Lawrence E.; Pace, Brett W.

    1996-01-01

    This report is written to present the fuel fabrication experience of Babcock and Wilcox using atomized spherical uranium silicide powder. The intent is to demonstrate the ability to fabricate fuel plates using spherical powder and to provide useful information proceeding into the next phase of work using this type of fuel. The limited quantity of resources- spherical powder and time, did not allow for much process optimizing in this work scope. However, the information contained within provides optimism for the future of spherical uranium silicide fuel plate fabrication at Babcock and Wilcox.The success of assembling fuel elements with spherical powder will enable Babcock and Wilcox to reduce overall costs to its customers while still maintaining our reputation for providing high quality research and test reactor products. (author)

  5. PHWR fuel fabrication with imported uranium - procedures and processes

    International Nuclear Information System (INIS)

    Rao, R.V.R.L.V.; Rameswara Rao, A.; Hemantha Rao, G.V.S.; Jayaraj, R.N.

    2010-01-01

    Following the 123 agreement and subsequent agreements with IAEA & NSG, Government of India has entered into bilateral agreements with different countries for nuclear trade. Department of Atomic Energy (DAE), Government of India, has entered into contract with few countries for supply of uranium material for use in the safeguarded PHWRs. Nuclear Fuel Complex (NFC), an industrial unit of DAE, established in the early seventies, is engaged in the production of Nuclear Fuel and Zircaloy items required for Nuclear Power Reactors operating in the country. NFC has placed one of its fuel fabrication facilities (NFC, Block-A, INE-) under safeguards. DAE has opted to procure uranium material in the form of ore concentrate and fuel pellets. Uranium ore concentrate was procured as per the ASTM specifications. Since no international standards are available for PHWR fuel pellets, Specifications have to be finalized based on the present fabrication and operating experience. The process steps have to be modified and fine tuned for handling the imported uranium material especially for ore concentrate. Different transportation methods are to be employed for transportation of uranium material to the facility. Cost of the uranium material imported and the recoveries at various stages of fuel fabrication have impact on the fuel pricing and in turn the unit energy costs. Similarly the operating procedures have to be modified for safeguards inspections by IAEA. NFC has successfully manufactured and supplied fuel bundles for the three 220 MWe safeguarded PHWRs. The paper describes various issues encountered while manufacturing fuel bundles with different types of nuclear material. (author)

  6. Prototype fuel fabrication for nuclear reactors of Laguna Verde

    International Nuclear Information System (INIS)

    Nocetti, C.; Torres, J.; Medrano, A.

    1996-01-01

    Four prototype fuel bundles for the Laguna Verde Nuclear Power Plant have been fabricated. the type of nuclear fuel produced is described and the process used is commented. As an example of the fabrication criteria adopted, the production model to determine the density of the U O 2 pellets for the different batches of ceramic powder is described. the results are evaluated using the statistical indexes C p and C pk . (author)

  7. Review of qualifications for fuel assembly fabrication

    International Nuclear Information System (INIS)

    Slabu, Dan; Zemek, Martin; Hellwig, Christian

    2013-01-01

    The required quality of nuclear fuel in industrial production can only be assured by applying processes in fabrication and inspection, which are well mastered and have been proven by an appropriate qualification. The present contribution shows the understanding and experiences of Axpo with respect to qualifications in the frame of nuclear fuel manufacturing and reflects some related expectations of the operator. (orig.)

  8. Powder metallurgy and fabricating processes of cermet and metmet fuel in Russia

    International Nuclear Information System (INIS)

    Vatulin, A.; Konovalov, I.; Savchenco, A.; Stetsky, Y.; Trifonov, Y.; Bochvar, A.A.

    2000-01-01

    Methods of powder metallurgy are widely used for manufacturing of various components of reactor core: beryllium reflectors, absorbers, parts of controlling and safety systems, fuel pellets for fuel elements of power reactors and etc. The new problems arising before atomic engineering associated with increasing requirements to safe operation of reactors, non-proliferation of the nuclear weapons and utilization of plutonium stockpile in the world, served as a push to development of new kinds of dispersion nuclear fuel CERMET, CERCER, METMET. The bases of fabricating processes of such compositions are the methods of powder metallurgy. In this report some results of research activities on the development of new kinds of CERMET and METMET fuel and fuel elements for different type reactors are presented. (author)

  9. Special equipment for the fabrication and quality control of nuclear fuel elements

    International Nuclear Information System (INIS)

    Guse, K.; Herbert, W.; Jaeger, K.

    1989-01-01

    For the fabrication of LWR fuel elements, columns are packed of up to 4 m in length, consisting of fuel pellets with different uranium enrichment, their weight and total length to be measured prior to further processing to fuel rods. An automated column packing device has been developed for this purpose. The packing jobs and other tasks are computer-controlled, measured data are stored and are printed out for quality documentation. The forces in the springs of fuel spacers of LWR fuel elements are to be measured and compared with the standard requirements, deviations to be documented. For this task, another computer-controlled, automated device has been developed for measuring the spring forces at all required positions after positioning and fixation of the spacers. (orig./DG) [de

  10. Subsoil exploration of the estimated building site for nuclear fuel development and fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Song, In Taek [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    The objective of this report, as the result of subsoil exploration, is to provide basic design data of structural plan for nuclear fuel development and fabrication facility that is builded on Duckjin 150, Yusong, Taejeon, Korea, and provide basic data for execution of work. The soft rock level of estimated building site is deep(18.0m:BH-1, 20.5m:BH-2, 25.5m:BH-3) and the hard rock level of it is very deep (33.0m:BH-1, 46.0m:BH-2, 34.5m:BH-3) , for structural design, the hard rock shall be the bottom of foundation. 9 figs., 19 tabs. (Author)

  11. Direct fabrication of 238PuO2 fuel forms

    International Nuclear Information System (INIS)

    Burney, G.A.; Congdon, J.W.

    1982-07-01

    The current process for the fabrication of 238 PuO 2 heat sources includes precipitation of small particle plutonium oxalate crystals (4 to 6 μm diameter), a calcination to PuO 2 , ball milling, cold pressing, granulation (60 to 125 μm), and granule sintering prior to hot pressing the fuel pellet. A new two-step direct-strike Pu(III) oxalate precipitation method which yields mainly large well-developed rosettes (50 to 100 μm diameter) has been demonstrated in the laboratory and in the plant. These large rosettes are formed by agglomeration of small (2 to 4 μm) crystals, and after calcining and sintering, were directly hot pressed into fuel forms, thus eliminating several of the powder conditioning steps. Conditions for direct hot pressing of the large heat-treated rosettes were determined and a full-scale General Purpose Heat Source pellet was fabricated. The pellet had the desired granule-type microstructure to provide dimensional stability at high temperature. 27 figures

  12. Development of alternative fuel for pressurized water reactors

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  13. Integrated fuel processor development

    International Nuclear Information System (INIS)

    Ahmed, S.; Pereira, C.; Lee, S. H. D.; Krumpelt, M.

    2001-01-01

    The Department of Energy's Office of Advanced Automotive Technologies has been supporting the development of fuel-flexible fuel processors at Argonne National Laboratory. These fuel processors will enable fuel cell vehicles to operate on fuels available through the existing infrastructure. The constraints of on-board space and weight require that these fuel processors be designed to be compact and lightweight, while meeting the performance targets for efficiency and gas quality needed for the fuel cell. This paper discusses the performance of a prototype fuel processor that has been designed and fabricated to operate with liquid fuels, such as gasoline, ethanol, methanol, etc. Rated for a capacity of 10 kWe (one-fifth of that needed for a car), the prototype fuel processor integrates the unit operations (vaporization, heat exchange, etc.) and processes (reforming, water-gas shift, preferential oxidation reactions, etc.) necessary to produce the hydrogen-rich gas (reformate) that will fuel the polymer electrolyte fuel cell stacks. The fuel processor work is being complemented by analytical and fundamental research. With the ultimate objective of meeting on-board fuel processor goals, these studies include: modeling fuel cell systems to identify design and operating features; evaluating alternative fuel processing options; and developing appropriate catalysts and materials. Issues and outstanding challenges that need to be overcome in order to develop practical, on-board devices are discussed

  14. Fabrication of the Spent Fuel Elements Rack on the ISFSF

    International Nuclear Information System (INIS)

    Slamet Wiranto; Sigit Purwanto; Safrul, H.

    2004-01-01

    The Interim Storage For Spent Fuel elements (ISFSF) was designed to be able to store the 33 spent fuel element racks with capacity of 1386 of normal spent fuel elements and 2 racks for 36 of defected ones. Until now, only 9 out of 33 racks of normal spent fuel elements and lout of 2 racks of defected fuel elements are available. Five of them have suffered from corrosion so that they are not fulfilled the requirements of the spent fuel elements storage anymore. Meanwhile, the spent fuel storage racks in the reactor are almost full. It means, the transfer of the spent fuel from reactor spent fuel storage to the ISFSF pool are compulsory needed. Therefore, it is necessary to provide the new ISFSF spent fuel storage rack with better material and fabrication method than the old one. In this design all materials consist of SS 316 L that are welded with the Argon TIG-welding. Right now there has been one new spent fuel storage rack fabricated with capacity of 42 normal spent fuel elements. (author)

  15. Design of a quality assurance system in the nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Garcia Rojas Palacios, L.

    1992-01-01

    A)For the first time a project on nuclear fuel fabrication is going to be lead in this country. For this reason the work is oriented to establish a quality assurance system for the different stages of fuel fabrication. C) The work of this thesis was developed first by means of an analysis of quality philosophies of Deming, Ishikawa, Juran and Crosby from which several important points were stracted to be used in the designed quality system. Metrology and normalization are so important for quality control that a study of them is made considering definitions, unit systems and type of errors (for Metrology) as well as standards for quality systems, qualification, destructive and non destructive tests, shipment, packing for nuclear power plants. With the standards as a basis, the working strategy for the system was reached, as well as the design of control cards and the design of documents for inspection control, personnel and its documentation and finally the diagrams for each one of the fabrication stages

  16. Fabrication drawings of fuel pins for FUJI project among PSI, JNC and NRG. Revised version 2

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Nagayama, Masahiro

    2002-10-01

    Irradiation tests and post-irradiation examinations in the framework of JNC-PSI-NRG collaboration project will be performed in 2003-2005. Irradiation fuel pins will be fabricated by the middle of 2003. The fabrication procedure for irradiation fuel pins has been started in 2001. Several fabrication tests and qualification tests in JNC and PSI (Paul Scherrer Institut, Switzerland) have been performed before the fuel pin fabrication. According to the design assignment between PSI and JNC in the frame of this project, PSI should make specification documents for the fuel pellet, the sphere-pac fuel particles, the vipac fuel fragments, and the fuel segment fabrication. JNC should make the fabrication drawings for irradiation pins. JNC has been performed the fuel design in cooperation with PSI and NRG (Nuclear Research and Consultancy Group, Holland). In this project, the pelletized fuel, the sphere-pac fuel, and the vipac fuel will be simultaneously irradiated on HFR (High Flux Reactor, Holland). The fabrication drawings have been made under the design assignment with PSI, and consist of the drawings of MOX pellet, thermal insulator pellet, pin components, fuel segments, and the constructed pin. The fabrication drawings were approved in October 2001, but after that, the optimization of specifications was discussed and agreed among all partners. According to this agreement, the fabrication drawings were revised in January 2002. After the earlier revision, the shape of particle retainer to be made by PSI was modified from its drawing beforehand delivered. In this report, the fabrication drawings revised again will be shown, and the fabrication procedure (welding Qualification Tests) will be modified in accordance with the result of discussion on the 3rd technical meeting held in September 2002. These design works have been performed in Fuel Design and Evaluation Group, Plutonium Fuel Fabrication Division, Plutonium Fuel Center under the commission of Plutonium Fuel

  17. Investigation into rationalization of low decontamination pellet fuel fabrication plant configuration

    International Nuclear Information System (INIS)

    Maekawa, Kazuhiko; Yoshimura, Tadahiro; Hoshino, Yasushi; Munekata, Hideki; Tamaki, Yoshihisa

    2005-02-01

    In feasibility studies on commercialized FBR cycle system, a comprehensive system investigation and properties evaluation for candidate FBR cycle systems has been implemented through view point of safety, economics, environmental burden reduction, non-proliferation resistivity, etc. As part of these studies, an investigation into rationalization of low decontamination pellet fuel fabrication plant configuration was carried out. Until last fiscal year, conceptual design studies of the fuel fabrication plant in 200t-HM/y scale were conducted, and system properties data concerning economics and environmental burden reduction of fuel fabrication plant was acquired. In addition to this, 50t-HM/y scale plant was also schematically studied. In this fiscal year, a rationalization study on conceptual design of 50t-HM/y scale plant was conducted with main aim of economic improvement, and the 200t-HM/y scale plant design was revised based on the recent R and D progress. The system properties data concerning economics and environmental burden reduction of fuel fabrication plant was also acquired. In both case of the 50t-HM/y and 200t-HM/y scale plant, it was suggested that the equipment costs were reduced in several percentages because of reduction of maintenance equipments and cut in line number at the pellet fabrication process although granulation process fro denitration converted powder and O/M control process for pellets were added. System properties data for comparative evaluation of candidate fuel fabrication systems was also prepared. (author)

  18. Dry process fuel performance technology development

    International Nuclear Information System (INIS)

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

    2006-06-01

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

  19. Dry process fuel performance technology development

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-06-15

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

  20. Development of end plug welding method in the fabrication of FBR fuel pins

    International Nuclear Information System (INIS)

    Ohtani, Seiji; Sawayama, Takeo; Tateishi, Yoshinori

    1977-01-01

    As a part of the development of the automatic and remote controlled fabrication of FBR fuel pins, welding of fuel pin end plugs has been examined. Cladding tubes and end plugs used for this experiment are made of SUS 316, and they are the components of fuel pins for the prototype fast breeder reactor (Monju) or the second core of Joyo (Joyo MK-II). The welding tests of cladding tubes and four kinds of end plugs were carried out by means of two techniques; tungsten inert gas welding and laser welding. It can be said that no considerable difference was observed in weld penetration, occurrence rate of weld defects and breaking strength between the tight fit and the loose fit plugs. The face-to-face fit welding requires the least welding heat input, but involves much difficulty in the control of weld penetration and bead zone diameter. The good concentrative property and high energy density of laser beam make the face of weld hollow due to the vaporization of weld metal. However, this problem can be easily solved by changing the shape of end plugs. Good results in the other characteristics of the weld also were obtained by this laser welding. Further experiment is needed in connection with the compatibility of weld metal with sodium and neutron irradiation before final judgement is made on the laser welding technique. (Nakai, Y.)

  1. Fabrication of uranium carbide/beryllium carbide/graphite experimental-fuel-element specimens

    International Nuclear Information System (INIS)

    Muenzer, W.A.

    1978-01-01

    A method has been developed for fabricating uranium carbide/beryllium carbide/graphite fuel-element specimens for reactor-core-meltdown studies. The method involves milling and blending the raw materials and densifying the resulting blend by conventional graphite-die hot-pressing techniques. It can be used to fabricate specimens with good physical integrity and material dispersion, with densities of greater than 90% of the theoretical density, and with a uranium carbide particle size of less than 10 μm

  2. Fabrication of fuel elements on the basis of increased concentration fuel composition

    International Nuclear Information System (INIS)

    Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

    2004-01-01

    As a part of Russian Program RERTR Reduced Enrichment for Research and Test Reactors), at NCCP, Inc. jointly with the State Scientific Centre VNIINM the mastering in industrial environment of design and fabrication process of fuel elements (FE) with increased concentration fuel compositions is performed. Fuel elements with fuel composition on the basis of dioxide uranium with nearly 4 g/cm 3 fuel concentration have been produced thus confirming the principal possibility of fuel enrichment reduction down to 20% for research reactors which were built up according to the projects of the former USSR, by increasing the oxide fuel concentration in fuel assemblies (FAs). The form and geometrical dimensions of FEs and FAs shall remain unchanged, only uranium mass in FA shall be increased. (author)

  3. Sol-gel process for thermal reactor fuel fabrication

    International Nuclear Information System (INIS)

    Mukerjee, S.K.

    2008-01-01

    Full text: Sol-gel processes have revolutionized conventional ceramic technology by providing extremely fine and uniform powders for the fabrication of ceramics. The use of this technology for nuclear fuel fabrication has also been explored in many countries. Unlike the conventional sol-gel process, sol-gel process for nuclear fuels tries to eliminate the preparation of powders in view of the toxic nature of the powders particularly those of plutonium and 233 U. The elimination of powder handling thus makes this process more readily amenable for use in glove boxes or for remote handling. In this process, the first step is the preparation of microspheres of the fuel material from a solution which is then followed by vibro-compaction of these microspheres of different sizes to obtain the required smear density of fuel inside a pin. The maximum achievable packing density of 92 % makes it suitable for fast reactors only. With a view to extend the applicability of sol-gel process for thermal reactor fuel fabrication the concept of converting the gel microspheres derived from sol-gel process, to the pellets, has been under investigation for several years. The unique feature of this process is that it combines the advantages of sol-gel process for the preparation of fuel oxide gel microspheres of reproducible quality with proven irradiation behavior of the pellet fuel. One of the important pre-requisite for the success of this process is the preparation of soft oxide gel microspheres suitable for conversion to dense pellets free from berry structure. Studies on the internal gelation process, one of the many variants of sol-gel process, for obtaining soft oxide gel microspheres suitable for gel pelletisation is now under investigation at BARC. Some of the recent findings related to Sol-Gel Microsphere Pelletisation (SGMP) in urania-plutonia and thoria-urania systems will be presented

  4. Estimation of costs for fabrication of pressurized-water reactor fuel

    International Nuclear Information System (INIS)

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

    1979-01-01

    To provide a reference case on which to base cost estimates of the several fuel cycles to be considered, the facility, equipment, and operating requirements for the fabrication of fuel for current-design pressurized-water reactors were examined. From an analysis of these requirements, the capital and operating costs of a plant with a capacity of two metric tons of heavy metal per day (MTHM/day) were estimated. In a cash flow analysis, the lifetime of the plant was assumed to be 20 y, and the income from the sale of nuclear fuel assemblies over this period was equated to the total capital and operating expenses of the plant, including a specified 15% return on investment. In this way a levelized unit price for the fuel was obtained. The effects of inflation were not considered since the purpose of these estimates and the determination of unit price was to permit comparison of different types of fuels. The capital costs of the fuel fabrication plant were estimated at $32 million for the facility--land, site preparation, building--and $34 million for equipment. Annual operating costs including labor, management, materials, and utilities were estimated to be $36.5 million. From these estimates, the unit price for fabricating the fuel for the reference pressurized-water reactor was determined to be $138/kg of heavy metal or $63,600 per fuel assembly

  5. Project fuel development

    International Nuclear Information System (INIS)

    Stratton, R.W.

    1981-05-01

    The activities continued on lab-scale production of uranium-plutonium carbide fuel for the fast reactor using gelation methods, irradiation testing and performance evaluation. Whereas in earlier years a balance was maintained between research and development or with emphasis on research, 1980 was marked by a concentrated equipment development effort for an increased throughput with improved process control and product reproducability and installation of new equipment for large pin fabrication. (Auth.)

  6. FY2015 ceramic fuels development annual highlights

    Energy Technology Data Exchange (ETDEWEB)

    Mcclellan, Kenneth James [Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)

    2015-09-22

    Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2015 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY15 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

  7. FY2016 Ceramic Fuels Development Annual Highlights

    Energy Technology Data Exchange (ETDEWEB)

    Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-24

    Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2016 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY16 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

  8. Advanced methods of quality control in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Onoufriev, Vladimir

    2004-01-01

    Under pressure of current economic and electricity market situation utilities implement more demanding fuel utilization schemes including higher burn ups and thermal rates, longer fuel cycles and usage of Mo fuel. Therefore, fuel vendors have recently initiated new R and D programmes aimed at improving fuel quality, design and materials to produce robust and reliable fuel. In the beginning of commercial fuel fabrication, emphasis was given to advancements in Quality Control/Quality Assurance related mainly to product itself. During recent years, emphasis was transferred to improvements in process control and to implementation of overall Total Quality Management (TQM) programmes. In the area of fuel quality control, statistical control methods are now widely implemented replacing 100% inspection. This evolution, some practical examples and IAEA activities are described in the paper. The paper presents major findings of the latest IAEA Technical Meetings (TMs) and training courses in the area with emphasis on information received at the TM and training course held in 1999 and other latest publications to provide an overview of new developments in process/quality control, their implementation and results obtained including new approaches to QC

  9. Design Evolutuion of Hot Isotatic Press Cans for NTP Cermet Fuel Fabrication

    Science.gov (United States)

    Mireles, O. R.; Broadway, J.; Hickman, R.

    2014-01-01

    Nuclear Thermal Propulsion (NTP) is under consideration for potential use in deep space exploration missions due to desirable performance properties such as a high specific impulse (> 850 seconds). Tungsten (W)-60vol%UO2 cermet fuel elements are under development, with efforts emphasizing fabrication, performance testing and process optimization to meet NTP service life requirements [1]. Fuel elements incorporate design features that provide redundant protection from crack initiation, crack propagation potentially resulting in hot hydrogen (H2) reduction of UO2 kernels. Fuel erosion and fission product retention barriers include W coated UO2 fuel kernels, W clad internal flow channels and fuel element external W clad resulting in a fully encapsulated fuel element design as shown.

  10. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    International Nuclear Information System (INIS)

    Knight, R.W.; Morin, R.A.

    1999-01-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U 3 O 8 powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated

  11. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    Energy Technology Data Exchange (ETDEWEB)

    Knight, R.W.; Morin, R.A.

    1999-12-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

  12. US Progress on Property Characterization to Support LEU U-10 Mo Monolithic Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Cole, James Irvin [Idaho National Laboratory; Rabin, Barry H [Idaho National Laboratory; Smith, James Arthur [Idaho National Laboratory; Scott, Clark Landon [Idaho National Laboratory; Benefiel, Bradley Curtis [Idaho National Laboratory; Larsen, Eric David [Idaho National Laboratory; Lind, Robert Paul [Idaho National Laboratory; Sell, David Alan [Idaho National Laboratory

    2016-03-01

    The US High Performance Research Reactor program is pursuing development and qualification of a new high density monolithic LEU fuel to facilitate conversion of five higher power research reactors located in the US (ATR, HFIR, NBSR, MIT and MURR). In order to support fabrication development and fuel performance evaluations, new testing capabilities are being developed to evaluate the properties of fuel specimens. Residual stress and fuel-cladding bond strength are two characteristics related to fuel performance that are being investigated. In this overview, new measurement capabilities being developed to assess these characteristics in both fresh and irradiated fuel are described. Progress on fresh fuel testing is summarized and on-going hot-cell implementation efforts to support future PIE campaigns are detailed. It is anticipated that benchmarking of as-fabricated fuel characteristics will be critical to establishing technical bases for specifications that optimize fuel fabrication and ensure acceptable in-reactor fuel performance.

  13. Projected developments in the US and European LWR fabrication market

    International Nuclear Information System (INIS)

    Anderson, C.K.; Varley, G.

    1996-01-01

    There are several important factors influencing change in the fuel fabrication supply industry in Western Europe today. The most important is the changing supply relationships resulting from lower costs in the USA. With US fabrication prices substantially lower than in Western Europe, a situation which is sustainable based on production costs, a major force for change prevails which will have significant implications for the commercial development of both markets over the next decade. The extent to which supply from the USA will drive the Western Europe market will depend, to some extent, on three other factors: the new US-Euratom agreement on nuclear trade; electricity market deregulation in the European Union (EU); near term MOX fuel fabrication capacity. The main purpose of this paper is to discuss the current situation in the USA and Europe, and the manner in which this will lead to changing fuel fabrication supply dynamics in both markets. (author)

  14. Quarterly Progress Report Fuels Development Operation: October - December 1959

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation; Tobin, J. C. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

    1960-01-15

    The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

  15. Quarterly Progress Report Fuels Development Operation: January - March 1958

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation; Tobin, J. C. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

    1958-04-15

    The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

  16. Quarterly Progress Report Fuels Development Operation: July - September 1957

    Energy Technology Data Exchange (ETDEWEB)

    Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Wallace, W. P. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

    1957-10-15

    The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

  17. Plant overview of JNFL MOX fuel fabrication plant (J-MOX)

    International Nuclear Information System (INIS)

    Hiruta, Kazuhiko; Suzuki, Masataka; Shimizu, Junji; Suzuki, Kazumi; Yamamoto, Yutaka; Deguchi, Morimoto; Fujimaki, Kazunori

    2005-01-01

    In April 2005, JNFL submitted METI an application for the permission of MOX fuel fabrication business for JNFL MOX Fuel Fabrication Plant (J-MOX). Accordingly, safeguards formalities and discussion with the Agency have been also started for J-MOX as an official project. This report describes J-MOX plant overview and also presents outline of J-MOX by focusing on safeguards features and planned material accountancy method. (author)

  18. Advanced fuel for fast breeder reactors: Fabrication and properties and their optimization

    International Nuclear Information System (INIS)

    1988-06-01

    The present design for FBR fuel rods includes usually MOX fuel pellets cladded into stainless steel tubes, together with UO 2 axial blanket and stainless steel hexagonal wrappers. Mixed carbide, nitride and metallic fuels have been tested as alternative fuels in test reactors. Among others, the objectives to develop these alternative fuels are to gain a high breeding ratio, short doubling time and high linear ratings. Fuel rod and assembly designers are now concentrating on finding the combination of optimized fuel, cladding and wrapper materials which could result in improvement of fuel operational reliability under high burnups and load-follow mode of operation. The purpose of the meeting was to review the experience of advanced FBR fuel fabrication technology, its properties before, under and after irradiation, peculiarities of the back-end of the nuclear fuel cycle, and to outline future trends. As a result of the panel discussion, the recommendations on future Agency activities in the area of advanced FBR fuels were developed. A separate abstract was prepared for each of the 10 presentations of this meeting. Refs, figs and tabs

  19. Fabrication of nano-structured UO2 fuel pellets

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Kang, Ki Won; Rhee, Young Woo; Kim, Dong Joo; Kim, Jong Heon; Kim, Keon Sik; Song, Kun Woo

    2007-01-01

    Nano-structured materials have received much attention for their possibility for various functional materials. Ceramics with a nano-structured grain have some special properties such as super plasticity and a low sintering temperature. To reduce the fuel cycle costs and the total mass of spent LWR fuels, it is necessary to extend the fuel discharged burn-up. In order to increase the fuel burn-up, it is important to understand the fuel property of a highly irradiated fuel pellet. Especially, research has focused on the formation of a porous and small grained microstructure in the rim area of the fuel, called High Burn-up Structure (HBS). The average grain size of HBS is about 300nm. This paper deals with the feasibility study on the fabrication of nano-structured UO 2 pellets. The nano sized UO 2 particles are prepared by a combined process of a oxidation-reducing and a mechanical milling of UO 2 powder. Nano-structured UO 2 pellets (∼300nm) with a density of ∼93%TD can be obtained by sintering nano-sized UO 2 compacts. The SEM study reveals that the microstructure of the fabricated nano-structure UO 2 pellet is similar to that of HBS. Therefore, this bulk nano-structured UO 2 pellet can be used as a reference pellet for a measurement of the physical properties of HBS

  20. Fuel fabrication and reprocessing at UKAEA Dounreay

    International Nuclear Information System (INIS)

    Anderson, B.

    1994-01-01

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

  1. Year One Summary of X-energy Pebble Fuel Development at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McMurray, Jake W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunt, Rodney D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jolly, Brian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Trammell, Michael P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Daniel R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Blamer, Brandon J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Reif, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kim, Howard T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    The Advanced Reactor Concepts X-energy (ARC-Xe) Pebble Fuel Development project at Oak Ridge National Laboratory (ORNL) has successfully completed its first year, having made excellent progress in accomplishing programmatic objectives. The primary focus of research at ORNL in support of X-energy has been the training of X-energy fuel fabrication engineers and the establishment of US pebble fuel production capabilities able to supply the Xe-100 pebble-bed reactor. These efforts have been strongly supported by particle fuel fabrication and characterization expertise present at ORNL from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program.

  2. Development of Metallic Fuels for Actinide Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, Steven Lowe [Idaho National Laboratory; Fielding, Randall Sidney [Idaho National Laboratory; Benson, Michael Timothy [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory; Carmack, William Jonathan [Idaho National Laboratory

    2015-09-01

    Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  4. Equipment system for advanced nuclear fuel development

    International Nuclear Information System (INIS)

    Kwon, Hyuk Il; Ji, C. G.; Bae, S. O.

    2002-11-01

    The purpose of the settlement of equipment system for nuclear Fuel Technology Development Facility(FTDF) is to build a seismic designed facility that can accommodate handling of nuclear materials including <20% enriched Uranium and produce HANARO fuel commercially, and also to establish the advanced common research equipment essential for the research on advanced fuel development. For this purpose, this research works were performed for the settlement of radiation protection system and facility special equipment for the FTDF, and the advanced common research equipment for the fuel fabrication and research. As a result, 11 kinds of radiation protection systems such as criticality detection and alarm system, 5 kinds of facility special equipment such as environmental pollution protection system and 5 kinds of common research equipment such as electron-beam welding machine were established. By the settlement of exclusive domestic facility for the research of advanced fuel, the fabrication and supply of HANARO fuel is possible and also can export KAERI-invented centrifugal dispersion fuel materials and its technology to the nations having research reactors in operation. For the future, the utilization of the facility will be expanded to universities, industries and other research institutes

  5. Description of fuel element brush assembly's fabrication for 105-K west

    International Nuclear Information System (INIS)

    Maassen, D.P.

    1997-01-01

    This report is a description of the process to redesign and fabricate, as well as, describe the features of the Fuel Element Brush Assembly used in the 105-K West Basin. This narrative description will identify problems that occurred during the redesigning and fabrication of the 105-K West Basin Fuel Element Brush Assembly and specifically address their solutions

  6. Sphere-pac fuel development program. First semi-annual progress report, October 1979-March 1980

    International Nuclear Information System (INIS)

    Felt, R.E.

    1980-05-01

    Development of processes by Exxon Nuclear Company, Inc. for fabricating spherical particle nuclear fuel (sphere-pac) during October 1979 through March 1980 is reported. The program surveyed available technology to develop an initial flowsheet as a design basis for process development. An 0.1 ton/day pilot plant was built to develop and demonstrate the fabrication of sphere-pac fuel. Process and equipment efforts have been directed towards the demonstration of processes and equipment necessary to fabricate sphere-pac fuel on a commercial scale

  7. Study on the nitride fuel fabrication for FBR cycle (1)

    International Nuclear Information System (INIS)

    Shinkai, Yasuo; Ono, Kiyoshi; Tanaka, Kenya

    2002-07-01

    In the phase-II of JNC's 'Feasibility Study on Commercialized Fuel Reactor Cycle System (the F/S)', the nitride fuels are selected as candidate for fuels for heavy metal cooled reactor, gas cooled reactor, and small scale reactor. In particular, the coated fuel particles are a promising concept for gas cooled reactor. In addition, it is necessary to study in detail the application possibility of pellet nitride fuel and vibration compaction nitride fuel for heavy metal cooled reactor and small scale reactor in the phase-II. In 2001, we studied more about additional equipments for the nitride fuel fabrication in processes from gelation to carbothermic reduction in the vibration compaction method. The result of reevaluation of off-gas mass flow around carbothermic reduction equipment in the palletizing method, showed that quantity of off-gas flow reduced and its reduction led the operation cost to decrease. We studied the possibility of fabrication of large size particles in the coated fuel particles for helium gas cooled reactor and we made basic technical issues clear. (author)

  8. Advanced methods for fabrication of PHWR and LMFBR fuels

    International Nuclear Information System (INIS)

    Ganguly, C.

    1988-01-01

    For self-reliance in nuclear power, the Department of Atomic Energy (DAE), India is pursuing two specific reactor systems, namely the pressurised heavy water reactors (PHWR) and the liquid metal cooled fast breeder reactors (LMFBR). The reference fuel for PHWR is zircaloy-4 clad high density (≤ 96 per cent T.D.) natural UO 2 pellet-pins. The advanced PHWR fuels are UO 2 -PuO 2 (≤ 2 per cent), ThO 2 -PuO 2 (≤ 4 per cent) and ThO 2 -U 233 O 2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O 2 pellets clad in SS 316 or D9 is the reference fuel for the first generation of prototype and commercial LMFBRs all over the world. However, (UPu)C and (UPu)N are considered as advanced fuels for LMFBRs mainly because of their shorter doubling time. The conventional method of fabrication of both high and low density oxide, carbide and nitride fuel pellets starting from UO 2 , PuO 2 and ThO 2 powders is 'powder metallurgy (P/M)'. The P/M route has, however, the disadvantage of generation and handling of fine powder particles of the fuel and the associated problem of 'radiotoxic dust hazard'. The present paper summarises the state-of-the-art of advanced methods of fabrication of oxide, carbide and nitride fuels and highlights the author's experience on sol-gel-microsphere-pelletisation (SGMP) route for preparation of these materials. The SGMP process uses sol gel derived, dust-free and free-flowing microspheres of oxides, carbide or nitride for direct pelletisation and sintering. Fuel pellets of both low and high density, excellent microhomogeneity and controlled 'open' or 'closed' porosity could be fabricated via the SGMP route. (author). 5 tables, 14 figs., 15 refs

  9. Structure, conduct, and sustainability of the international low-enriched fuel fabrication industry

    International Nuclear Information System (INIS)

    Rothwell, Geoffrey

    2008-01-01

    This paper examines the cost structures of fabricating Low-Enriched Uranium fuel (LEU, enriched to 5% enrichment) light water reactor fuels. The LEU industry is decades old, and (except for high entry cost, i.e., the cost of designing and licensing a fuel fabrication facility and its fuel), labor and additional fabrication lines can be added by industry incumbents at Nth-of-a-Kind cost to the maximum capacity allowed by the license. On the other hand, new entrants face higher First-of-a-Kind costs and high new-facility licensing costs, increasing the scale required for entry thus discouraging small scale entry by countries with only a few nuclear power plants. Therefore, the industry appears to be competitive with sustainable investment in fuel-cycle states, and structural barriers-to-entry increase its proliferation resistance. (author)

  10. Development of 3-Pin Fuel Test Loop and Utilization Technology

    International Nuclear Information System (INIS)

    Lee, Chung Young; Sim, B. S.; Lee, C. Y.

    2007-06-01

    The principal contents of this project are to design, fabricate and install the steady-state fuel test loop in HANARO for nuclear technology development. Procurement and, fabrication of main equipment, licensing and installation for fuel test loop have been performed. Following contents are described in the report. 1. Design - Design of the In-pile system and Out pile system 2. Fabrication and procurement of the equipment - Fabrication of the In-pile system and In-pool piping - Fabrication and procurement of the equipment of the out-pile system 3. Acquisition of the license - Preparation of the safety analysis report and acquisition of the license - Pre-service inspection of the facility 4. Installation and commissioning - Installation of the FTL - Development of the commissioning procedure

  11. The fabrication of nuclear fuel elements in Mexico

    International Nuclear Information System (INIS)

    Guerrero Morillo, H.L.

    1977-01-01

    The situation of nuclear electricity generation in Mexico in 1976 is described: two nuclear reactors were under construction but no definite programme on the type and start-up dates for the next power plants existed. However, the existence of a general plan on nuclear power plants is mentioned, which, according to the latest estimates, will provide 10,000MW installed by 1990. The national intention, as laid down in an appropriate Law, is to supply domestic nuclear fuel to the power reactors operating in the country, starting with the first reloading of the two BWRs at the first national station in Laguna Verde, required at the end of 1981 and 1982, respectively. Before this can be achieved and to provide the relatively small amounts of fuel elements for the two reactors, Mexico must adopt a strategy of fuel elements fabrication. The two main options are analysed: (1) to delay local fabrication until a national nuclear programme has been defined, meanwhile purchasing abroad the necessary initial cores and refuelling; (2) to start local fabrication of fuel elements as soon as possible in order to provide the first refuelling of the first unit of Laguna Verde, confronting the economic risks of such a decision with the advantages of immediate action. Both options are analysed in detail, comparing them especially from the economic point of view. Current information from potential licensors for design and manufacture are used in the analysis. (author)

  12. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    International Nuclear Information System (INIS)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

  13. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  14. Effects of fabrication requirements on fuel performance in relation to operating conditions. The views of Electricite de France (EDF)

    International Nuclear Information System (INIS)

    Ponticq, M.; Richer, P.; Scribe, G.

    1979-01-01

    Because of the operational constraints relating to fuel behaviour, imperfect knowledge of the behaviour of a defective fuel assembly and, in the near future, the need to adapt reactor power to grid following (load following and remote control), EDF is aiming to reduce the present rate of fuel failure. While the phenomena affecting fuel behaviour have now been listed and analysed, the efforts at reducing their consequences have yet to be completed. This can be achieved, firstly, by reducing or eliminating fabrication defects, which are responsible for failure at the beginning of fuel life, through establishment of a good quality assurance organization and the search for still higher efficiency of quality control and fabrication equipment, and secondly, by developing fabrication techniques minimizing in particular cladding-pellet interactions and the stress corrosion of the cladding, which are responsible for fuel failure as from mid-life. However, the reactor operating conditions likely to apply in the near future may lead, for a given fuel configuration, to a re-evaluation of the fabrication parameters of cladding and UO 2 pellets. (author)

  15. Modern methods of material accounting for mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Eggers, R.F.; Pindak, J.L.; Brouns, R.J.; Williams, R.C.; Brite, D.W.; Kinnison, R.R.; Fager, J.E.

    1981-01-01

    The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MCandA) philosophy have been applied to a mixed oxide fuel fabrication plant to produce a detailed preliminary MCandA system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed oxide fuel fabrication plant study

  16. Modern methods of material accounting for mixed-oxide fuel-fabrication facility

    International Nuclear Information System (INIS)

    Eggers, R.F.; Brouns, R.J.; Brite, D.W.; Pindak, J.L.

    1981-07-01

    The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MC and A) philosophy have been applied to a mixed-oxide fuel-fabrication plant to produce a detailed preliminary MC and A system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed-oxide fuel-fabrication plant study. Topics covered in this paper include: mixed-oxide fuel-fabrication process description, process disaggregation into MC and A system control units, quantitative results of analysis of control units for abrupt and recurring loss-detection capability, impact of short- and long-term holdup on loss-detection capability, response to alarms for abrupt loss, and response to alarms for recurring loss

  17. SEM and TEM Characterization of As-Fabricated U-7Mo Disperson Fuel Plates

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Yao, B.; Perez, E.; Sohn, Y.H.

    2009-01-01

    The starting microstructure of a dispersion fuel plate can have a dramatic impact on the overall performance of the plate during irradiation. To improve the understanding of the as-fabricated microstructures of dispersion fuel plates, SEM and TEM analysis have been performed on RERTR-9A archive fuel plates, which went through an additional hot isostatic procsssing (HIP) step during fabrication. The fuel plates had depleted U-7Mo fuel particles dispersed in either Al-2Si or 4043 Al alloy matrix. For the characterized samples, it was observed that a large fraction of the ?-phase U-7Mo alloy particles had decomposed during fabrication, and in areas near the fuel/matrix interface where the transformation products were present significant fuel/matrix interaction had occurred. Relatively thin Si-rich interaction layers were also observed around the U-7Mo particles. In the thick interaction layers, (U)(Al,Si)3 and U6Mo4Al43 were identified, and in the thin interaction layers U(Al,Si)3, U3Si3Al2, U3Si5, and USi1.88-type phases were observed. The U3Si3Al2 phase contained some Mo. Based on the results of this work, exposure of dispersion fuel plates to relatively high temperatures during fabrication impacts the overall microstructure, particularly the nature of the interaction layers around the fuel particles. The time and temperature of fabrication should be carefully controlled in order to produce the most uniform Si-rich layers around the U-7Mo particles.

  18. Status of the atomized uranium silicide fuel development at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C.K.; Kim, K.H.; Park, H.D.; Kuk, I.H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-08-01

    While developing KMRR fuel fabrication technology an atomizing technique has been applied in order to eliminate the difficulties relating to the tough property of U{sub 3}Si and to take advantage of the rapid solidification effect of atomization. The comparison between the conventionally comminuted powder dispersion fuel and the atomized powder dispersion fuel has been made. As the result, the processes, uranium silicide powdering and heat treatment for U{sub 3}Si transformation, become simplified. The workability, the thermal conductivity and the thermal compatibility of fuel meat have been investigated and found to be improved due to the spherical shape of atomized powder. In this presentation the overall developments of atomized U{sub 3}Si dispersion fuel and the planned activities for applying the atomizing technique to the real fuel fabrication are described.

  19. Modernization of RTC for fabrication of MOX fuel, Vibropac fuel pins and BN-600 FA with weapon grade plutonium

    International Nuclear Information System (INIS)

    Grachyov, A.F.; Kalygin, V.V.; Skiba, O.V.; Mayorshin, A. A.; Bychkov, A.V.; Kisly, V.A.; Ovsyannikov, Y.F.; Bobrov, D.A.; Mamontov, S.I.; Tsyganov, A.N.; Churutkin, E.I.; Davydov, P.I.; Samosenko, E.A; Shalak, A.R.; Ojima, Hisao

    2004-01-01

    Since mid 70's RIAR has been performing activities on plutonium involvement in fuel cycle. These activities are considered a stage within the framework of the closed fuel cycle development. Developed at RIAR fuel cycle is based on two technologies: 'dry' process of fuel reprocessing and vibro-packing method for fuel pin fabrication. Due to the available scientific capabilities and a gained experience in operating the technological facilities (ORYOL, SIC) for plutonium (various grade) blending into fuel for fast reactors, RIAR is a participant of the activities aimed at solving these tasks. Under international program RIAR with financial support of JNC (Japan) is modernizing the facility for granulated fuel production, vibro-pac fuel pins and FA fabrication to provide the BN-600 'hybrid' core. In order to provide 'hybrid' core it is necessary to produce (per year): - 1775 kg of granulated MOX-fuel, 6500 fuel pins, 50 fuel assemblies. Potential output of the facility under construction is as follows: - 1800 kg of granulated MOX-fuel per year, 40 fuel pins per shift, 200 FAs for the BN-600 reactor per year. Taking into account domestic and foreign experience in MOX-fuel production, different options were discussed of the equipment layouts in the available premises of chemical technological division of RIAR: - in the shielded manipulator boxes, in the existing hot cells. During construction of the facility in the building under operation the following requirements should be met: - facility must meet all standards and regulations set for nuclear facilities, installation work at the facility must not influence other production programs implemented in the building, engineering supply lines of the facility must be connected to the existing service lines of the building, cost of the activities must not exceed amount of JNC funding. The paper presents results of comparison between two options of the process equipment layout: in boxes and hot cells. This equipment is intended

  20. Selection and development of advanced nuclear fuel products

    International Nuclear Information System (INIS)

    Stucker, David L.; Miller, Richard S.; Arnsberger, Peter L.

    2004-01-01

    The highly competitive international marketplace requires a continuing product development commitment, short development cycle times and timely, on-target product development to assure customer satisfaction and continuing business. Westinghouse has maintained its leadership position within the nuclear fuel industry with continuous developments and improvements to fuel assembly materials and design. This paper presents a discussion of the processes used by Westinghouse in the selection and refinement of advanced concepts for deployment in the highly competitive US and international nuclear fuel fabrication marketplace. (author)

  1. Design and fuel fabrication processes for the AC-3 mixed-carbide irradiation test

    International Nuclear Information System (INIS)

    Latimer, T.W.; Chidester, K.M.; Stratton, R.W.; Ledergerber, G.; Ingold, F.

    1992-01-01

    The AC-3 test was a cooperative U.S./Swiss irradiation test of 91 wire-wrapped helium-bonded U-20% Pu carbide fuel pins irradiated to 8.3 at % peak burnup in the Fast Flux Test Facility. The test consisted of 25 pins that contained spherepac fuel fabricated by the Paul Scherrer Institute (PSI) and 66 pins that contained pelletized fuel fabricated by the Los Alamos National Laboratory. Design of AC-3 by LANL and PSI was begun in 1981, the fuel pins were fabricated from 1983 to 1985, and the test was irradiated from 1986 to 1988. The principal objective of the AC-3 test was to compare the irradiation performance of mixed-carbide fuel pins that contained either pelletized or sphere-pac fuel at prototypic fluence and burnup levels for a fast breeder reactor

  2. The development and localization of nuclear fuel technology for KMRR

    International Nuclear Information System (INIS)

    Kim, Seong Yun; Lee, Ji Bok; Suk, Ho Chun; Kuk, Il Hyun; Hwang, Woan; Kim, Bong Goo; Park, Joo Hwan; Kim, Young Jin; Kang, Thae Khapp; Lee, Jae Choon

    1988-05-01

    This project was implemented aiming at localizing the fabrication of the KMRR fuel by october 1993. The contents of this project were divided into three parts: fuel design, fuel fabrication and process criticality analysis. In the fuel design, the radial power distribution in the fuel core was modeled and formulated taking account of the neutron flux depression in the radial direction. It was also performed to model and formulate the thermal characteristics such as the thermal conductivity and specific heat of the fuel core, U3Si-Al, the swelling and the film coefficient of heat transfer between the aluminum clad and light water coolant. The two dimensional heat transfer in the finned fuel element was equated based on the general equation governing the heat transfer in materials in order to develope a computer code, TEMP2D. TEMP2D solves finite differenced equations to calculate a two dimensional fuel temperature distribution under the steady and transient states. In the fuel fabrication, the technologies of fabricating uranium silicide fuel meat were tried by using depleted uranium as a raw material. These were extended to find the problems in technologies and to establish the ways of approach. The end product, so called fuel meat, was a metallic powder compound, U3Six(1≤x≤2), dispersed in Al matrix. The fuel meat was fabricated by the horizontal extrusion technique, and powder extrusion technique. Fabrication technologies comprise five different continuous processes: melting and casting of metallic uranium with silicon and aluminum, heat treatment, chipping and crushing, pulverizing, and extrusion. In the process criticality analysis, AMPX-KENO benchmark calculation was performed and calculational error of AMPX-KENO system was established. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-08-01

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

  4. Nuclear fuels technologies fiscal year 1998 research and development test plan

    International Nuclear Information System (INIS)

    Alberstein, D.; Blair, H.T.; Buksa, J.J.

    1998-06-01

    A number of research and development (R and D) activities are planned at Los Alamos National Laboratory (LANL) in FY98 in support of the Department of Energy Office of Fissile Materials Disposition (DOE-MD). During the past few years, the ability to fabricate mixed oxide (MOX) nuclear fuel using surplus-weapons plutonium has been researched, and various experiments have been performed. This research effort will be continued in FY98 to support further development of the technology required for MOX fuel fabrication for reactor-based plutonium disposition. R and D activities for FY98 have been divided into four major areas: (1) feed qualification/supply, (2) fuel fabrication development, (3) analytical methods development, and (4) gallium removal. Feed qualification and supply activities encompass those associated with the production of both PuO 2 and UO 2 feed materials. Fuel fabrication development efforts include studies with a new UO 2 feed material, alternate sources of PuO 2 , and determining the effects of gallium on the sintering process. The intent of analytical methods development is to upgrade and improve several analytical measurement techniques in support of other R and D and test fuel fabrication tasks. Finally, the purpose of the gallium removal system activity is to develop and integrate a gallium removal system into the Pit Disassembly and Conversion Facility (PDCF) design and the Phase 2 Advanced Recovery and Integrated Extraction System (ARIES) demonstration line. These four activities will be coordinated and integrated appropriately so that they benefit the Fissile Materials Disposition Program. This plan describes the activities that will occur in FY98 and presents the schedule and milestones for these activities

  5. Establishment of technological basis for fabrication of U-Pu-Zr ternary alloy fuel pins for irradiation tests in Japan

    International Nuclear Information System (INIS)

    Kikuchi, Hironobu; Iwai, Takashi; Nakajima, Kunihisa; Arai, Yasuo; Nakamura, Kinya; Ogata, Takanari

    2011-01-01

    A high-purity Ar gas atmosphere glove box accommodating injection casting and sodium-bonding apparatuses was newly installed in the Plutonium Fuel Research Facility of Oarai Research and Development Center, Japan Atomic Energy Agency, in which several nitride and carbide fuel pins were fabricated for irradiation tests. The experiences led to the establishment of the technological basis of the fabrication of U-Pu-Zr alloy fuel pins for the first time in Japan. After the injection casting of the U-Pu-Zr alloy, the metallic fuel pins were fabricated by welding upper and lower end plugs with cladding tubes of ferritic-martensitic steel. Subsequent to the sodium bonding for filling the annular gap region between the U-Pu-Zr alloy and the cladding tube with the melted sodium, the fuel pins for irradiation tests are inspected. This paper shows the apparatuses and the technological basis for the fabrication of U-Pu-Zr alloy fuel pins for the irradiation test planned at the experimental fast test reactor Joyo. (author)

  6. Fabrication development of full-sized components for GCFR core assemblies

    International Nuclear Information System (INIS)

    Lindgren, J.R.; Flynn, P.W.; Foster, L.C.

    1980-05-01

    This paper presents the status of the development of full-sized components for gas-cooled fast reactor (GCFR) core assemblies. Methods for ribbing of the fuel rod cladding, fabrication of grid spacers of two different designs, drawing of assembly flow ducts, and fabrication of fission gas collection manifolds by several methods are discussed

  7. MELOX fuel fabrication plant: Operational feedback and future prospects

    International Nuclear Information System (INIS)

    Hugelmann, D.; Greneche, D.

    2000-01-01

    As of December 1, 1998, 32 Europeans LWRs are loaded with MOX fuel. It clearly means that plutonium recycling in MOX fuels is a mature industry, with successful operational experience in fabrication plants in some European countries, especially in France. Indeed, the recycling of plutonium generated in LWRs is one of the objectives of the full Reprocessing-Conditioning-Recycling (RCR) strategy chosen by France in the 70's. The most impressive results of this strategy, is the fact that 31 of the 32 reactors are loaded with MOX fuels supplied by the COGEMA Group from the same efficient fabrication process, the MIMAS process, improved for the MELOX plant to become the A-MIMAS process. In France, 17 reactors are already loaded and 11 additional reactors are technically suited to do so. Indeed, the EDF MOX program plans to use MOX in 28 of its 57 reactors. An EDF 900 MWe reactor core contains 157 assemblies of 264 rods each. 52 fuel assemblies per year are necessary for a 'UO 2 3-batches-MOX 3-batches' core management. In this case, a third of the UO 2 and a third of the MOX assemblies are replaced yearly, that means 36 UO 2 fuel assemblies and 16 MOX fuel assemblies. Some MOX fuelled reactors have now switched from the previously described core management to a so-called 'hybrid core management'. In this case, a quarter of UO 2 assemblies is replaced yearly. The first EDF reactor loaded with MOX fuel was Saint-Laurent B1, in 1987. The in-core experience, based on several hundred assemblies loaded, with reloading on a 1/3 cycle basis, shows that there is no operational difference between UO 2 and MOX fuels, both in terms of performance and safety. MOX fueling of 900 MWe EDF's PWRs, with a limited in-core MOX ratio of 30%, has needed only minor adaptations, such as addition of control rods, modification of the boron concentration in the cooling system and precaution against radiation exposure, easy to set up (optimisation of the fresh MOX fuel handling process, remote

  8. Automated fuel fabrication- a vision comes true

    International Nuclear Information System (INIS)

    Hemantha Rao, G.V.S.; Prakash, M.S.; Setty, C.R.P.; Gupta, U.C.

    1997-01-01

    When New Uranium Fuel Assembly Project at Nuclear Fuel Complex (NFC) begins production, its operator will have equipment provided with intramachine handling systems working automatically by pressing a single button. Additionally simple low cost inter machine handling systems will further help in critical areas. All these inter and intra machine handling systems will result in improved reliability, productivity and quality. The fault diagnostics, mimics and real time data acquisition systems make the plant more operator friendly. The paper deals with the experience starting from layout, selection of product carriers, different handling systems, the latest technology and the integration of which made the vision on automation in fuel fabrication come true. (author)

  9. Dissolution of mixed oxide fuel as a function of fabrication variables

    International Nuclear Information System (INIS)

    Lerch, R.E.

    1979-08-01

    Dissolution properties of mechanically blended mixed oxide fuel were very dependent on the six fuel fabrication variables studied. Fuel sintering temperature, source of PuO 2 and PuO 2 content of the fuel had major effects: (1) as the sintering temperature was increased from 1400 to 1700 0 C, pellet dissolution was more complete; (2) pellets made from burned metal derived PuO 2 were more completely dissolved than pellets made from calcined nitrate derived PuO 2 which in turn were more completely dissolved than pellets made from calcined nitrate derived PuO 2 ; (3) as the PuO 2 content decreased from 25 to 15 wt % PuO 2 , pellet dissolution was more complete. Preferential dissolution of uranium occurred in all the mechanically blended mixed oxide. Unirradiated mixed oxide fuel pellets made by the Sol Gel process were generally quite soluble in nitric acid. Unirradiated mixed oxide fuel pellets made by the coprecipitation process dissolved completely and rapidly in nitric acid. Fuel made by the coprecipitation process was more completely dissolved than fuel made by the Sol Gel process which, in turn, was more completely dissolved than fuel made by mechanically blending UO 2 and PuO 2 as shown below. Addition of uncomplexed fluoride to nitric acid during fuel dissolution generally rendered all fuel samples completely dissolvable. In boiling 12M nitric acid, 95 to 99% of the plutonium which was going to dissolve did so in the first hour. Irradiated mechanically blended mixed oxide fuel with known fuel fabrication conditions was also subjected to fuel dissolution tests. While irradiation was shown to increase completeness of plutonium dissolution, poor dissolubility due to adverse fabrication conditions (e.g., low sintering temperature) remained after irradiation

  10. Fuel development at CERCA. Status of development - September 1984

    International Nuclear Information System (INIS)

    Fanjas, Y.; Dewez, Ph.; Savornin, B.

    1985-01-01

    Since 1978, CERCA has developed high density aluminide (UAl x ), oxide (U 3 O 8 ) and silicides (U 3 Si 2 , U 3 Si) fuels allowing the use of 19.75 enriched uranium in research and test reactors. An extensive irradiation program has been carried out to test the full size fuel plates and fuel elements fabricated by CERCA. So far, all the irradiation tests have given satisfactory results whatever the uranium density, the burn-up level and the type of fuel. In particular, silicides which cover the whole density range from 1 to 7 g U/cm 3 appear more and more as the standard fuels for the future. (author)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  12. U-10Mo Baseline Fuel Fabrication Process Description

    Energy Technology Data Exchange (ETDEWEB)

    Hubbard, Lance R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Arendt, Christina L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dye, Daniel F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Clayton, Christopher K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lerchen, Megan E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lombardo, Nicholas J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zacher, Alan H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-09-27

    This document provides a description of the U.S. High Power Research Reactor (USHPRR) low-enriched uranium (LEU) fuel fabrication process. This document is intended to be used in conjunction with the baseline process flow diagram (PFD) presented in Appendix A. The baseline PFD is used to document the fabrication process, communicate gaps in technology or manufacturing capabilities, convey alternatives under consideration, and as the basis for a dynamic simulation model of the fabrication process. The simulation model allows for the assessment of production rates, costs, and manufacturing requirements (manpower, fabrication space, numbers and types of equipment, etc.) throughout the lifecycle of the USHPRR program. This document, along with the accompanying PFD, is updated regularly

  13. Radiological safety aspects in the fabrication of mixed oxide fuel elements

    International Nuclear Information System (INIS)

    Krishnamurthi, T.N.; Janardhanan, S.; Soman, S.D.

    1981-01-01

    The problems of radiological safety in the fabrication of (U, Pu)O 2 fuel assemblies for fast reactors utilising high exposure plutonium are discussed. Derived working limits for plutonium as a function of the burn-up of RAPS (Rajasthan Atomic Power Station) fuel, external gamma and neutron exposures from feed product batches, finished fuel pins and assemblies are presented. Shielding requirements for the various glove box operations are also indicated. In general, high exposure plutonium handling calls for remote fabrication and automation at various stages would play a key role in minimising exposures to personnel in a large production plant. (author)

  14. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    Science.gov (United States)

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  15. Interfuel: development of fuel fabrication techniques

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    On July 5 1972, an understanding was reached between Rijn-Schelde-Verolme NV(RSV), Shell Kernenergie NV, Gemeenschappelijke Kernenergiecentrale Nederland NV (GKN), Comprimo BV and Stichting Reactor Centrum Nederland (RCN), regarding the formation of a company to co-operate concerning the fuel cycle for nuclear reactors-special emphasis being given to the production of fuel elements for light water reactor systems. (Auth.)

  16. Design, fabrication and operating experience of Monju ex-vessel fuel storage tank

    International Nuclear Information System (INIS)

    Yokota, Yoshio; Yamagishi, Yoshiaki; Kuroha, Mitsuo; Inoue, Tatsuya

    1995-01-01

    In FBRs there are two methods of storing and cooling the spent fuel - the in-vessel storage and the ex-vessel storage. Because of the sodium leaks through the tank at the beginning of pre-operation, the utilization of the ex-vessel fuel storage tank (EVST) of some FBR plant has been changed from the ex-vessel fuel storage to the interim fuel transfer tank. This led to reactor designers focusing on the material, structure and fabrication of the carbon steel sodium storage tanks worldwide. The Monju EVST was at the final stage of the design, when the leaks occurred. The lesson learned from that experience and the domestic fabrication technology are reflected to the design and fabrication of the Monju EVST. This paper describes the design, fabrication and R and D results for the tank, and operating experience in functional test. The items to be examined are as follows: (1) Overall structure of the tank and design philosophy on the function, (2) Structure of the cover shielding plug and its design philosophy, (3) Structures of the rotating rack and its bearings, and their design philosophy, (4) Cooling method and its design philosophy, (5) Structure and fabrication of the cooling coil support inside EVST with comparison of leaked case, (6) R and D effort for items above. The fabrication of the Monju EVST started in August 1986 and it was shipped to the site in March 1990. Installation was completed in November 1990, and sodium fill after pre-heating started in 1991. The operation has been continued since September 1992. In 1996 when the first spent fuel is stored, its total functions will be examined. (author)

  17. Investigation of small scale sphere-pac fuel fabrication plant with external gelation process

    International Nuclear Information System (INIS)

    Maekawa, Kazuhiko; Yoshimura, Tadahiro; Kikuchi, Toshiaki; Hoshino, Yasushi; Munekata, Hideki; Shimizu, Makoto

    2005-02-01

    In feasibility studies on commercialized FBR cycle system, comprehensive system investigation and properties evaluation for candidate FBR cycle systems have been implemented through view point of safety, economics, environmental burden reduction, non-proliferation resistivity, etc. As part of these studies, an investigation of small scale sphere-pac fuel fabrication plant with external gelation process was conducted. Until last fiscal year, equipment layout in cells and overall layout design of the 200t-HM/y scale fuel fabrication plant were conducted as well as schematical design studies on main equipments in gelation and reagent recovery processes of the plant. System property data concerning economics and environmental burden reduction of fuel fabrication plant was also acquired. In this fiscal year, the processes from vibropacking to fuel assemblies storage were added to the investigation range, and a conceptual design of whole fuel fabrication plant was studied as well as deepening the design study on main equipments. The conceptual design study was mainly conducted for small 50t-HM/y scale plant and a revising investigation was done for 200t-HM/y scale plant. Taking the planed comparative evaluation with pellet fuel fabrication system into account, design of equipments which should be equivalent with pellet system, especially in post-vibropacking processes, were standardized in each system. Based on these design studies, system properties data concerning economics and environmental burden reduction of the plant was also acquired. In comparison with existing design, the cell height was lowered on condition that plug type pneumatic system was adopted and fuel fabrication building was downsized by applying rationalized layout design of pellet system to post-vibropacking processes. Reduction of reagent usage at gelation process and rationalization of sintering and O/M controlling processes etc., are foremost tasks. (author)

  18. Research and development of nitride fuel cycle technology in Europe

    International Nuclear Information System (INIS)

    Wallenius, Janne

    2004-01-01

    Research and development on nitride fuels for minor actinide burning in accelerator driven systems is performed in Europe in context of the CONFIRM project. Dry and wet methods for fabrication of uranium free nitride fuels have been developed with the assistance of thermo-chemical modelling. Four (Pu, Zr) pins have been fabricated by PSI and will be irradiated in Studsvik at a rating of 40-50 kW/m. The thermal conductivity of (Pu, Zr)N has been measured and was found to be in agreement with earlier theoretical assessments. Safety modeling indicates that americium bearing nitride fuels, in spite of their relatively poor high temperature stability under atmospheric pressure, can survive power transients as long as the fuel cladding remains intact. (author)

  19. Status of research reactor fuel development in KAERI

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Ryu, Woo-Seok; Park, Jong-Man; Lee, Don-Bae; Kim, Ki-Hwan; Kuk, Il-Hyun

    1996-01-01

    The development of uranium silicide dispersion fuel fabrication technology has been carried out in KAERI. LEU fuel bundle was prepared for irradiation test. In order to compare the performance of atomized and comminuted U 3 Si dispersed fuels, the bundle of two kinds of fuel elements were prepared. Irradiation test will be performed in the OR-hole of HANARO in the near future. U 3 Si 2 atomization technology has been improved by using ceramic crucible and nozzle. Irradiation test for atomized U 3 Si 2 plate type fuel will be carried out in cooperation with ANL by using HANARO in connection with RERTR advanced fuel development. (author)

  20. Atomics International fuel fabrication facility and low enrichment program. Part 2

    International Nuclear Information System (INIS)

    Hassel, H.W.

    1993-01-01

    Most of you know our company from the last meeting in May in Vienna, so I won't steal your time with explaining and demonstrating the same techniques that we have heard this morning f rom the other speakers. I would just take some words to explain the order of business with highly enriched uranium. NUKEM handles around almost two tons of highly enriched uranium a year and it was necessary to satisfy all the new physical protection philosophies. That means that we have to install storage and safe fabrication sites for a lot of money, 2.5 meter thick concrete walls, and different alarm systems. So just to demonstrate how silly this business is, we have just overcome this for highly enriched uranium, and now we speak about low enriched uranium for which we don't need all of these investments to make this business safe. I would just like to concentrate my words on the status of fabrication and considerations in my company concerning the medium enriched uranium and low enriched uranium. In TABLE I are the different fuel types (see column 1) and then we have the fabrication in column 2; (The reason that I use the blackboard this morning is that I try to demonstrate all the techniques. However, all the speakers before me did this and in theory we are not so far away from each other.) the experience of my company in kg. In column 3 is the irradiation experience of these fuels types. Column 4 shows the studies and calculations made in our company for lower and medium enriched fuels. The preliminary fabrication tests and calculations are in column 5, and in column 6 we have the delivery time for a prototype core in months after UF 6 supply. Column 7 shows the time for the development of specifications including irradiation time in years for 6 and 7, and column 8 is the estimated cost of 6 and 7. There is just one fuel that is not in this summary and that is U-Zr

  1. Development of manufacturing equipment and QC equipment for DUPIC fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Park, J.J.; Lee, J.W.; Kim, S.S.; Yim, S.P.; Kim, J.H.; Kim, K.H.; Na, S.H.; Kim, W.K.; Shin, J.M.; Lee, D.Y.; Cho, K.H.; Lee, Y.S.; Sohn, J.S.; Kim, M.J.

    1999-05-01

    In this study, DUPIC powder and pellet fabrication equipment, welding system, QC equipment, and fission gas treatment are developed to fabricate DUPIC fuel at IMEF M6 hot cell. The systems are improved to be suitable for remote operation and maintenance with the manipulator at hot cell. Powder and pellet fabrication equipment have been recently developed. The systems are under performance test to check remote operation and maintenance. Welding chamber and jigs are designed and developed to remotely weld DUPIC fuel rod with manipulators at hot cell. Remote quality control equipment are being tested for analysis and inspection of DUPIC fuel characteristics at hot cell. And trapping characteristics is analyzed for cesium and ruthenium released under oxidation/reduction and sintering processes. The design criteria and process flow diagram of fission gas treatment system are prepared incorporating the experimental results. The fission gas treatment system has been successfully manufactured. (Author). 33 refs., 14 tabs., 91 figs

  2. Cost evaluation of a commercial-scale DUPIC fuel fabrication facility (Part I) -Summary

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-08-01

    A conceptual design of a commercial scale DUPIC fuel fabrication facility was initiated to provide some insights into the costs associated with construction, operation, and decommissioning. The primary conclusion of this report is that it is feasible to design, license, construct, test, and operate a facility that will process 400 MTHE/yr of spent PWR fuel and reconfigure the fuel into CANDU fuel bundles at a reasonable unit cost of the fuel material. Although DUPIC fuel fabrication by vibropacking method is clearly cheaper than that of the pellet method, the feasibility of vibropac technology for DUPIC fuel fabrication and use of vibroac fuel in CANDU reactors may has to be studied in depth in order to use as an alternative to the conventional pellet fuel method. Especially, there are some questions on meeting the CANDU requirements in thermal and mechanical terms as well as density of fuel. Wherever possible, this report used representative costs of currently available technologies as the bases for cost estimation. It should also be noted that the conceptual design and cost information contained in this report was extracted from the public domain and general open literature. Later studies have to focus on other important areas of concern such as safety, security, safeguards, process optimization etc. 7 figs., 6 tabs. (Author)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  4. Development of the high temperature sintering furnace for DUPIC fuel fabrication

    International Nuclear Information System (INIS)

    Lee, Jung Won; Kim, B. G.; Park, J. J.; Yang, M. S.; Kim, K. H.; Kim, J. H.; Cho, K. H.; Lee, D. Y.; Lee, Y. S.

    1998-11-01

    This report describes the development of the high temperature sintering furnace for manufacturing DUPIC (Direct Use of spent PWR fuel in CANDU reactors) fuel pellets. The furnace has to be remotely operated and maintained in a high radioactive hot cell using master-slave manipulators. The high temperature sintering furnace for manufacturing DUPIC fuel pellets, which is satisfied with the requirements of remote operation and maintenance in a hot cell, was successfully developed and installed in the M6 hot cell at IMEF (Irradiated Material Examination Facility). The functional and thermal performance test was also successfully completed. The technology accumulated during developing this sintering furnace became the basis of other DUPIC equipment development, and will be very helpful in the development of equipment for use in hot cell in the future. (author). 20 figs

  5. Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

    Science.gov (United States)

    Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

    2017-12-01

    A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

  6. Development of uranium reduction system for incineration residue generated at LWR nuclear fuel fabrication plants in Japan

    International Nuclear Information System (INIS)

    Sampei, T.; Sato, T.; Suzuki, N.; Kai, H.; Hirata, Y.

    1993-01-01

    The major portion of combustible solid wastes generated at LWR nuclear fuel fabrication plants in Japan is incinerated and stored in a warehouse. The uranium content in the incineration residue is higher compared with other categories of wastes, although only a small amount of incineration residue is generated. Hence, in the future uranium should be removed from incineration residues before they are reduced to a level appropriate for the final disposal. A system for processing the incineration residue for uranium removal has been developed and tested based on the information obtained through laboratory experiments and engineering scale tests

  7. Fabrication of simulated plate fuel elements: Defining role of out-of-plane residual shear stress

    Energy Technology Data Exchange (ETDEWEB)

    Rakesh, R., E-mail: rakesh.rad87@gmail.com [DAE Graduate Fellows, IIT Bombay, Powai, Mumbai 400076 (India); Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Kohli, D. [DAE Graduate Fellows, IIT Bombay, Powai, Mumbai 400076 (India); Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Sinha, V.P.; Prasad, G.J. [Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Samajdar, I. [Department of Metallurgical Engineering and Materials Science, IIT Bombay, Powai, Mumbai 400076 (India)

    2014-02-01

    Bond strength and microstructural developments were investigated during fabrication of simulated plate fuel elements. The study involved roll bonding of aluminum–aluminum (case A) and aluminum–aluminum + yttria (Y{sub 2}O{sub 3}) dispersion (case B). Case B approximated aluminum–uranium silicide (U{sub 3}Si{sub 2}) ‘fuel-meat’ in an actual plate fuel. Samples after different stages of fabrication, hot and cold rolling, were investigated through peel and pull tests, micro-hardness, residual stresses, electron and micro-focus X-ray diffraction. Measurements revealed a clear drop in bond strength during cold rolling: an observation unique to case B. This was related to significant increase in ‘out-of-plane’ residual shear stresses near the clad/dispersion interface, and not from visible signatures of microstructural heterogeneities.

  8. Description of a reference mixed oxide fuel fabrication plant (MOFFP)

    International Nuclear Information System (INIS)

    1978-01-01

    In order to evaluate the environment impact, due to the Mixed Oxide Fuel Fabrication Plants, work has been initiated to describe the general design and operating conditions of a reference Mixed Oxide Fuel Fabrication Plant (MOFFP) for the 1990 time frame. The various reference data and basic assumptions for the reference MOFFP plant have been defined after discussion with experts. The data reported in this document are only made available to allow an evaluation of the environmental impact due to a reference MOFFP plant. These data have therefore not to be used as recommandation, standards, regulatory guides or requirements

  9. The development of flow test technology for PWR fuel assembly

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

    1988-05-01

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

  10. The fabrication of nuclear fuel elements in Mexico

    International Nuclear Information System (INIS)

    Guerrero Morillo, H.L.

    1977-01-01

    The situation of the nucleoelectrical generation in Mexico by 1976 is described: two nuclear reactors under construction but no defined program on the type and start-up dates for the next power plants. However the existence of a general plan on nuclear power plants is mentioned, which, according to the last estimates reaches to 10,000 MW installed by 1990. The national intension, definitely expressed in the Law, is to supply domestic nuclear fuel to the power reactors operating in the country, starting with the first reload for the two BWR's at the first national station in Laguna Verde, which will be required at the end of 1981 and of 1982, respectively. Before such circumstances and the relatively short amounts of fuel elements that should be produced for those two unique reactors, Mexico already has to adopt a strategy to follow in respect to fuel elements fabrication. The two main options are analyzed: 1. To delay the local fabrication until a National Nuclear Program may be defined, meanwhile purchasing abroad the necessary reloads and initial cores; and 2. To start as soon as possible the local fuel elements fabrication in order to supply fuel for the first reload of the first unit of Laguna Verde, confronting the economical risks of such posture with the advantages of an immediate action. Both options are analyzed in detail comparing them specially under the economic point of view, standing out immediately the big effect of some factors which are economically imponderable, as experience and independance that would be gained with the second option. Emphasis is made on the advantages and risks of any case. According to the first option and once a National Program is defined, the work would be heavy but of simple strategy. On the contrary, the second option requires the adoption of a more complicated strategy, as either the project of the factory as its initial operation should be made under transient conditions, in view of the expected future expansion still

  11. Radiological surveillance in the nuclear fuel fabrication in Mexico

    International Nuclear Information System (INIS)

    Garcia A, J.; Reynoso V, R.; Delgado A, G.

    1996-01-01

    The objective of this report is to present the obtained results related to the application of the radiological safety programme established at the Nuclear Fuel Fabrication Pilot Plant (NFFPF) in Mexico, such as: surveillance methods, radiological protection criteria and regulations, radiation control and records and the application of ALARA recommendation. During the starting period from April 1994 to April 1995, at the NFFPF were made two nuclear fuel bundles a Dummy and other to be burned up in a BWR the mainly process activities are: UO 2 powder receiving, powder pressing for the pellets formation, pellets grinding, cleaning and drying, loading into a rod, Quality Control testing, nuclear fuel bundles assembly. The NFFPF is divided into an unsealed source area (pellets manufacturing plant) and into a sealed source area (rods fabrication plant). The control followed have helped to detect failures and to improve the safety programme and operation. (authors). 1 ref., 3 figs

  12. Further developments of PWR and BWR fuel elements

    International Nuclear Information System (INIS)

    Sofer, G.A.; Busselman, G.J.; Federico, L.J.

    1988-01-01

    The performance, safety, and economy of nuclear power plants in inluenced very decisively by the quality of their fuel elements. This is why quality assurance in fuel fabrication has been a factor of great importance from the outset. Operating experince and more stringent performance requirements have resulted in a continuous process of further development of fuel elements, which has been reflected also in lower and lower failure rates and increasingly higher burn-ups. Next to further development also innovation has been an important factor contributing to the present high quality level of fuel elements, which also has allowed fuel cycle costs to be decreased quite considerably. (orig.) [de

  13. The industrial production of fuel elements; La fabrication en france des elements combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Boussard, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Nadal, J [Societe Industrielle de Combustible Nucleaire (SICN), 75 - Paris (France); Pellen, A [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques (CERCA), 75 - Paris (France)

    1964-07-01

    -pool type reactors. The authors show how the problem of the industrial production of rolled fuel elements has been solved in France, and give the three steps involved: 1 - Assembly of the plates made in the U.S.A., 2 - Rolling of the cores made in the U.S.A. to obtain the plates, 3 - Fabrication of the U-Al alloy and production of the cores. They then recall briefly the characteristics of the different fuel elements now in production. A description is given of the various stages of the production including information about the equipment; stress is laid on the extent of the controls carried out at each stage. In conclusion the authors consider the future development of this type of production taking into account the improvements planned and those which are possible. (authors) [French] Les auteurs traitent successivement de la fabrication industrielle des elements combustibles pour reacteurs de puissance de la filiere U naturel graphite-gaz et plus particulierement pour les centrales energetiques d'E.D.F. et de celle des elements combustibles a base d'U enrichi destines aux reacteurs experimentaux du type 'piscine'. 1ere Partie - LES ELEMENTS COMBUSTIBLES AVANCES POUR LES REACTEURS E.D.F.: Apres un bref rappel des caracteristiques des elements combustibles actuellement fabriques industriellement pour les reacteurs de MARCOULE et de CHINON, les auteurs indiquent les differentes etapes suivies pour aboutir au stade de la fabrication industrielle d'un element combustible nouveau, tant en ce qui concerne la gaine et eventuellement la chemise de graphite que le combustible lui-meme. Pour ce qui est de l'elaboration du combustible, ils decrivent les differentes operations en insistant sur les points originaux de la fabrication et de l'appareillage tels que: - coulees en moules chauds, - traitement thermique des alliages U.Mo 1 p. 100, - soudure des pastilles de fermeture des tubes, - gainage - controle aux differents stades. En ce qui concerne la fabrication des gaines, ils

  14. Environmental aspects based on operation performance of nuclear fuel fabrication facilities

    International Nuclear Information System (INIS)

    2001-07-01

    This publication was prepared within the framework of the IAEA Project entitled Development and Upgrading of Guidelines, Databases and Tools for Integrating Comparative Assessment into Energy System Analysis and Policy Making, which included the collection, review and input of data into a database on health and environmental impacts related to operation of nuclear fuel cycle facilities. The objectives of the report included assembling environmental data on operational performance of nuclear fabrication facilities in each country; compiling and arranging the data in a database, which will be easily available to experts and the public; and presenting data that may be of value for future environmental assessment of nuclear fabrication facilities

  15. Fast breeder fuel element development

    International Nuclear Information System (INIS)

    Marth, W.; Muehling, G.

    1983-08-01

    This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

  16. Caramel fuel for research reactors: experience acquired in the fabrication, monitoring and irradiation of Osiris core

    International Nuclear Information System (INIS)

    Contenson, Ghislain de; Foulquier, Henri; Trotabas, Maria; Vignesoult, Nicole; Cerles, J.-M.; Delafosse, Jacques.

    1981-06-01

    A plate type nuclear fuel (Caramel fuel) has been developed in France in the framework of the various activities pursued in the design, fabrication and development of nuclear fuels by the CEA. This fuel can be adapted to various different categories of water cooled reactor (power reactors, marine propulsion reactors, urbain heating reactors, research reactors). The successful work conducted in this field led the realization of a complete core and reloads for the high performance research reactor, Osiris, at Saclay. The existing highly enriched U-Al alloy fuel was replaced by a non-proliferating low enrichment (7%) caramel fuel. This new core has been operating successfully since january 1980. A brief description of Caramel and its main advantages is given. The way in which it is fabricated is described together with the quality controls to which it is subjected. The qualification program and the main results deduced from it are also presented. The program used to monitor its in-pile behavior is described. The essential purpose of this program is to ensure the high performance of the fuel under irradiation. The successful operation of Osiris, which terminated 11 irradiation cycles on the 21st of April 1981 confirmed the correctness of the decisions made and the excellent performance that could be achieved with these fuel elements under the severe conditions encountered in a high performance research reactor [fr

  17. Nondestructive assay of special nuclear material for uranium fuel-fabrication facilities

    International Nuclear Information System (INIS)

    Smith, H.A. Jr.; Schillebeeckx, P.

    1997-01-01

    A high-quality materials accounting system and effective international inspections in uranium fuel-fabrication facilities depend heavily upon accurate nondestructive assay measurements of the facility's nuclear materials. While item accounting can monitor a large portion of the facility inventory (fuel rods, assemblies, storage items), the contents of all such items and mass values for all bulk materials must be based on quantitative measurements. Weight measurements, combined with destructive analysis of process samples, can provide highly accurate quantitative information on well-characterized and uniform product materials. However, to cover the full range of process materials and to provide timely accountancy data on hard-to-measure items and rapid verification of previous measurements, radiation-based nondestructive assay (NDA) techniques play an important role. NDA for uranium fuel fabrication facilities relies on passive gamma spectroscopy for enrichment and U isotope mass values of medium-to-low-density samples and holdup deposits; it relies on active neutron techniques for U-235 mass values of high-density and heterogeneous samples. This paper will describe the basic radiation-based nondestructive assay techniques used to perform these measurements. The authors will also discuss the NDA measurement applications for international inspections of European fuel-fabrication facilities

  18. Glovebox glove deterioration in the Hanford Engineering Development Laboratory fuel fabrication facility

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.; Smith, R.C.; Powell, D.L.

    1979-07-01

    Neoprene glovebox gloves have been found susceptible to periodic rapid deterioration under normal operating conditions in fuel fabrication facilities. Examinations of glove failure histories and measurements of the atmospheres in inert atmosphere dry-boxes indicated ozone at low concentrations of 100 to 500 ppB was probably the most important factor in rapid glove deterioration. Testing of a varity of new glove materials indicated that Hypalon and ethylene-propylene-diamine monomer (EDPM) gloves have greater than 30 times the longevity of neoprene in low-level ozone concentration atmospheres. comparative tests over a 30-month period have also confirmed that the two glove candidates have a significantly longer operative life. 14 figures

  19. Characterization of aerosols from industrial fabrication of mixed-oxide nuclear reactor fuels

    International Nuclear Information System (INIS)

    Hoover, M.D.; Newton, G.J.

    1997-01-01

    Recycling plutonium into mixed-oxide (MOX) fuel for nuclear reactors is being given serious consideration as a safe and environmentally sound method of managing plutonium from weapons programs. Planning for the proper design and safe operation of the MOX fuel fabrication facilities can take advantage of studies done in the 1970s, when recycling of plutonium from nuclear fuel was under serious consideration. At that time, it was recognized that the recycle of plutonium and uranium in irradiated fuel could provide a significant energy source and that the use of 239 Pu in light water reactor fuel would reduce the requirements for enriched 235 U as a reactor fuel. It was also recognized that the fabrication of uranium and plutonium reactor fuels would not be risk-free. Despite engineered safety precautions such as the handling of uranium and plutonium in glove-box enclosures, accidental releases of radioactive aerosols from normal containment might occur. Workers might then be exposed to the released materials by inhalation

  20. Development of molten carbonate fuel cells for power generation

    Science.gov (United States)

    1980-04-01

    The broad and comprehensive program included elements of system definition, cell and system modeling, cell component development, cell testing in pure and contaminated environments, and the first stages of technology scale up. Single cells, with active areas of 45 sq cm and 582 sq cm, were operated at 650 C and improved to state of the art levels through the development of cell design concepts and improved electrolyte and electrode components. Performance was shown to degrade by the presence of fuel contaminants, such as sulfur and chlorine, and due to changes in electrode structure. Using conventional hot press fabrication techniques, electrolyte structures up to 20" x 20" were fabricated. Promising approaches were developed for nonhot pressed electrolyte structure fabrication and a promising electrolyte matrix material was identified. This program formed the basis for a long range effort to realize the benefits of molten carbonate fuel cell power plants.

  1. Treat upgrade fuel fabrication

    International Nuclear Information System (INIS)

    Davidson, K.V.; Schell, D.H.

    1979-01-01

    An extrusion and thermal treatment process was developed to produce graphite fuel rods containing a dispersion of enriched UO 2 . These rods will be used in an upgraded version of the Transient Reactor Test Facility (TREAT). The improved fuel provides a higher graphite matrix density, better fuel dispersion and higher thermal capabilities than the existing fuel

  2. Coordinated irradiation plan for the Fuel Refabrication and Development Program

    International Nuclear Information System (INIS)

    Barner, J.O.

    1979-04-01

    The Department of Energy's Fuel Refabrication and Development (FRAD) Program is developing a number of proliferation-resistant fuel systems and forms for alternative use in nuclear reactors. A major portion of the program is the development of irradiation behavioral information for the fuel system/forms with the ultimate objective of qualifying the design for licensing and commercial utilization. The nuclear fuel systems under development include denatured thoria--urania fuels and spiked urania--plutonia or thoria--plutonia fuels. The fuel forms being considered include pellet fuel produced from mechanically mixed or coprecipitated feed materials, pellet fuel fabricated from partially calcined gel-derived or freeze-dried spheres (hybrid fuel) and packed-particle fuel produced from sintered gel-derived spheres (sphere-pac). This document describes the coordinated development program that will be used to test and demonstrate the irradiation performance of alternative fuels

  3. Statistical methods to assess and control processes and products during nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Weidinger, H.

    1999-01-01

    Very good statistical tools and techniques are available today to access the quality and the reliability of fabrication process as the original sources for a good and reliable quality of the fabricated processes. Quality control charts of different types play a key role and the high capability of modern electronic data acquisition technologies proved, at least potentially, a high efficiency in the more or less online application of these methods. These techniques focus mainly on stability and the reliability of the fabrication process. In addition, relatively simple statistical tolls are available to access the capability of fabrication process, assuming they are stable, to fulfill the product specifications. All these techniques can only result in as good a product as the product design is able to describe the product requirements necessary for good performance. Therefore it is essential that product design is strictly and closely performance oriented. However, performance orientation is only successful through an open and effective cooperation with the customer who uses or applies those products. During the last one to two decades in the west, a multi-vendor strategy has been developed by the utility, sometimes leading to three different fuel vendors for one reactor core. This development resulted in better economic conditions for the user but did not necessarily increase an open attitude with the vendor toward the using utility. The responsibility of the utility increased considerably to ensure an adequate quality of the fuel they received. As a matter of fact, sometimes the utilities had to pay a high price because of unexpected performance problems. Thus the utilities are now learning that they need to increase their knowledge and experience in the area of nuclear fuel quality management and technology. This process started some time ago in the west. However, it now also reaches the utilities in the eastern countries. (author)

  4. LWR fuel fabrication: a mature and competitive industry

    International Nuclear Information System (INIS)

    Schwartz, M.H.

    1997-01-01

    The pressures on fuel fabricators - to avoid losing existing clients as well as to win any new business that is put up to tender in this overly supplied market - is driving them to reduce costs and to improve designs and performance. (author)

  5. Gel-sphere-pac fuel for thermal reactors: assessment of fabrication technology and irradiation performance

    Energy Technology Data Exchange (ETDEWEB)

    Beatty, R.L. Norman, R.E.; Notz, K.J. (comps.)

    1979-11-01

    Recent interest in proliferation-resistant fuel cycles for light-water reactors has focused attention on spiked plutonium and /sup 233/U-Th fuels, requiring remote refabrication. The gel-sphere-pac process for fabricating metal-clad fuel elements has drawn special attention because it involves fewer steps. Gel-sphere-pac fabrication technology involves two major areas: the preparation of fuel spheres of high density and loading these spheres into rods in an efficiently packed geometry. Gel sphere preparation involves three major steps: preparation of a sol or of a special solution (broth), gelation of droplets of sol or broth to give semirigid spheres of controlled size, and drying and sintering these spheres to a high density. Gelation may be accomplished by water extraction (suitable only for sols) or ammonia gelation (suitable for both sols and broths but used almost exclusively with broths). Ammonia gelation can be accomplished either externally, via ammonia gas and ammonium hydroxide, or internally via an added ammonia generator such as hexamethylenetetramine. Sphere-pac fuel rod fabrication involves controlled blending and metering of three sizes of spheres into the rod and packing by low- to medium-energy vibration to achieve about 88% smear density; these sizes have diametral ratios of about 40:10:1 and are blended in size fraction amounts of about 60% coarse, 18% medium, and 22% fine. Irradiation test results indicate that sphere-pac fuel performs at least as well as pellet fuel, and may in fact offer an advantage in significantly reducing mechanical and chemical interaction between the fuel and cladding. The normal feed for gel sphere preparation, heavy metal nitrate solution, is the usual product of fuel reprocessing, so that fabrication of gel spheres performs all the functions performed by both conversion and pellet fabrication in the case of pellet technology.

  6. Gel-sphere-pac fuel for thermal reactors: assessment of fabrication technology and irradiation performance

    International Nuclear Information System (INIS)

    Beatty, R.L.; Norman, R.E.; Notz, K.J.

    1979-11-01

    Recent interest in proliferation-resistant fuel cycles for light-water reactors has focused attention on spiked plutonium and 233 U-Th fuels, requiring remote refabrication. The gel-sphere-pac process for fabricating metal-clad fuel elements has drawn special attention because it involves fewer steps. Gel-sphere-pac fabrication technology involves two major areas: the preparation of fuel spheres of high density and loading these spheres into rods in an efficiently packed geometry. Gel sphere preparation involves three major steps: preparation of a sol or of a special solution (broth), gelation of droplets of sol or broth to give semirigid spheres of controlled size, and drying and sintering these spheres to a high density. Gelation may be accomplished by water extraction (suitable only for sols) or ammonia gelation (suitable for both sols and broths but used almost exclusively with broths). Ammonia gelation can be accomplished either externally, via ammonia gas and ammonium hydroxide, or internally via an added ammonia generator such as hexamethylenetetramine. Sphere-pac fuel rod fabrication involves controlled blending and metering of three sizes of spheres into the rod and packing by low- to medium-energy vibration to achieve about 88% smear density; these sizes have diametral ratios of about 40:10:1 and are blended in size fraction amounts of about 60% coarse, 18% medium, and 22% fine. Irradiation test results indicate that sphere-pac fuel performs at least as well as pellet fuel, and may in fact offer an advantage in significantly reducing mechanical and chemical interaction between the fuel and cladding. The normal feed for gel sphere preparation, heavy metal nitrate solution, is the usual product of fuel reprocessing, so that fabrication of gel spheres performs all the functions performed by both conversion and pellet fabrication in the case of pellet technology

  7. Cermet fuel for fast reactor – Fabrication and characterization

    Energy Technology Data Exchange (ETDEWEB)

    Mishra, Sudhir, E-mail: sudhir@barc.gov.in [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Kutty, P.S.; Kutty, T.R.G. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Das, Shantanu [Uranium Extraction Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Dey, G.K. [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Kumar, Arun [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2013-11-15

    (U, Pu)O{sub 2} ceramic fuel is the well-established fuel for the fast reactors and (U, Pu, Zr) metallic fuel is the future fuel. Both the fuels have their own merits and demerits. Optimal solution may lie in opting for a fuel which combines the favorable features of both fuel systems. The choice may be the use of cermet fuel which can be either (U, PuO{sub 2}) or (Enriched U, UO{sub 2}). In the present study, attempt has been made to fabricate (Natural U, UO{sub 2}) cermet fuel by powder metallurgy route. Characterization of the fuel has been carried out using dilatometer, differential thermal analyzer, X-ray diffractometer, and Scanning Electron Microscope. The results show a high solidus temperature, high thermal expansion, presence of porosities, etc. in the fuel. The thermal conductivity of the fuel has also been measured. X-ray diffraction study on the fuel compact reveals presence of α U and UO{sub 2} phases in the matrix of the fuel.

  8. Pilot-scale demonstration of the modified direct denitration process to prepare uranium oxide for fuel fabrication evaluation

    International Nuclear Information System (INIS)

    Kitts, F.G.

    1994-04-01

    The Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) Program has the objective of developing a cost-competitive enrichment process that will ultimately replace the gaseous diffusion process used in the United States. Current nuclear fuel fabricators are set up to process only the UF 6 product from gaseous diffusion enrichment. Enriched uranium-iron alloy from the U-AVLIS separator system must be chemically converted into an oxide form acceptable to these fabricators to make fuel pellets that meet American Society for Testing and Materials (ASTM) and utility company specifications. A critical step in this conversion is the modified direct denitration (MDD) that has been selected and presented in the AVLIS Conceptual Design for converting purified uranyl nitrate to UO 3 to be shipped to fabricators for making UO 2 pellets for power reactor fuel. This report describes the MDD process, the equipment used, and the experimental work done to demonstrate the conversion of AVLIS product to ceramic-grade UO 3 suitable for making reactor-grade fuel pellets

  9. Key differences in the fabrication of US and German TRISO-coated particle fuel, and their implications on fuel performance

    International Nuclear Information System (INIS)

    Petti, D.A.; Buongiorno, J.; Maki, J.T.; Miller, G.K.; Hobbins, R.R.

    2002-01-01

    Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the US. German fuel generally displayed in-pile gas release values that were three orders of magnitude lower than US fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the US and German and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the US fuel has not faired as well, and what process/production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer US irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, degree of acceleration, power per particle) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance. (author)

  10. Review of training methods employed in nuclear fuel fabrication plants

    International Nuclear Information System (INIS)

    Box, W.D.; Browder, F.N.

    1975-01-01

    A search of the literature through the Nuclear Safety Information Center revealed that 86 percent of the incidents that have occurred in fuel fabrication plants can be traced directly or indirectly to insufficient operator training. In view of these findings, a review was made of the training programs now employed by the nuclear fuel fabrication industry. Most companies give the new employee approximately 20 hours of orientation courses, followed by 60 to 80 hours of on-the-job training. It was concluded that these training programs should be expanded in both scope and depth. A proposed program is outlined to offer guidance in improving the basic methods currently in use

  11. The series production in a standardized fabrication line for silicide fuels and commercial aspects

    International Nuclear Information System (INIS)

    Wehner, E.L.; Hassel, H.W.

    1987-01-01

    NUKEM has been responsible for the development and fabrication of LEU fuel elements for MTR reactors under the frame of the German AF program since 1979. The AF program is part of the international RERTR efforts, which were initiated by the INFCE Group in 1978. This paper describes the actual status of development and the transition from the prototype to the series production in a standardized manufacturing line for silicide fuels at NUKEM. Technical provisions and a customer oriented standardized product range aim at an economized manufacturing. (Author)

  12. Fabrication and characterization of MX-type fuels and fuel pins

    International Nuclear Information System (INIS)

    Richter, K.; Bartscher, W.; Benedict, U.; Gueugnon, J.F.; Kutter, H.; Sari, C.; Schmidt, H.E.

    1978-01-01

    This paper summarizes the most important fabrication parameters and characterization of fuel and fuel pins obtained during the investigation of uranium-plutonium carbides, oxicarbides, carbonitrides and nitrides in the past years at the European Institute for Transuranium Elements at Karlsruhe. All preparation methods discussed are based on carbothermic reduction of a mechanical blend of uranium-plutonium oxide and carbon powder. General data for carbothermic reduction processes are discussed (influence of starting material, homogeneity, control of degree of reaction, etc). A survey of different preparation methods investigated is given. Limitations with respect to temperature and atmosphere for both carbothermic reduction processes and sintering conditions for the different compounds are summarized. A special preparation process for mixed carbonitrides with low nitrogen content (U,Pu)sub(1-x)Nsub(x) in the range 0.1 0 C to 1400 0 C by means of a modulated electron beam technique. A scheme is proposed, which allows to predict the thermal properties of MX fuels on the basis of their chemical composition and porosity. Preparation, preirradiation characterization and final controls of fuel test pins for pellet and vibrocompacted type of pins are described and the most important data summarized for all advanced fuels irradiated at Dounreay (DN1) and Rapsodie Fast Reactor (DN2) within the TU irradiation programme

  13. Design and fabrication procedures of Super-Phenix fuel elements

    International Nuclear Information System (INIS)

    Leclere, J.; Vialard, J.-L.; Delpeyroux, P.

    1975-01-01

    For Super-Phenix fuel assemblies, Phenix technological arrangements will be used again, but they will be simplified as far as possible. The maximum fuel can temperature has been lowered in order to obtain a good behavior of hexagonal tubes and cans at high irradiation levels. An important experimental programme and the experience gained from Phenix operation will confirm the merits of the options retained. The fuel element fabrication is envisaged to take place in the plutonium workshop at Cadarache. Usual procedures will be employed and both reliability and automation will be increased [fr

  14. Fabrication and irradiation testing of LEU [low enriched uranium] fuels at CRNL status as of 1987 September

    International Nuclear Information System (INIS)

    Sears, D.F.; Berthiaume, L.C.; Herbert, L.N.

    1987-01-01

    The current status of Chalk River Nuclear Laboratories' (CRNL) program to develop and test low-enriched uranium (LEU), proliferation-resistant fuels for use in research reactors is reviewed. CRNL's fuel manufacturing process has been qualified by the successful demonstration irradiation of 7 full-size rods in the NRU reactor. Now industrial-scale production equipment has been commissioned, and a fuel-fabrication campaign for 30 NRU rods and a MAPLE-X core is underway. Excess capacity could be used for commercial fuel fabrication. In the irradiation testing program, mini-elements with deliberately included core surface defects performed well in-reactor, swelling by only 7 to 8 vol% at 93 atomic percent burnup of the original U-235. The additional restraint provided by the aluminium cladding which flowed into the defects during extrusion contributed to this good performance. Mini-elements containing a variety of particle size distributions were also successfully irradiated to 93 at% burnup in NRU, as part of a study to establish the optimum particle size distribution. Swelling was found to be proportional to the percentage of fines (<44μm particles) contained in the cores. The mini-elements containing the composition normally used at CRNL had swollen by 5.8 vol%, and mini-elements with a much higher percentage of fines had swollen by 6.8 vol%, at 93 at% burnup. Also, a program to develop LEU targets for Mo-99 production, via the technology developed to fabricate dispersed silicide fuel, has started, and preliminary scoping studies are underway. (Author)

  15. Comparative Study of Laboratory-Scale and Prototypic Production-Scale Fuel Fabrication Processes and Product Characteristics

    International Nuclear Information System (INIS)

    Marshall, Douglas W.

    2014-01-01

    An objective of the High Temperature Gas Reactor fuel development and qualification program for the United States Department of Energy has been to qualify fuel fabricated in prototypic production-scale equipment. The quality and characteristics of the tristructural isotropic (TRISO) coatings on fuel kernels are influenced by the equipment scale and processing parameters. The standard deviations of some TRISO layer characteristics were diminished while others have become more significant in the larger processing equipment. The impact on statistical variability of the processes and the products, as equipment was scaled, are discussed. The prototypic production-scale processes produce test fuels meeting all fuel quality specifications. (author)

  16. Process for the fabrication of a nuclear fuel

    International Nuclear Information System (INIS)

    Hirose, Yasuo.

    1970-01-01

    Herein disclosed is a process for fabricating a nuclear fuel incorporating either uranium or plutonium. A pellet-like substrate consisting of a packed powder ceramic fuel such as uranium or plutonium is prepared with the horizontal surface of the body provided with a masking. Next, after impregnating the substrate voids with a solution consisting of a fissile material or mixture of fissile material and poison, the solvent is removed by a chemical deposition process which causes the impregnated material to migrate through capillary action toward the vicinity of the fuel body surface. Sintering and pyrolysis of the deposited material and masking are subsequently carried out to yield a fuel body having adjacent to its surface an intensely concentrated layer of either fissile material or a mixture of fissile material and poison. (Owens, K.J.)

  17. Waste management state-of-the-art review for mixed-oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Woodsum, H.C.; Goodman, J.

    1977-11-01

    This report provides a state-of-the-art review of the waste management for mixed-oxide (MOX) fuel fabrication facilities. The intent of this report is to focus on those processes and regulatory issues which have a direct bearing on existing and anticipated future management of transuranic (TRU) wastes from a commercial MOX fuel fabrication faciity. Recent government agency actions are reviewed with regard to their impact on existing and projected waste management regulations; and it is concluded that acceleration in the development of regulations, standards, and criteria is one of the most important factors in the implementation of improved MOX plant waste management techniques. ERDA development programs pertaining to the management of TRU wastes have been reviewed and many promising methods for volume reduction of both solid and liquid wastes are discussed. For solid wastes, these methods include compaction, shredding and baling, combustion, acid digestion, and decontamination by electropolishing or by electrolytic treatment. For liquid wastes, treatment options include evaporation, drying, calcination, flocculation, ion exchange, filtration, reverse osmosis, combustion (of combustible organics), and bioprocessing. Based on this review, it is recommended that ERDA continue with its combustible solid waste volume reduction program and complete these development activities by 1979. Following this, a critical evaluation of solid waste volume reduction techniques should be made to select the most promising systems for a commercial MOX fuel facility

  18. Quality control in nuclear fuel fabrication on the inspection basis

    International Nuclear Information System (INIS)

    Fuentes S, A.

    1997-01-01

    Every plant productive of electric power requires the use of energetics for the transformation to electricity. In the nucleo electric plant the energetic is the uranium, in which it makes ensembles and is used as fuel in the reactor. To assure that the fuel ensembles fulfill the specifications and requirements of design stipulated in the nucleo electric plant is that under a quality control through inspections during the fabrication process. The purpose of this work is to study and verify that the lineaments of the standard 10 CFR 50 appendix B 'Quality assurement for nuclear plants' specially in the criteria 'Inspections' that is used to guarantee the quality of the ensembles. This standard is the one that rules every activity and operation inside the pilot plant and its established in the quality program in the production of nuclear fuel for the Laguna Verde plant. The quality of the assemble is verified through each one of the tests or inspections due to the importance of it in the fabrication of fuel. (Author)

  19. Evaluation of bioassay program at uranium fuel fabrication plants

    International Nuclear Information System (INIS)

    Biggs, D.

    1981-03-01

    Results of a comprehensive study of urinalysis, lung burden and personal air sample measurements for workers at a uranium fuel fabrication plant are presented. Correlations between measurements were found and regression models used to explain the relationship between lung burden, daily intakes and urinary excretions of uranium. Assuming the ICRP lung model, the lung burden histories of ten workers were used to estimate the amounts in each of the long-term compartments of the lung. Estimates of the half lives of each compartment and of the maximum relative contributions to the urine from each compartment are given. These values were then used to predict urinary excretions from the long-term compartments for workers at another fuel fabrication plant. The standard error of estimate compared well with the daily variation in urinary excretion. (author)

  20. A Review on the Fabrication of Electrospun Polymer Electrolyte Membrane for Direct Methanol Fuel Cell

    Directory of Open Access Journals (Sweden)

    Hazlina Junoh

    2015-01-01

    Full Text Available Proton exchange membrane (PEM is an electrolyte which behaves as important indicator for fuel cell’s performance. Research and development (R&D on fabrication of desirable PEM have burgeoned year by year, especially for direct methanol fuel cell (DMFC. However, most of the R&Ds only focus on the parent polymer electrolyte rather than polymer inorganic composites. This might be due to the difficulties faced in producing good dispersion of inorganic filler within the polymer matrix, which would consequently reduce the DMFC’s performance. Electrospinning is a promising technique to cater for this arising problem owing to its more widespread dispersion of inorganic filler within the polymer matrix, which can reduce the size of the filler up to nanoscale. There has been a huge development on fabricating electrolyte nanocomposite membrane, regardless of the effect of electrospun nanocomposite membrane on the fuel cell’s performance. In this present paper, issues regarding the R&D on electrospun sulfonated poly (ether ether ketone (SPEEK/inorganic nanocomposite fiber are addressed.

  1. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  2. Paper-based membraneless hydrogen peroxide fuel cell prepared by micro-fabrication

    Science.gov (United States)

    Mousavi Ehteshami, Seyyed Mohsen; Asadnia, Mohsen; Tan, Swee Ngin; Chan, Siew Hwa

    2016-01-01

    A paper-based membraneless single-compartment hydrogen peroxide power source prepared by micro-electromechanical systems (MEMS) technology is reported. The cell utilizes hydrogen peroxide as both fuel and oxidant in a low volume cell fabricated on paper. The fabrication method used is a simple method where precise, small-sized patterns are produced which include the hydrophilic paper bounded by hydrophobic resin. Open circuit potentials of 0.61 V and 0.32 V are achieved for the cells fabricated with Prussian Blue as the cathode and aluminium/nickel as the anode materials, respectively. The power produced by the cells is 0.81 mW cm-2 at 0.26 V and 0.38 mW cm-2 at 0.14 V, respectively, even after the cell is bent or distorted. Such a fuel cell provides an easily fabricated, environmentally friendly, flexible and cost saving power source. The cell may be integrated within a self-sustained diagnostic system to provide the on-demand power for future bio-sensing applications.

  3. Development of wire wrapping technology for FBR fuel pin

    International Nuclear Information System (INIS)

    Nogami, Tetsuya; Seki, Nobuo; Sawayama, Takeo; Ishibashi, Takashi

    1991-01-01

    For the FBR fuel assembly, the spacer wire is adopted to maintain the space between fuel pins. The developments have been carried out to achieve automatically wire wrapping with high precision. Based on the fundamental technology developed through the mock-up test operation, Joyo 'MK-I', fuel pin fabrication was started using partially mechanized wire wrapping machine in 1973. In 1978, an automated wire wrapping machine for Joyo 'MK-II' was developed by the adoption of some improvements for the wire inserting system to end plug hole and the precision of wire pitch. On the bases of these experiences, fully automated wire wrapping machine for 'Monju' fuel pin was installed at Plutonium Fuel Production Facility (PFPF) in 1987. (author)

  4. Fabrication and control of fuels made of mixed carbides (U, Pu)C

    International Nuclear Information System (INIS)

    Lorenzelli, R.; Delaroche, P.

    1980-01-01

    Fabrication of this type of advanced fuel is described. The fuel is prepared by reduction of oxides with carbon and natural sintering. Density, thermal stability and thermal conductibility are more particularly studied [fr

  5. Development of remote disassembly technology for liquid-metal reactor (LMR) fuel

    International Nuclear Information System (INIS)

    Bradley, E.C.; Evans, J.H.; Metz, C.F. III; Weil, B.S.

    1990-01-01

    A major objective of the Consolidated Fuel Reprocessing Program (CFRP) is to develop equipment and demonstrate technology to reprocess fast breeder reactor fuel. Experimental work on fuel disassembly cutting methods began in the 1970s. High-power laser cutting was selected as the preferred cutting method for fuel disassembly. Remotely operated development equipment was designed, fabricated, installed, and tested at Oak Ridge National Laboratory (ORNL). Development testing included remote automatic operation, remote maintenance testing, and laser cutting process development. This paper summarizes the development work performed at ORNL on remote fuel disassembly. 2 refs., 1 fig

  6. A Brief Description of High Temperature Solid Oxide Fuel Cell’s Operation, Materials, Design, Fabrication Technologies and Performance

    Directory of Open Access Journals (Sweden)

    Muneeb Irshad

    2016-03-01

    Full Text Available Today’s world needs highly efficient systems that can fulfill the growing demand for energy. One of the promising solutions is the fuel cell. Solid oxide fuel cell (SOFC is considered by many developed countries as an alternative solution of energy in near future. A lot of efforts have been made during last decade to make it commercial by reducing its cost and increasing its durability. Different materials, designs and fabrication technologies have been developed and tested to make it more cost effective and stable. This article is focused on the advancements made in the field of high temperature SOFC. High temperature SOFC does not need any precious catalyst for its operation, unlike in other types of fuel cell. Different conventional and innovative materials have been discussed along with properties and effects on the performance of SOFC’s components (electrolyte anode, cathode, interconnect and sealing materials. Advancements made in the field of cell and stack design are also explored along with hurdles coming in their fabrication and performance. This article also gives an overview of methods required for the fabrication of different components of SOFC. The flexibility of SOFC in terms fuel has also been discussed. Performance of the SOFC with varying combination of electrolyte, anode, cathode and fuel is also described in this article.

  7. Fabrication of uranium-plutonium mixed nitride fuel pins (88F-5A) for first irradiation test at JMTR

    International Nuclear Information System (INIS)

    Suzuki, Yasufumi; Iwai, Takashi; Arai, Yasuo; Sasayama, Tatsuo; Shiozawa, Ken-ichi; Ohmichi, Toshihiko; Handa, Muneo

    1990-07-01

    A couple of uranium-plutonium mixed nitride fuel pins was fabricated for the first irradiation tests at JMTR for the purpose of understanding the irradiation behavior and establishing the feasibility of nitride fuels as advanced FBR fuels. The one of the pins was fitted with thermocouples in order to observe the central fuel temperature. In this report, the fabrication procedure of the pins such as pin design, fuel pellet fabrication and characterizations, welding of fuel pins, and inspection of pins are described, together with the outline of the new TIG welder installed recently. (author)

  8. Powder fabrication of U-Mo alloys for nuclear dispersion fuels

    Energy Technology Data Exchange (ETDEWEB)

    Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

  9. Powder fabrication of U-Mo alloys for nuclear dispersion fuels

    International Nuclear Information System (INIS)

    Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes

    2011-01-01

    For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

  10. Review of training methods employed in nuclear fuel fabrication plants

    International Nuclear Information System (INIS)

    Box, W.D.; Browder, F.N.

    A search of the literature through the Nuclear Safety Information Center revealed that approximately 86 percent of the incidents that have occurred in fuel fabrication plants can be traced directly or indirectly to insufficient operator training. In view of these findings, a review was made of the training programs now employed by the nuclear fuel fabrication industry. Most companies give the new employee approximately 20 h of orientation courses, followed by 60 to 80 h of on-the-job training. It was concluded that these training programs should be expanded in both scope and depth. A proposed program is outlined to offer guidance in improving the basic methods currently in use. (U.S.)

  11. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    International Nuclear Information System (INIS)

    Souza, Jose Antonio Batista de

    2011-01-01

    IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm 3 for U 3 Si 2 -Al dispersion-based and 2.3 gU/cm 3 for U 3 O 8 -Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm 3 in U 3 Si 2 -Al dispersion and 3.2 gU/cm 3 U 3 O 8 -Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U 3 Si 2 -Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U 3 O 8 -Al dispersion fuel plates with 3.2 gU/cm 3 showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U 3 Si 2 production at 4.8 gU/cm 3 , with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

  12. Comparative Study of Laboratory-Scale and Prototypic Production-Scale Fuel Fabrication Processes and Product Characteristics

    International Nuclear Information System (INIS)

    2014-01-01

    An objective of the High Temperature Gas Reactor fuel development and qualification program for the United States Department of Energy has been to qualify fuel fabricated in prototypic production-scale equipment. The quality and characteristics of the tristructural isotropic coatings on fuel kernels are influenced by the equipment scale and processing parameters. Some characteristics affecting product quality were suppressed while others have become more significant in the larger equipment. Changes to the composition and method of producing resinated graphite matrix material has eliminated the use of hazardous, flammable liquids and enabled it to be procured as a vendor-supplied feed stock. A new method of overcoating TRISO particles with the resinated graphite matrix eliminates the use of hazardous, flammable liquids, produces highly spherical particles with a narrow size distribution, and attains product yields in excess of 99%. Compact fabrication processes have been scaled-up and automated with relatively minor changes to compact quality to manual laboratory-scale processes. The impact on statistical variability of the processes and the products as equipment was scaled are discussed. The prototypic production-scale processes produce test fuels that meet fuel quality specifications.

  13. Fabrication and characterization of CeO{sub 2} pellets for simulation of nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    García-Ostos, C.; Rodríguez-Ortiz, J.A. [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain); Arévalo, C., E-mail: carevalo@us.es [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain); Cobos, J. [CIEMAT, Avenida Complutense, 40, Madrid (Spain); Gotor, F.J. [Materials Science Institute of Seville (CSIC-US), Av. Américo Vespucio, 49, 41092 Seville (Spain); Torres, Y. [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain)

    2016-03-15

    Highlights: • CeO{sub 2} is presented as a surrogate material for UO{sub 2} to study nuclear fuel. • Powder-metallurgy methods are applied to fabricate CeO{sub 2} pellets with controlled porosity. • An optimization of the fabrication parameters is established. • Microstructural and tribo-mechanical characterizations are performed. • Properties are compared to those of the nuclear fuel. - Abstract: Cerium Oxide, CeO{sub 2}, has been shown as a surrogate material to understand irradiated Mixed Oxide (MOX) based matrix fuel for nuclear power plants due to its similar structure, chemical and mechanical properties. In this work, CeO{sub 2} pellets with controlled porosity have been developed through conventional powder-metallurgy process. Influence of the main processing parameters (binder content, compaction pressure, sintering temperature and sintering time) on porosity and volumetric contraction values has been studied. Microstructure and physical properties of sintered compacts have also been characterized through several techniques. Mechanical properties such as dynamic Young's modulus, hardness and fracture toughness have been determined and connected to powder-metallurgy parameters. Simulation of nuclear fuel after reactor utilization with radial gradient porosity is proposed.

  14. Development of fuel performance and thermal hydraulic technology

    International Nuclear Information System (INIS)

    Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

    2000-03-01

    Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

  15. The Fabrication Problem Of U3Si2-Al Fuel With Uranium High Loading

    International Nuclear Information System (INIS)

    Supardjo

    1996-01-01

    The quality of U 3 Si 2 -Al dispersion fuel product is the main aim for each fabricator. Low loading of uranium fuel element is easily fabricated, but with the increased, uranium loading, homogeneity of uranium distribution is difficult to achieve and it always formed white spots, blister, and dogboning in the fuel plates. The problem can be eliminated by the increasing treatment of the fuel/Al powder. The precise selection of fuel/Al particles diameter is needed indeed to make easier in the homogeneous process of powder and the porosities arrangement in the fuel plates. The increasing of uranium loading at constant meat thickness will increase the meat hardness, therefore to withdraw the dogboning forming, the use of harder cladding materials is necessity

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    Directory of Open Access Journals (Sweden)

    Chan Bock Lee

    2016-10-01

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

  18. Environmental control aspects for fabrication, reprocessing and waste disposal of alternative LWR and LMFBR fuels

    International Nuclear Information System (INIS)

    Nolan, A.M.; Lewallen, M.A.; McNair, G.W.

    1979-11-01

    Environmental control aspects of alternative fuel cycles have been analyzed by evaluating fabrication, reprocessing, and waste disposal operations. Various indices have been used to assess potential environmental control requirements. For the fabrication and reprocessing operations, 50-year dose commitments were used. Waste disposal was evaluated by comparing projected nuclide concentrations in ground water at various time periods with maximum permissible concentrations (MPCs). Three different fabrication plants were analyzed: a fuel fabrication plant (FFP) to produce low-activity uranium and uranium-thorium fuel rods; a plutonium fuel refabrication plant (PFRFP) to produce plutonium-uranium and plutonium-thorium fuel rods; and a uranium fuel refabrication plant (UFRFP) to produce fuel rods containing the high-activity isotopes 232 U and 233 U. Each plant's dose commitments are discussed separately. Source terms for the analysis of effluents from the fuel reprocessing plant (FRP) were calculated using the fuel burnup codes LEOPARD, CINDER and ORIGEN. Effluent quantities are estimated for each fuel type. Bedded salt was chosen for the waste repository analysis. The repository site is modeled on the Waste Isolation Pilot Program site in New Mexico. Wastes assumed to be stored in the repository include high-level vitrified waste from the FRP, packaged fuel residue from the FRP, and transuranic (TRU) contaminated wastes from the FFP, PFRFP, and UFRFP. The potential environmental significance was determined by estimating the ground-water concentrations of the various nuclides over a time span of a million years. The MPC for each nuclide was used along with the estimated ground-water concentration to generate a biohazard index for the comparison among fuel compositions

  19. Hydrogen-bromine fuel cell advance component development

    Science.gov (United States)

    Charleston, Joann; Reed, James

    1988-01-01

    Advanced cell component development is performed by NASA Lewis to achieve improved performance and longer life for the hydrogen-bromine fuel cells system. The state-of-the-art hydrogen-bromine system utilizes the solid polymer electrolyte (SPE) technology, similar to the SPE technology developed for the hydrogen-oxygen fuel cell system. These studies are directed at exploring the potential for this system by assessing and evaluating various types of materials for cell parts and electrode materials for Bromine-hydrogen bromine environment and fabricating experimental membrane/electrode-catalysts by chemical deposition.

  20. Technology developments for Japanese BWR MOX fuel utilization

    International Nuclear Information System (INIS)

    Oguma, M.; Mochida, T.; Nomata, T.; Asahi, K.

    1997-01-01

    The Long-Term Program for Research, Development and Utilization of Nuclear Energy established by the Atomic Energy Commission of Japan asserts that Japan will promote systematic utilization of MOX fuel in LWRs. Based on this Japanese nuclear energy policy, we have been pushing development of MOX fuel technology aimed at future full scale utilization of this fuel in BWRs. In this paper, the main R and D topics are described from three subject areas, MOX core and fuel design, MOX fuel irradiation behaviour, and MOX fuel fabrication technology. For the first area, we explain the compatibility of MOX fuel with UO 2 core, the feasibility of the full MOX core, and the adaptability of MOX design methods based on a mock-up criticality experiment. In the second, we outline the Tsuruga MOX irradiation program and the DOMO program, and suggest that MOX fuel behaviour is comparable to ordinary BWR UO 2 fuel behaviour. In the third, we examine the development of a fully automated MOX bundle assembling apparatus and its features. (author). 14 refs, 11 figs, 3 tabs

  1. Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

  2. Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung

    2016-01-01

    In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

  3. Low enrichment fuel development at INEL

    International Nuclear Information System (INIS)

    Newton, D.G.

    1993-01-01

    EG and G Idaho, Inc. is under contract to the Department of Energy to operate the Idaho National Engineering Laboratory (INEL). The INEL is located in southeastern Idaho. This facility has been operating since 1949 and was originally called the National Reactor Testing Station. Several contractors manage projects on this facility. Most projects at INEL are concerned with either reactor safety or irradiation testing. At Test Area North, for example, experiments are being conducted on the effects of loss of coolant. At the Test Reactor Area the ATR (Advanced Test Reactor) and ETR (Engineering Test Reactor) are used for irradiation testing and, of course, those of you working at Argonne will recognize the Experimental Breeder Reactors I and II. SPERT is an acronym for Special Power Excursion Reactor Test. A part of this former reactor facility has been converted into a fuel fabrication laboratory facility. At SPERT IV a miniature fabrication facility has been set up to duplicate the aluminide plate fuel processing line at Atomics International. In other words, a model of the supplier's processing has been created, so that what process changes are developed here can then be scaled up to production. The process is described showing: making UAI x powder, making compact for fuel core, making experimental fuel plate and compact assembly, inspection and testing the fuel plate. Main concern was related to possible swelling

  4. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  5. Research and development of nitride fuel cycle technology in Japan

    International Nuclear Information System (INIS)

    Minato, Kazuo; Arai, Yasuo; Akabori, Mitsuo; Tamaki, Yoshihisa; Itoh, Kunihiro

    2004-01-01

    The research on the nitride fuel was started for an advanced fuel, (U, Pn)N, for fast reactors, and the research activities have been expanded to minor actinide bearing nitride fuels. The fuel fabrication, property measurements, irradiation tests and pyrochemical process experiments have been made. In 2002 a five-year-program named PROMINENT was started for the development of nitride fuel cycle technology within the framework of the Development of Innovative Nuclear Technologies by the Ministry of Education, Culture, Sports, Science and Technology of Japan. In the research program PROMINENT, property measurements, pyrochemical process and irradiation experiments needed for nitride fuel cycle technology are being made. (author)

  6. Application of plasma deposition technology for nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Jung, I. H.; Moon, J. S.; Park, H. S.; Song, K. C.; Lee, C. Y.; Kang, K. H.; Ryu, H. J.; Kim, H. S.; Yang, M. S.

    2001-01-01

    Yttria-stabilized-zirconia (m.p. 2670.deg. C), was deposited by induction plasma spraying system with a view to develop a new nuclear fuel fabrication technology. To fabricate the dense pellets, the spraying condition was optimized through the process parameters such as, chamber pressure, plasma plate power, powder spraying distance, sheath gas composition, probe position particle size and its morphology. The results with a 5mm thick deposit on rectangular planar graphite substrates showed 97.11% theoretical density, when the sheath gas flow rate was Ar/H 2 120/20 L/min, probe position 8cm, particle size-75 μm and spraying distance 22cm. The microstructure of YSZ deposit by ICP was lamellae and columnar perpendicular to the spraying direction. In the bottom part near the substrate, small equiaxed grains bounded in a layer. In the middle part, relatively regular size of columnar grains with excellent bonding each other were distinctive

  7. Fast reactor fuel design and development

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  8. Development of the continuous casting technology for fabrication of the tubular fuels

    International Nuclear Information System (INIS)

    Kim, H. S.; Lee, Y. S.; Kim, C. K.; Lee, D. B.; Oh, S. J.

    2003-01-01

    In the irradiation test of the U-Mo dispersed nuclear fuel that is used as nuclear fuels for research reactors, it was recognized that the swelling due to reaction between U-Mo particle and Al matrix caused some failures of the fuel claddings. The development of new style nuclear fuel that could minimize the reaction between U-Mo particles and Al matrix was needed. Tube style nuclear fuel was judged to be suitable as new style nuclear fuel. We targeted to make U-Mo tube of diameter 10mm, thinner than 1mm thick, because temperature distribution of tube style nuclear fuel will be expected to have a good performance. We used continuous casting technology to make tube style nuclear fuel. In this research, we have tried to make tube using copper before we make U-Mo tube style nuclear fuel by continuous casting method. As a result of the experiment, we succeeded to make copper tube of diameter 10mm, thickness 1mm

  9. Radiation shielding calculation for the MOX fuel fabrication plant Melox

    International Nuclear Information System (INIS)

    Lee, Y.K.; Nimal, J.C.; Chiron, M.

    1994-01-01

    Radiation shielding calculation is an important engineering work in the design of the MOX fuel fabrication plant MELOX. Due to the recycle of plutonium and uranium from UO2 spent fuel reprocessing and the large capacity of production (120t HM/yr.), the shielding design requires more attention in this LWR fuel plant. In MELOX, besides several temporary storage facilities of massive fissile material, about one thousand radioactive sources with different geometries, forms, densities, quantities and Pu concentrations, are distributed through different workshops from the PuO 2 powder reception unit to the fuel assembly packing room. These sources, with or without close shield, stay temporarily in different locations, containers and glove boxes. In order to optimize the dimensions, the material and the cost of shield as well as to limit the calculation work in a reasonable engineer-hours, a calculation scheme for shielding design of MELOX is developed. This calculation scheme has been proved to be useful in consideration of the feedback from the evolutionary design and construction. The validated shielding calculations give a predictive but reliable radiation doses information. (authors). 2 figs., 10 refs

  10. Status of SFR Metal Fuel Development

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Lee, Byoung Oon; Kim, Ki Hwan; Kim, Sung Ho

    2013-01-01

    Conclusion: • Metal fuel recycling in SFR: - Enhanced utilization of uranium resource; - Efficient transmutation of minor actinides; - Inherent passive reactor safety; - Proliferation resistance with pyro-electrochemical fuel recycling. • Demonstration of technical feasibility of recycling TRU metal fuel by 2020: - Remote fuel fabrication; - Irradiation performance up to high burnup

  11. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    International Nuclear Information System (INIS)

    Jackson, D.R.; Meeuwsen, P.V.

    1997-09-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems' Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment

  12. Fabrication and use of zircaloy/tantalum-sheathed cladding thermocouples and molybdenum/rhenium-sheathed fuel centerline thermocouples

    International Nuclear Information System (INIS)

    Wilkins, S.C.; Sepold, L.K.

    1985-01-01

    The thermocouples described in this report are zircaloy/tantalum-sheathed and molybdenum/rhenium alloy-sheathed instruments intended for fuel rod cladding and fuel centerline temperature measurements, respectively. Both types incorporate beryllium oxide insulation and tungsten/rhenium alloy thermoelements. These thermocouples, operated at temperatures of 2000 0 C and above, were developed for use in the internationally sponsored Severe Fuel Damage test series in the Power Burst Facility. The fabrication steps for both thermocouple types are described in detail. A laser-welding attachment technique for the cladding-type thermocouple is presented, and experience with alternate materials for cladding and fuel therocouples is discussed

  13. A Review on the Fabrication of Electro spun Polymer Electrolyte Membrane for Direct Methanol Fuel Cell

    International Nuclear Information System (INIS)

    Junoh, H.; Jaafar, J.; Norddin, M.N.A.M.; Ismail, A.F.; Othman, M.H.D.; Rahman, M.A.; Yusof, N.; Salleh, W.N.W.; Junoh, H.; Jaafar, J.; Norddin, M.N.A.M.; Ismail, A.F.; Othman, M.H.D.; Rahman, M.A.; Yusof, N.; Salleh, W.N.W.; Hamid Ilbeygi, H.

    2014-01-01

    Proton exchange membrane (PEM) is an electrolyte which behaves as important indicator for fuel cell’s performance. Research and development (R and D) on fabrication of desirable PEM have burgeoned year by year, especially for direct methanol fuel cell (DMFC). However, most of the R and Ds only focus on the parent polymer electrolyte rather than polymer inorganic composites. This might be due to the difficulties faced in producing good dispersion of inorganic filler within the polymer matrix, which would consequently reduce the DMFC’s performance. Electro spinning is a promising technique to cater for this arising problem owing to its more widespread dispersion of inorganic filler within the polymer matrix, which can reduce the size of the filler up to nano scale. There has been a huge development on fabricating electrolyte nano composite membrane, regardless of the effect of electro spun nano composite membrane on the fuel cell’s performance. In this present paper, issues regarding the R and D on electro spun sulfonated poly (ether ether ketone) (SPEEK)/inorganic nano composite fiber are addressed.

  14. Interim report spent nuclear fuel retrieval system fuel handling development testing

    Energy Technology Data Exchange (ETDEWEB)

    Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

    1997-06-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

  15. The KNK II/1 fuel assembly NY-205: Compilation of the irradiation history and the fuel and fuel pin fabrication data of the INTERATOM data bank system BESEX

    International Nuclear Information System (INIS)

    Patzer, G.; Geier, F.

    1988-01-01

    The fuel assembly NY-205 has been irradiated during the first and the second core of KNK II with a total residence time of 832 equivalent full-power days. A maximum burnup of 175.000 MWd/tHM or 18.6 % was reached with a maximum steel damage of 66 dpa-NRT. For the cladding the materials 1.4970 and 1.4981 have been used in different metallurgical conditions, and for the Uranium/Plutonium mixed- oxide fuel the most important variants of the major fabrication parameters had been realized. The assembly will be brought to the Hot Cells of the KfK Karlsruhe for post-irradiation examination in February 1988, so that the knowledge of the fabrication data is of interest for the selection of fuel pins and for the evaluation of the examination results. Therefore this report compiles the fuel and fuel pin fabrication data from the INTERATOM data bank system BESEX and additionally, an overview of the irradiation history of the assembly is given [de

  16. Status of LEU fuel development and conversion of NRU

    International Nuclear Information System (INIS)

    Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

    1989-11-01

    The status of the low-enrichment uranium (LEU) fuel development and NRU conversion program at Chalk River Nuclear Laboratories is reviewed. Construction of a new fuel fabrication facility is essentially completed and installation of LEW fuel manufacturing equipment has begun. The irradiation of 31 prototype Al-61 wt% U 3 Si dispersion fuel rods, approximately one third of a full NRU core, is continuing without incident. Recent post-irradiation examination of spent fuel rods revealed that the prototype LEU fuel achieved the design burnup (80 at%) in excellent condition, confirming that the Al-U 3 Si 2 dispersion fuel to complement out Al-U 3 Si capability. Three full-size NRU rods containing Al-U 3 Si 2 dispersion fuel have been fabricated for a qualification irradiation in NRU. Post-irradiation examinations of mini-elements containing Al-U 3 Si 2 fuel revealed that the U 3 Si 2 behaved similarly to U 3 Si 2 fuel revealed that the U 3 Si 2 particles and the aluminum matrix, and fission gas bubbles up to 10 μm in diameter, could be seen in the particles after 60 at% and 80 at% burnup. The mini-elements contained a variety of silicide particle sizes; however, no significant swelling dependence on particle size distribution was observed

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

    International Nuclear Information System (INIS)

    Cho, S. W.; Park, J. S.; Kwon, S.J.; Lee, K. W.; Kim, I. J.; Yu, C. H.

    2003-09-01

    The design and construction of the fuel technology development facility was aimed to accommodate general nuclear fuel research and development for the HANARO fuel fabrication and advanced fuel researches. 1. Building size and room function 1) Building total area : approx. 3,618m 2 , basement 1st floor, ground 3th floor 2) Room function : basement floor(machine room, electrical room, radioactive waste tank room), 1st floor(research reactor fuel fabrication facility, pyroprocess lab., metal fuel lab., nondestructive lab., pellet processing lab., access control room, sintering lab., etc), 2nd floor(thermal properties measurement lab., pellet characterization lab., powder analysis lab., microstructure analysis lab., etc), 3rd floor(AHU and ACU Room) 2. Special facility equipment 1) Environmental pollution protection equipment : ACU(2sets), 2) Emergency operating system : diesel generator(1set), 3) Nuclear material handle, storage and transport system : overhead crane(3sets), monorail hoist(1set), jib crane(2sets), tank(1set) 4) Air conditioning unit facility : AHU(3sets), packaged air conditioning unit(5sets), 5) Automatic control system and fire protection system : central control equipment(1set), lon device(1set), fire hose cabinet(3sets), fire pump(3sets) etc

  18. Development of TIG Welding System for a Nuclear Fuel Test Rig

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Changyoung; Ahn, Sungho; Hong, Jintae; Kim, Kahye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers.

  19. Development of TIG Welding System for a Nuclear Fuel Test Rig

    International Nuclear Information System (INIS)

    Joung, Changyoung; Ahn, Sungho; Hong, Jintae; Kim, Kahye

    2013-01-01

    The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers

  20. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1977-01-01

    As regards an application for permission of an fabricating business of nuclear fuel materials, it should describe the site of the fabricating facilities and the structure and equipments of buildings (fire-resistant, aseismatic, waterproof, ventilating and air-tight structures), etc. The business plan to be attached to the foregoing application should contain 1) scheduled date when the fabricating business starts, 2) scheduled amounts of products classified by the kinds in each business year within 5 years since the business starts, 3) the amount and the procurement plan of funds necessary for the operation, etc. For the permission of change of a fabricating business, an application must be filed. One who wants to obtain the permission of design and construction of fabricating facilities must file an application. One who wants to undergo inspection of the construction of fabricating facilities must file an application in which various items must be written. After such inspection has been done and it is regarded as passable, a certificate of passing inspection will be given. (Rikitake, Y.)

  1. Fuel canister and blockage pin fabrication for SLSF Experiment P4

    International Nuclear Information System (INIS)

    Rhude, H.V.; Folkrod, J.R.; Noland, R.A.; Schaus, P.S.; Benecke, M.W.; Delucchi, T.A.

    1983-01-01

    As part of its fast breeder reactor safety research program, Argonne National Laboratory (ANL) has conducted an experiment (SLSF Experiment P4) to determine the extent of fuel-failure propagation resulting from the release of molten fuel from one or more heat-generating fuel canisters. The test conditions consisted of 37 full-length FTR fuel pins operating at FTR rated core nominal peak fuel/reduced coolant conditions. Thirty-four of the the fuel pins were prototypical FTR mixed-oxide fuel pins. The other three fuel pins were fabricated with a mid-core section having an enlarged canister containing fully enriched UO 2 . Two of the canisters were cylindrical and one was fluted. The cylindrical canisters were designed to fail and release molten fuel into the 37-pin fuel cluster at near full power

  2. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye

    2013-01-01

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

  3. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

  4. Binder Jetting: A Novel Solid Oxide Fuel-Cell Fabrication Process and Evaluation

    Science.gov (United States)

    Manogharan, Guha; Kioko, Meshack; Linkous, Clovis

    2015-03-01

    With an ever-growing concern to find a more efficient and less polluting means of producing electricity, fuel cells have constantly been of great interest. Fuel cells electrochemically convert chemical energy directly into electricity and heat without resorting to combustion/mechanical cycling. This article studies the solid oxide fuel cell (SOFC), which is a high-temperature (100°C to 1000°C) ceramic cell made from all solid-state components and can operate under a wide range of fuel sources such as hydrogen, methanol, gasoline, diesel, and gasified coal. Traditionally, SOFCs are fabricated using processes such as tape casting, calendaring, extrusion, and warm pressing for substrate support, followed by screen printing, slurry coating, spray techniques, vapor deposition, and sputter techniques, which have limited control in substrate microstructure. In this article, the feasibility of engineering the porosity and configuration of an SOFC via an additive manufacturing (AM) method known as binder jet printing was explored. The anode, cathode and oxygen ion-conducting electrolyte layers were fabricated through AM sequentially as a complete fuel cell unit. The cell performance was measured in two modes: (I) as an electrolytic oxygen pump and (II) as a galvanic electricity generator using hydrogen gas as the fuel. An analysis on influence of porosity was performed through SEM studies and permeability testing. An additional study on fuel cell material composition was conducted to verify the effects of binder jetting through SEM-EDS. Electrical discharge of the AM fabricated SOFC and nonlinearity of permeability tests show that, with additional work, the porosity of the cell can be modified for optimal performance at operating flow and temperature conditions.

  5. Development and fabrication of a new concept planar-tubular solid oxide fuel cell (PT-SOFC)

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Chen, F. [CAS Key Laboratory of Materials for Energy Conversion, Department of Materials Science and Engineering, University of Science and Technology of China, Hefei, 230026 Anhui (China); Department of Mechanical Engineering, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States); Ding, D. [School of Materials Science and Engineering, Georgia Institute of Technology, Atlanta, GA 30332 (United States); Gao, J. [CAS Key Laboratory of Materials for Energy Conversion, Department of Materials Science and Engineering, University of Science and Technology of China, Hefei, 230026 Anhui (China)

    2011-06-15

    The paper reports a new concept of planar-tubular solid oxide fuel cell (PT-SOFC). Emphasis is on the fabrication of the required complex configuration of Ni-yttria-stabilised zirconia (YSZ) porous anode support by tert-butyl alcohol (TBA) based gelcasting, particularly the effects of solid loading, amounts of monomers and dispersant on the rheological behaviour of suspension, the shrinkage of a wet gelcast green body upon drying, and the properties of final sample after sintering at 1350 C and reduction from NiO-YSZ to Ni-YSZ. The results show that the gelcasting is a powerful method for preparation of the required complex configuration anode support. The anode support resulted from an optimised suspension with the solid loading of 25 vol% has uniform microstructure with 37% porosity, bending strength of 44 MPa and conductivity of 300 S cm{sup -} {sup 1} at 700 C, meeting the requirements for an anode support of SOFC. Based on the as-prepared anode support, PT-SOFC single cell of Ni-YSZ/YSZ/LSCF has been fabricated by slurry coating and co-sintering technique. The cell peak power density reaches 63, 106 and 141 mW cm {sup -} {sup 2} at 700, 750 and 800 C, respectively, using hydrogen as fuel and ambient air as oxidant. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  6. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost

    International Nuclear Information System (INIS)

    Choi, Hangbok; Ko, Won Il; Yang, Myung Seung

    2001-01-01

    A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (∼49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC

  7. Fabrication of mixed oxide fuel using plutonium from dismantled weapons

    International Nuclear Information System (INIS)

    Blair, H.T.; Chidester, K.; Ramsey, K.B.

    1996-01-01

    A very brief summary is presented of experimental studies performed to support the use of plutonium from dismantled weapons in fabricating mixed oxide (MOX) fuel for commercial power reactors. Thermal treatment tests were performed on plutonium dioxide powder to determine if an effective dry gallium removal process could be devised. Fabrication tests were performed to determine the effects of various processing parameters on pellet quality. Thermal tests results showed that the final gallium content is highly dependent on the treatment temperature. Fabrication tests showed that the milling process, sintering parameters, and uranium feed did effect pellet properties. 1 ref., 1 tab

  8. Development of Coated Particle Fuel Technology

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, B. G.; Kim, S. H.

    2007-06-01

    Uranium kernel fabrication technology using a wet chemical so-gel method, a key technology in the coated particle fuel area, is established up to the calcination step and the first sintering of UO2 kernel was attempted. Experiments on the parametric study of the coating process using the surrogate ZrO2 kernel give the optimum conditions for the PyC and SiC coating layer and ZrC coating conditions were obtained for the vaporization of the ZrCl4 precursor and coating condition from ZrC coating experiments using plate-type graphite substrate. In addition, by development of fuel performance analysis code a part of the code system is completed which enables the participation to the benchmark calculation and comparison in the IAEA collaborated research program. The technologies for irradiation and post irradiation examination, which are important in developing the HTGR fuel technology of its first kind in Korea was started to develop and, through a feasibility study and preliminary analysis, the technologies required to be developed are identified for further development as well as the QC-related basic technologies are reviewed, analyzed and identified for the own technology development. Development of kernel fabrication technology can be enhanced for the remaining sintering technology and completed based on the technologies developed in this phase. In the coating technology, the optimum conditions obtained using a surrogate ZrO2 kernel material can be applied for the uranium kernel coating process development. Also, after completion of the code development in the next phase, more extended participation to the international collaboration for benchmark calculation can be anticipated which will enable an improvement of the whole code system. Technology development started in this phase will be more extended and further focused on the detailed technology development to be required for the related technology establishment

  9. Nuclear fuel control in fuel fabrication plants

    International Nuclear Information System (INIS)

    Seki, Yoshitatsu

    1976-01-01

    The basic control problems of measuring uranium and of the environment inside and outside nuclear fuel fabrication plants are reviewed, excluding criticality prevention in case of submergence. The occurrence of loss scraps in fabrication and scrap-recycling, the measuring error, the uranium going cut of the system, the confirmation of the presence of lost uranium and the requirement of the measurement control for safeguard make the measurement control very complicated. The establishment of MBA (material balance area) and ICA (item control area) can make clearer the control of inventories, the control of loss scraps and the control of measuring points. Besides the above basic points, the following points are to be taken into account: 1) the method of confirmation of inventories, 2) the introduction of reliable NDT instruments for the rapid check system for enrichment and amount of uranium, 3) the introduction of real time system, and 4) the clarification of MUF analysis and its application to the reliability check of measurement control system. The environment control includes the controls of the uranium concentration in factory atmosphere, the surface contamination, the space dose rate, the uranium concentration in air and water discharged from factories, and the uranium in liquid wastes. The future problems are the practical restudy of measurement control under NPT, the definite plan of burglary protection and the realization of the disposal of solid wastes. (Iwakiri, K.)

  10. Development of coated particle fuel technology

    International Nuclear Information System (INIS)

    Cho, Moonsung; Kim, B. G.; Kim, D. J.

    2011-06-01

    Ammonia contacting method for prehardenning the surfaces of ADU liquid droplets and the ageing/washing/drying method and equipment for spherical dried-ADU particles were improved and tested with laboratory sacle. After the improvement of fabrication process, the sphericity of UO 2 kernel obtained to 1.1, and the sintered density and O/U ratio of final UO 2 kernel were above 10.60g/cm 3 . 2.01 respectively. Defects of SiC coating layer could be minimized by optimization of gas flow rate. The fracture strength of SiC layer increased from 450 MPa to 530 MPa by controlling the coating defects. An effort was made to develop the fundamental technology for the fuel element compact for use in High Temperature Gas-cooled Reactor(HTGR) through an establishment of fabrication process, required materials and process equipment as well as performing experiments to identify the basic process conditions and optimize them. Thermal load simulation and verification experiments were carried out for an assesment of the design feasibility of the irradiation rod. Out-of-pile testing of irradiation device such as measurement of pressure drop and vibration, endurance test was performed and the validity of its design was confirmed. A fuel performance analysis code, COPA has been developed to calculate the fuel temperature, the failure fractions of coated fuel particles, the release of fission products. The COPA code can be used to evaluate the performance of the high temperature reactor fuel under the reactor operation, irradiation, heating conditions. KAERI participated in the round robin test of IAEA CRP-6 program to characterize the diameter, sphericity, coating thickness, density and anisotropy of coated particles provided by Korea, USA and South Africa. QC technology was established for TRISO-coated fuel particle. A method for accurate measurement of the optical anisotropy factor for PyC layers of coated particles was developed. Technology and inspection procedures for density

  11. 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

  12. 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.

  13. A review on the status of development in thorium-based nuclear fuels

    International Nuclear Information System (INIS)

    Lee, Young Woo; Na, S. H.; Lee, Y. W.; Kim, H. S.; Kim, S. H.; Joung, C.Y.

    2000-02-01

    Thorium as an alternative nuclear energy source had been widely investigated in the 1950s-1960s because it is more abundant than uranium, but the studies of thorium nuclear fuel cycle were discontinued by political and economic reasons in the 1970s. Recently, however, renewed interest was vested in thorium-based nuclear fuel cycle because it may generate less long-lived minor actinides and has a lower radiotoxicity of high level wastes after reprocessing compared with the thorium fuel cycle. In this state-of the art report, thorium-based nuclear cycle. In this state-of the art report, thorium-based nuclear fuel cycle and fuel fabrication processes developed so far with different reactor types are reviewed and analyzed to establish basic technologies of thorium fuel fabrication which could meet our situation. (author)

  14. Thoria-based nuclear fuels thermophysical and thermodynamic properties, fabrication, reprocessing, and waste management

    CERN Document Server

    Bharadwaj, S R

    2013-01-01

    This book presents the state of the art on thermophysical and thermochemical properties, fabrication methodologies, irradiation behaviours, fuel reprocessing procedures, and aspects of waste management for oxide fuels in general and for thoria-based fuels in particular. The book covers all the essential features involved in the development of and working with nuclear technology. With the help of key databases, many of which were created by the authors, information is presented in the form of tables, figures, schematic diagrams and flow sheets, and photographs. This information will be useful for scientists and engineers working in the nuclear field, particularly for design and simulation, and for establishing the technology. One special feature is the inclusion of the latest information on thoria-based fuels, especially on the use of thorium in power generation, as it has less proliferation potential for nuclear weapons. Given its natural abundance, thorium offers a future alternative to uranium fuels in nuc...

  15. Development of advanced fuels in the Swiss Federal Institute for Reactor Research (EIR)

    International Nuclear Information System (INIS)

    Stratton, R.W.

    1984-02-01

    The work of the project Fuel Development over the three years 1981-83 is reported. In this period virtually all of the development, demonstration and preparatory work for the fabrication of mixed carbide sphere-pac fuel pins for the FFTF experiment was completed. As well as describing the background to and the progress of the work, selected details are given of some of the results achieved in all areas of activity - fuel fabrication, pin manufacture, quality assurance, pin behaviour and modelling. Names of all principle contributers to each activity are given and in addition to references the complete list of publications over the period is provided. (Auth.)

  16. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

  17. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

  18. 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

  19. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  20. Fuel reprocessing/fabrication interface

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  1. Reprocessing and fuel fabrication systems

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  2. Recent progress in the development of metallic fuel

    International Nuclear Information System (INIS)

    Seidel, B.R.; Batte, G.L.; Dodds, N.E.; Lahm, C.E.; Pahl, R.G.; Tsai, H.C.

    1990-01-01

    Tests to date demonstrate that metallic fuel for advanced liquid metal reactors performs well, is easily reprocessed and refabricated and provides inherent reactor safety within an economic design. The behavior and performance of metallic fuel is key to the demonstration of the Integral Fast Reactor (IFR) concept at Argonne National Laboratory. Since 1985, more than 40 assemblies of experimental fuel in addition to the standard metallic driver fuel for Experimental Breeder Reactor 2 (EBR-2)have been irradiated; several more continue to be designed and fabricated. Results have characterized the influence of a wide range of fabrication, design and material variables upon irradiation behavior throughout the fuel lifetime under normal and upset conditions including operation with breached cladding. Results of test, both in- and out-of-reactor, indicate that metallic fuel is readily and economically fabricated, capable of achieving high exposure and long reactor residence times, and possesses unique and promising safety features. 9 refs., 6 figs

  3. Deep-Burn Modular Helium Reactor Fuel Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    McEachern, D

    2002-12-02

    This document contains the workscope, schedule and cost for the technology development tasks needed to satisfy the fuel and fission product transport Design Data Needs (DDNs) for the Gas Turbine-Modular Helium Reactor (GT-MHR), operating in its role of transmuting transuranic (TRU) nuclides in spent fuel discharged from commercial light-water reactors (LWRs). In its application for transmutation, the GT-MHR is referred to as the Deep-Burn MHR (DB-MHR). This Fuel Development Plan (FDP) describes part of the overall program being undertaken by the U.S. Department of Energy (DOE), utilities, and industry to evaluate the use of the GT-MHR to transmute transuranic nuclides from spent nuclear fuel. The Fuel Development Plan (FDP) includes the work on fuel necessary to support the design and licensing of the DB-MHR. The FDP is organized into ten sections. Section 1 provides a summary of the most important features of the plan, including cost and schedule information. Section 2 describes the DB-MHR concept, the features of its fuel and the plan to develop coated particle fuel for transmutation. Section 3 describes the knowledge base for fabrication of coated particles, the experience with irradiation performance of coated particle fuels, the database for fission product transport in HTGR cores, and describes test data and calculations for the performance of coated particle fuel while in a repository. Section 4 presents the fuel performance requirements in terms of as-manufactured quality and performance of the fuel coatings under irradiation and accident conditions. These requirements are provisional because the design of the DB-MHR is in an early stage. However, the requirements are presented in this preliminary form to guide the initial work on the fuel development. Section 4 also presents limits on the irradiation conditions to which the coated particle fuel can be subjected for the core design. These limits are based on past irradiation experience. Section 5 describes

  4. Plutonium fuel program

    International Nuclear Information System (INIS)

    1979-09-01

    A review is presented of the development of the (UPu)C sphere-pac fuel project during 1978. In particular, the problems encountered in obtaining good fuel quality in the fabrication process and their solution is discussed. The development of a fabrication pilot plant is considered, and the post-irradiation examination of fuel pins is presented. (Auth.)

  5. Endplug Welding Techniques developed for SFR Metallic Fuel Elements

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established.

  6. Endplug Welding Techniques developed for SFR Metallic Fuel Elements

    International Nuclear Information System (INIS)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan

    2013-01-01

    In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established

  7. Development of very high-density low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

    1997-02-01

    The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

  8. Research on plant of metal fuel fabrication using casting process (2)

    International Nuclear Information System (INIS)

    Senda, Yasuhide; Yamada, Seiya

    2005-02-01

    In this research work for the metal fuel fabrication system (38 tHM/y), the studies of the concept of the main process equipments were performed based on the previous studies on the process design and the quality control system design. In this study the handling equipment of the products were also designed, according to these designs the handling periods were evaluated. Consequently the numbers of the equipments were assessed taking into account for the method of the blending the fuel composition. (1) Structural concept design of the major equipments, the fuel handling machine and the gravimetries in the main fabrication process. The structural concept were designed for the fuel composition blending equipment, the fuel pin assembling equipment, the sodium bonding equipment, the handling equipment for fuel slug palettes, the handling equipment for fuel pins and the gravimetries. (2) Re-assessment of the numbers of the equipments taking account of the handling periods. Based on the results of item (1) the periods were evaluated for the fuel slug and pin handling. Processing time of demolder is short, then the number of it is increased to two. Three vehicles are also added to transfer the slugs and a heel smoothly. (3) Design of the buffer storages. The buffer storages among the equipments were designed through the comparison of the process speed between the equipments taking into account for the handling periods. The required amount of the structural parts (for example cladding materials) was assessed for the buffer in the same manner and the amount of the buffer facilities were optimized. (author)

  9. Fabrication of fully ceramic microencapsulated fuel by hot pressing

    International Nuclear Information System (INIS)

    Lee, H. G.; Kim, D. J; Park, J. Y.; Kim, W. J.; Lee, S. J

    2014-01-01

    Fully ceramic microencapsulated(FCM) nuclear fuel is one of the recently suggested concept to enhance stability nuclear fuel itself. The requirements to increase the accident tolerance of nuclear fuel are mainly two parts: First, the performance has to be maintained compared to the existing UO 2 nuclear fuel and zircaloy cladding system under the normal operation condition. Second, under the severe accident condition, the high temperature structural integrity has to be kept and the generation rate of hydrogen has to be decrease largely. FCM nuclear fuel consists of tristructural isotropic(TRISO) fuel particle and SiC matrix. The relative thermal conductivity of the SiC matrix as compared to UO 2 is quite good, yielding as-irradiated fuel centerline temperature compared to high temperature for the existing fuel leading to reduced stored energy in the core and reduced operational release of fission products from the fuel. Generally SiC ceramics are fabricated via liquid phase sintering due to strong covalent bonding property and low self-diffusivity coefficient. Hot pressing is very effective method to conduct sintering of SiC powder including different second phase. In this study, SiC-matrix composite including TRISO particles were sintered by hot pressing with Al 2 O 3 -Y 2 O 3 additive system. Various sintering condition were investigated to obtain high relative density above 95%. The internal distribution of TRISO particles within SiC-matrix composite was observed by x-ray radiograph. From the analysis of the cross-section of SiC-matrix composite, the fracture of TRISO particles was investigated. In order to uniform distribution of TRISO particle embedded in the SiC matrix, SiC powder overcoating is considered. SiC matrix composite including TRISO was fabricated by hot pressing. FCM pallets with full density were obtained with Al 2 O 3 -Y 2 O 3 additive system. From the microstructure image, the effect of the sintering additive contents and sintering mechanism

  10. Atomics international fuel fabrication facility and low enrichment program [contributed by H.W. Hassel, NUKEM

    International Nuclear Information System (INIS)

    Hassel, H.W.

    1993-01-01

    NUKEM handles around almost two tons of highly enriched uranium a year and it was necessary to satisfy all the new physical protection philosophies. That means that we have to install storage and safe fabrication sites for a lot of money, 25 meter thick concrete walls, and different alarm systems. So just to demonstrate how silly this business is, we have just overcome this for highly enriched uranium, and now we speak about low enriched uranium for which we don't need all of these investments to make this business safe. I would-just like to concentrate my words on the status of fabrication and considerations in my company concerning the medium enriched uranium and low enriched uranium. In the table are the different fuel types (see column) and then we have the fabrication in column 2 the experience of my comp any in kg. In column 3 is the irradiation experience of these fuels types. Column 4 shows the studies and calculations made in our company for lower and medium enriched fuels. The preliminary fabrication tests and calculations are in column 5, and in column 6 we have the delivery time for a prototype core in months after UF 6 supply. Column 7 shows the time for the development of specifications including irradiation time in years for 6 and 7 and column is the estimated cost of 6 and 7 There is just one fuel that is not in this summary and that is U-Zr. We now see how complex and sophisticated this business is. I have told you already that we have installed for a lot of millions of Deutsche Mark the physical protection, storage vaults and things like that. Now we have to investigate all these different types of fuels for, as you see, a lot of money. Maybe these are a lot of optimistic figures; anyway the question is, does this make all the overall nuclear situation worldwide easier or not. One cannot answer for the moment, but anyway we have a lot of problems

  11. A development of solid oxide fuel cell technology

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Hee Chun; Lee, Chang Woo [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Kim, Kwy Youl; Yoon, Moon Soo; Kim, Ho Ki; Kim, Young Sik; Mun, Sung In; Eom, Sung Wuk [Korea Electrotechnology Research Inst., Changwon (Korea, Republic of)

    1996-12-31

    Solid oxide fuel cell which was consisted of ceramics has high power density and is very simple in shape. The project named A development of SOFC(Solid Oxide Fuel Cell) technology is to develop the unit cell fabrication processing and to evaluate the unit cell of solid oxide full cell. In this project, a manufacturing process of cathode by citrate method and polymeric precursor methods were established. By using tape casting method, high density thin electrolyte was manufactured and has high performance. Unit cell composed with La{sub 17}Sr{sub 13}Mn{sub 3} as cathode, 8YSZ electrolyte and 50% NiYSZ anode had a performance of O.85 W/cm{sup 2} and recorded 510 hours operation time. On the basis of these results. 100 cm{sup 2} class unit cell will be fabricated and tests in next program (author). 59 refs., 120 figs.

  12. A development of solid oxide fuel cell technology

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Hee Chun; Lee, Chang Woo [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Kim, Kwy Youl; Yoon, Moon Soo; Kim, Ho Ki; Kim, Young Sik; Mun, Sung In; Eom, Sung Wuk [Korea Electrotechnology Research Inst., Changwon (Korea, Republic of)

    1995-12-31

    Solid oxide fuel cell which was consisted of ceramics has high power density and is very simple in shape. The project named A development of SOFC(Solid Oxide Fuel Cell) technology is to develop the unit cell fabrication processing and to evaluate the unit cell of solid oxide full cell. In this project, a manufacturing process of cathode by citrate method and polymeric precursor methods were established. By using tape casting method, high density thin electrolyte was manufactured and has high performance. Unit cell composed with La{sub 17}Sr{sub 13}Mn{sub 3} as cathode, 8YSZ electrolyte and 50% NiYSZ anode had a performance of O.85 W/cm{sup 2} and recorded 510 hours operation time. On the basis of these results. 100 cm{sup 2} class unit cell will be fabricated and tests in next program (author). 59 refs., 120 figs.

  13. Development of alkaline fuel cells.

    Energy Technology Data Exchange (ETDEWEB)

    Hibbs, Michael R.; Jenkins, Janelle E.; Alam, Todd Michael; Janarthanan, Rajeswari; Horan, James L.; Caire, Benjamin R.; Ziegler, Zachary C.; Herring, Andrew M.; Yang, Yuan; Zuo, Xiaobing; Robson, Michael H.; Artyushkova, Kateryna; Patterson, Wendy; Atanassov, Plamen Borissov

    2013-09-01

    This project focuses on the development and demonstration of anion exchange membrane (AEM) fuel cells for portable power applications. Novel polymeric anion exchange membranes and ionomers with high chemical stabilities were prepared characterized by researchers at Sandia National Laboratories. Durable, non-precious metal catalysts were prepared by Dr. Plamen Atanassovs research group at the University of New Mexico by utilizing an aerosol-based process to prepare templated nano-structures. Dr. Andy Herrings group at the Colorado School of Mines combined all of these materials to fabricate and test membrane electrode assemblies for single cell testing in a methanol-fueled alkaline system. The highest power density achieved in this study was 54 mW/cm2 which was 90% of the project target and the highest reported power density for a direct methanol alkaline fuel cell.

  14. A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

    International Nuclear Information System (INIS)

    Vasudevamurthy, G.; Radecka, A.; Massey, C.

    2015-01-01

    High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

  15. A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vasudevamurthy, G.; Radecka, A.; Massey, C. [Virginia Commonwealth Univ., Richmond, VA (United States). High Temperature Materials Lab.

    2015-07-01

    High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

  16. Waste management in MOX fuel fabrication plants

    International Nuclear Information System (INIS)

    Schneider, V.

    1982-01-01

    After a short description of a MOX fuel fabrication plant's activities the waste arisings in such a plant are discussed according to nature, composition, Pu-content. Experience has shown that proper recording leads to a reduction of waste arisings by waste awareness. Aspects of the treatment of α-waste are given and a number of treatment processes are reviewed. Finally, the current waste management practice and the α-waste treatment facility under construction at ALKEM are outlined. (orig./RW)

  17. AECL's progress in developing the DUPIC fuel fabrication process

    International Nuclear Information System (INIS)

    Sullivan, J.D.; Cox, D.S.

    1995-01-01

    Spent Pressurized Water Reactor (PWR) fuel can be used directly in CANDU reactors without the need for wet chemical reprocessing or reenrichment. Considerable experimental progress has been made in verifying the practicality of this fuel cycle, including hot-cell experiments using spent PWR fuels and out-cell trials using surrogate fuels. This paper describes the current status of these experiments. (author)

  18. Informal presentations by fuel fabricators and others [contributed by W. Ross, U.S. NRC

    International Nuclear Information System (INIS)

    Ross, W.

    1993-01-01

    Al-28 wt.% U alloy and its ductility certainly will not increase with irradiation. The effect of this brittleness on behavior of the long NRX/NRU rods during thermal cycling in reactor would have to be investigated. Therefore from a fabrication point of view, it may be possible to make the rods from Al-40 wt.% U but it is obvious that an extensive fabrication, safety, and irradiation study would be required before a definitive answer could be given. Development of Al-50 wt.% U alloy for such rods would be even more difficult, the probability of success smaller, and the development program somewhat larger. The development of a completely new design driver fuel for the reactors using Zircaloy clad powder packed UO 2 or dispersion type fuels would require an even larger and more expensive program. At 20% enrichment level, the current Al-U designs of driver fuel could not be used even by increasing the number of fuel rods and/or by removing experimental facilities. A new fuel design would have to be tested to burnups of 60-70% for NRX and over 80% for NRU before it could be considered to be acceptable. The defect performance of the higher uranium alloys would also have to be checked out. In summary, the NRX and NRU are high performance, high flux research reactors having a very heavy experimental and radioisotope load. With the present designs of driver fuel and reactor loadings the use of 50% enriched uranium would be possible only if the brittle Al-40 to 50 wt.% U alloys can be successfully developed for high burnups. Extensive fabrication, safety, and physics work and irradiation of intact and defective elements would have to be done for each of the fuel designs considered before a definitive answer could be given. If these alloys could not be successfully developed, new designs of driver fuel would be required. The fuel developments would be in three phases: fuel alloy development, pin development, and irradiation testing and would probably take 3 to years. A change

  19. Minimization of waste from uranium purification, enrichment and fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-10-01

    As any industry, nuclear industry generates a diverse range of waste which has to be managed in a safe manner to be acceptable to the public and the environment. The cost of waste management, the risks to the public and employees, and the detriment to the environment are dependent on the quantity and radioactive content of the waste generated. Waste minimization is a necessary activity needed to reduce the impact from nuclear fuel cycle operations and it is included in the national policy of some countries. In recognition of the importance of the subject, the IAEA has decided to review the current status of the work aimed at waste minimization in the nuclear fuel cycle. The waste minimization issues related to the back end of the nuclear fuel cycle are covered in Technical Reports Series No. 377 'Minimization of Radioactive Waste from Nuclear Power Plants and the Back End of the Nuclear Fuel Cycle' published in 1995. The present report deals with the front end of the nuclear fuel cycle, including existing options, approaches, developments and some specific considerations to be taken into account in decision making on waste minimization. It has been recognized that, in comparison with the back end of the nuclear fuel cycle, much less information is available, and this report should be considered as a first attempt to analyse waste minimization practices and opportunities in uranium purification, conversion, enrichment and fuel fabrication. Although mining and milling is an important part of the front end of the nuclear fuel cycle, these activities are excluded from consideration since relevant activities are covered in other IAEA publications.

  20. Minimization of waste from uranium purification, enrichment and fuel fabrication

    International Nuclear Information System (INIS)

    1999-10-01

    As any industry, nuclear industry generates a diverse range of waste which has to be managed in a safe manner to be acceptable to the public and the environment. The cost of waste management, the risks to the public and employees, and the detriment to the environment are dependent on the quantity and radioactive content of the waste generated. Waste minimization is a necessary activity needed to reduce the impact from nuclear fuel cycle operations and it is included in the national policy of some countries. In recognition of the importance of the subject, the IAEA has decided to review the current status of the work aimed at waste minimization in the nuclear fuel cycle. The waste minimization issues related to the back end of the nuclear fuel cycle are covered in Technical Reports Series No. 377 'Minimization of Radioactive Waste from Nuclear Power Plants and the Back End of the Nuclear Fuel Cycle' published in 1995. The present report deals with the front end of the nuclear fuel cycle, including existing options, approaches, developments and some specific considerations to be taken into account in decision making on waste minimization. It has been recognized that, in comparison with the back end of the nuclear fuel cycle, much less information is available, and this report should be considered as a first attempt to analyse waste minimization practices and opportunities in uranium purification, conversion, enrichment and fuel fabrication. Although mining and milling is an important part of the front end of the nuclear fuel cycle, these activities are excluded from consideration since relevant activities are covered in other IAEA publications

  1. Introduction to Exxon nuclear fuel fabrication plant

    International Nuclear Information System (INIS)

    Schneider, R.A.

    1985-01-01

    The Exxon Nuclear low-enriched uranium fuel fabrication plant in Richland, Washington produces fuel assemblies for both pressurized water and boiling water reactors. The Richland plant was the first US bulk-handling facility selected by the IAEA for inspection under the US-IAEA Safeguards Agreement. The plant was under IAEA inspection from March 1981 through October 1983. This text provides a written description of the plant layout, operation and process. The text also includes a one ton-a-day model (or reference) plant which was adapted from the Exxon Nuclear plant. The Model Plant provides a generic example of a low-enriched uranium (LEU) bulk-handling facility. The Model Plant is used to illustrate in a more quantitative way some of the key safeguards requirements for a bulk-handling facility

  2. Radioactive waste management of experimental DUPIC fuel fabrication process

    International Nuclear Information System (INIS)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Hong, K. P.

    2001-01-01

    The concept of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) is a dry processing technology to manufacture CANDU compatible DUPIC fuel from spent PWR fuel material. Real spent PWR fuel was used in IMEF M6 hot cell to carry out DUPIC experiment. Afterwards, about 200 kg-U of spent PWR fuel is supposed to be used till 2006. This study has been conducted in some hot cells of PIEF and M6 cell of IMEF. There are various forms of nuclear material such as rod cut, powder, green pellet, sintered pellet, fabrication debris, fuel rod, fuel bundle, sample, and process waste produced from various manufacturing experiment of DUPIC fuel. After completing test, the above nuclear wastes and test equipment etc. will be classified as radioactive waste, transferred to storage facility and managed rigorously according to domestic and international laws until the final management policy is determined. It is desirable to review management options in advance for radioactive waste generated from manufacturing experiment of DUPIC nuclear fuel as well as residual nuclear material and dismantled equipment. This paper includes basic plan for DUPIC radwaste, arising source and estimated amount of radioactive waste, waste classification and packing, transport cask, transport procedures

  3. High U-density nuclear fuel development with application of centrifugal atomization technology

    International Nuclear Information System (INIS)

    Kim, Chang Kyu; Kim, Ki Hwan; Lee, Don Bae

    1997-01-01

    In order to simplify the preparation process and improve the properties of uranium silicide fuels prepared by mechanical comminution, a fuel fabrication process applying rotating-disk centrifugal atomization technology was invented in KAERI in 1989. The major characteristic of atomized U 3 Si and U 3 Si 2 powders have been examined. The out-pile properties, including the thermal compatibility between atomized particle and aluminum matrix in uranium silicide dispersion fuels, have generally showed a superiority to the comminuted fuels. Moreover, the RERTR (reduced enrichment for research and test reactors) program, which recently begins to develop very-high-density uranium alloy fuels, including U-Mo fuels, requires the centrifugal atomization process to overcome the contaminations of impurities and the difficulties of the comminution process. In addition, a cooperation with ANL in the U.S. has been performed to develop high-density fuels with an application of atomization technology since December 1996. If the microplate and miniplate irradiation tests of atomized fuels, which have been performed with ANL, demonstrated the stability and improvement of in-reactor behaviors, nuclear fuel fabrication technology by centrifugal atomization could be most-promising to the production method of very-high-uranium-loading fuels. (author). 22 refs., 2 tabs., 12 figs

  4. Effect of the fabrication process on fatigue performance of U3Si2 fuel plate with sandwich structure

    International Nuclear Information System (INIS)

    Wang Xishu; Li Shuangshou; Wang Qingyuan; Xu Yong

    2005-01-01

    U 3 Si 2 -Al fuel plate is one of the dispersion fuel structure materials recently developed and widely used in research reactors. The mechanical properties of this structural material, especially the fatigue performance, are strongly dependent on its fabrication process. To investigate the effects of these processing technologies, the fatigue tests for the different specimens were carried out. The S-N curves indicate that the fabrication processing technologies of U 3 Si 2 fuel plate, such as the addition of U 3 Si 2 particles into aluminum powder to form the fuel meat, holding and rolling the processes of meat and cladding of 6061-Al alloy, plays an important role in improving the mechanical properties and fatigue performance of this fuel plate. In addition, some factors that influence the crack initiation and propagation are summarized based on the fatigue images that are in situ observations with SEM. The critical criterion for fatigue damage is proposed based on the fatigue data of the structural material, which were obtained at the different conditions

  5. Development of MOX fuel database

    International Nuclear Information System (INIS)

    Ikusawa, Yoshihisa; Ozawa, Takayuki

    2007-03-01

    We developed MOX Fuel Database, which included valuable data from several irradiation tests in FUGEN and Halden reactor, for help of LWR MOX use. This database includes the data of fabrication and irradiation, and the results of post-irradiation examinations for seven fuel assemblies, i.e. P06, P2R, E03, E06, E07, E08 and E09, irradiated in FUGEN. The highest pellet peak burn-up reached ∼48GWd/t in MOX fuels, of which the maximum plutonium content was ∼6 wt%, irradiated in E09 fuel assembly without any failure. Also the data from the instrumented MOX fuels irradiated in HBWR to study the irradiation behavior of BWR MOX fuels under the steady state condition (IFA-514/565 and IFA-529), under the load-follow operation condition (IFA-554/555) and under the transit condition (IFA-591) are included in this database. The highest assembly burn-up reached ∼56 GWd/t in IFA-565 steady state irradiation test, and the maximum linear power of MOX fuel rods was 58.3-68.4 kW/m without any failure in IFA-591 ramp test. In addition, valuable instrument data, i.e. cladding elongation, fuel stack elongation, fuel center temperature and rod inner pressure were obtained from IFA-554/555 load-follow test. (author)

  6. Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

    International Nuclear Information System (INIS)

    Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed

    2013-01-01

    The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed

  7. Dimensional Behavior of Matrix Graphite Compacts during Heat Treatments for HTGR Fuel Element Fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung

    2015-01-01

    The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K. This carbonization step is followed by the final high temperature heat treatment where the carbonized compacts are heat treated at 2073-2173 K in vacuum for a relatively short time (about 2 hrs). In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions, which has a strong influence on the further steps and the material properties of fuel element. In this work, the dimensional changes of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed, keeping other process parameters constant, such as the binder content, carbonization time, temperature and atmosphere (two hours ant 1073K and N2 atmosphere). In this work, the dimensional variations of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed

  8. Fabrication of high-uranium-loaded U{sub 3}O{sub 8}-Al developmental fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Copeland, G L; Martin, M M [Oak Ridge National Laboratory, TN (United States)

    1983-08-01

    A common plate-type fuel for Research and Test Reactors (RERTR) is U{sub 3}0{sub 8} dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the {sup 235}U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for non-peaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service. We fabricated developmental fuel plates with cores containing from 60 to 100 wt U{sub 3}0{sub 8} in aluminum encapsulated in 6061 aluminum alloy and evaluated them for aspects of fabricability, nondestructive testing, and expected performance. We recommend 75 wt U{sub 3}0{sub 8}-Al 3.1 Mg U/m{sup 3}) as the highest loading in the initial irradiation test. This upper limit is based on a qualitative assessment of the mechanical integrity of the core made by using current fabrication techniques and materials. As the oxide loading is increased beyond this point, planar areas and extensive stringers of oxide and voids develop, which leave little strength in the thickness direction. Fuel plates may then blister over these areas as fission gases collect during irradiation. Current size plates are easily fabricable to the 75 wt % U{sub 3}0{sub 8}-Al core loading by current fabrication techniques. Dogboning is a potential problem at this loading for some applications; however, this can be easily solved by using tapered compact ends. Current nondestructive radiography and transmission x-ray scanning are applicable to the highly loaded plates. Ultrasonic testing for non-bonds is marginal because of the abrupt change in conductance at the cladding-core interface. Plate thickness can be increased if desired; we fabricated 75 wt % plates with

  9. The status of uranium-silicon alloy fuel development for the RERTR program

    International Nuclear Information System (INIS)

    Domagala, R.F.; Wiencek, T.C.; Thresh, H.R.; Stahl, D.

    1983-01-01

    As part of the national Reduced Enrichment Research and Test Reactor (RERTR) Program, Argonne National Laboratory (ANL) is engaged in a fuel-alloy development project. The fuel alloys are dispersed in an aluminum matrix and metallurgically roll-bonded within 6061 Al alloy. To date, 'miniplates' with up to 40 vol. fuel alloy have been successfully fabricated. Thirty-one of these plates have been or are being irradiated in the Oak Ridge Reactor (ORR). Three different fuels have been used in the ANL miniplates: U 3 Si (U + 4 wt.% Si), U 3 Si 2 (U + 7.4 wt.% Si), or ''U 3 SiAl'' (U + 3.5 wt.% Si + 1.5 wt.% Al). All three are candidates for permitting higher fuel loadings and thus lower enrichments of 235 U than would be possible with either UAl x or U 3 O 8 , the current fuels for plate-type elements. The enrichment level employed at ANL is ∼19.8%. Continuing effort involves the production of miniplates with up to ∼60 vol. % fuel, the development of a technology for full-size plate fabrication, and post-irradiation examination of miniplates already removed from the ORR. (author)

  10. Design, fabrication and installation of irradiation facilities -Advanced nuclear material development-

    International Nuclear Information System (INIS)

    Kim, Yong Seong; Lee, Jeong Yeong; Lee, Seong Ho; Ji, Dae Yeong; Kim, Seok Hoon; An, Seong Ho; Kim, Dong Hoon; Seok, Ho Cheon; Kim, Joon Yeon; Yang, Seong Hong

    1994-07-01

    The objective of this study is to design and construct the steady state fuel test loop and non-instrumented capsules to be installed in KMRR. The principle contents of this project are to design, fabricate the steady-state fuel test loop and non-instrumented capsule to be installed in KMRR for nuclear technology development. This project will be completed in 1996, so preparation of design criteria for fuel test loop have been performed in 1993 as the first year of the first phase in implementing this project. Also design and pressure drop test of non-instrumented capsule have been performed in 1993

  11. Overview of current research and development programmes for fuel in Japan

    International Nuclear Information System (INIS)

    Shiozawa, S.

    1991-01-01

    The Research and Development (R and D) programmes for HTGR fuel have been performed since 1969 by Japan Atomic Energy Research Institute (JAERI) as a leading organization in Japan. The R and D covers all fields necessary for the construction of the High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan. This R and D includes fuel fabrication, fuel property data, irradiation performance under normal operating conditions, safety-related research and fuel inspection technology. The R and D for the HTTR has been completed from a licensing point of view. Some R and D including future advanced fuel development continue. 2 figs, 3 tabs

  12. Fabrication of Metallic Fuel Slugs for Irradiation Experiments in Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arun; Prasad, G.J.

    2013-01-01

    Advantages of Metallic fuels for future FBR: → High heavy metal atom density; → Higher thermal conductivity at room temperature that increases with temperature; → Metal fuels can be relatively easily fabricated with close dimensional tolerances; → They have excellent compatibility with liquid metal coolants

  13. Development of very-high-density low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Snegrove, J.L.; Hofmann, G.L.; Trybus, C.L.; Wiencek, T.C.

    1997-01-01

    The RERTR (=Reduced Enrichment for Research and Test Reactors) program has begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun. (author)

  14. Airborne effluent control at fuel enrichment, conversion, and fabrication plants

    International Nuclear Information System (INIS)

    Mitchell, M.E.

    1976-01-01

    Uranium conversion, enrichment, and fuel fabrication facilities generate gaseous wastes that must be treated prior to being discharged to the atmosphere. Since all three process and/or handle similar compounds, they also encounter similar gaseous waste disposal problems, the majority of which are treated in a similar manner. Ventilation exhausts from personnel areas and equipment off-gases that do not contain corrosive gases (such as HF) are usually passed through roughening and/or HEPA filters prior to release. Ventilation exhausts that contain larger quantities of particles, such as the conversion facilities' U 3 O 8 sampling operation, are passed through bag filters or cyclone separators, while process off-gases containing corrosive materials are normally treated by sintered metal filters or scrubbers. The effectiveness of particle removal varies from about 90 percent for a scrubber alone to more than 99.9 percent for HEPA filters or a combination of the various filters and scrubbers. The removal of nitrogen compounds (N 2 , HNO 3 , NO/sub x/, and NH 3 ) is accomplished by scrubbers in the enrichment and fuel fabrication facilities. The conversion facility utilizes a nitric acid recovery facility for both pollution control and economic recovery of raw materials. Hydrogen removal from gaseous waste streams is generally achieved with burners. Three different systems are currently utilized by the conversion, enrichment, and fuel fabrication plants to remove gaseous fluorides from airborne effluents. The HF-rich streams, such as those emanating from the hydrofluorination and fluorine production operations of the conversion plant, are passed through condensers to recover aqueous hydrofluoric acid

  15. Estimation of radiation exposure for hot cell workers during DUPIC fuel fabrication process in IMEF M6 cell

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Yong Bum; Baek, Sang Yeol; Park, Dae Kyu

    1997-06-01

    DUPIC(Direct Use of spent PWR fuel In CANDU) fuel cycle to utilize the PWR spent fuel in fabricating CANDU fuel, which is expected to reduce not only the total amount of high level radwastes but the energy sources is underway. IMEF M6 cell to be used as DUPIC fuel fabrication facility is refurbished and retrofitted. Radiation exposure for the hot cell worker by dispersion of the radioactive materials during the DUPIC process were estimated on the basis of the hot cell design information. According to the estimation results, DUPIC fuel fabrication process could be run without any severe impacts to the hot cell workers when the ventilation system to maintain the sufficient pressure difference between hotcell and working area and radiation monitoring system is supports the hot cell operation properly. (author). 4 tabs., 6 figs.

  16. Recent advances in fuel fabrication techniques and prospects for the nineties

    International Nuclear Information System (INIS)

    Frain, R.G.; Caudill, H.L.; Faulhaber, R.

    1987-01-01

    Advanced Nuclear Fuels Corporation's approach and experience with the application of a flexible, just-in-time manufacturing philosophy to the production of customized nuclear fuel is described. Automation approaches to improve productivity are described. The transfer of technology across product lines is discussed as well as the challenges presented by a multiple product fabrication facility which produces a wide variety of BWR and PWR designs. This paper also describes the method of managing vendor quality control programs in support of standardization and clarity of documentation. Process simplification and the ensuing experience are discussed. Prospects for fabrication process advancements in the nineties are given with emphasis on the benefits of dry conversion of UF 6 to UO 2 powder, and increased use of automated and computerized inspection techniques. (author)

  17. Recent development in safety regulation of nuclear fuel cycle activities

    International Nuclear Information System (INIS)

    Kato, S.

    2001-01-01

    Through the effort of deliberation and legislation over five years, Japanese government structure was reformed this January, with the aim of realizing simple, efficient and transparent administration. Under the reform, the Agency for Nuclear and Industrial Safety (ANIS) was founded in the Ministry of Economy, Trade and Industry (METI) to be responsible for safety regulation of energy-related nuclear activities, including nuclear fuel cycle activities, and industrial activities, including explosives, high-pressure gasses and mining. As one of the lessons learned from the JCO criticality accident of September 1999, it was pointed out that the government's inspection function was not enough for fuel fabrication facilities. Accordingly, new statutory regulatory activities were introduced, namely, inspection of observance of safety rules and procedures for all kinds of nuclear operators and periodic inspection of fuel fabrication facilities. In addition, in order to cope with insufficient safety education and training of workers in nuclear facilities, licensees of nuclear facilities are required by law to specify safety education and training for their workers. ANIS is committed to enforce these new regulatory activities effectively and efficiently. In addition, it is going to be prepared, in its capacity as safety regulatory authority, for future development of Japanese fuel cycle activities, including commissioning of JNFL Rokkasho reprocessing plant and possible application for licenses for JNFL MOX fabrication plant and for spent fuel interim storage facilities. (author)

  18. Fabrication of a pressurized water reactor fuel element prototype with Zy-control rod guide tubes

    International Nuclear Information System (INIS)

    Bezold, H.; Romeiser, H.J.

    1978-10-01

    A prototype fuel assembly with zircaloy guide was fabricated by mass production methods. The fastening of the Inconel spacer grids to the guide tubes and the transition joint for fixing the tubes to the stainless stell upper end-fitting of the assembly were investigated. Tools and welding devices were developed for the construction of the skeleton. (orig.) [de

  19. Development of machine vision system for PHWR fuel pellet inspection

    Energy Technology Data Exchange (ETDEWEB)

    Kamalesh Kumar, B.; Reddy, K.S.; Lakshminarayana, A.; Sastry, V.S.; Ramana Rao, A.V. [Nuclear Fuel Complex, Hyderabad, Andhra Pradesh (India); Joshi, M.; Deshpande, P.; Navathe, C.P.; Jayaraj, R.N. [Raja Ramanna Centre for Advanced Technology, Indore, Madhya Pradesh (India)

    2008-07-01

    Nuclear Fuel Complex, a constituent of Department of Atomic Energy; India is responsible for manufacturing nuclear fuel in India . Over a million Uranium-di-oxide pellets fabricated per annum need visual inspection . In order to overcome the limitations of human based visual inspection, NFC has undertaken the development of machine vision system. The development involved designing various subsystems viz. mechanical and control subsystem for handling and rotation of fuel pellets, lighting subsystem for illumination, image acquisition system, and image processing system and integration. This paper brings out details of various subsystems and results obtained from the trials conducted. (author)

  20. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  1. Prevention of significant deterioration permit application for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    Energy Technology Data Exchange (ETDEWEB)

    1989-08-01

    This New Source Review'' has been submitted by the US Department of Energy-Richland Operations Office (PO Box 550, Richland, Washington 99352), pursuant to WAC 173-403-050 and in compliance with the Department of Ecology Guide to Processing A Prevention Of Significant Deterioration (PSD) Permit'' for three new sources of radionuclide emissions at the Hanford Site in Washington State. The three new sources, the Fueled Clad Fabrication System (FCFS), the Radioisotope Power Systems Facility (RPSF), and the Fuel Assembly Area (FAA), will be located in one facility, the Fuels and Materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post-irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies for use in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, these systems will be dealt with separately to the extent possible. The FAA is a comparatively independent operation though it will share the FMEF complex.

  2. Development of very-high-density fuels by the RERTR program

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

    1996-01-01

    The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

  3. Low-enriched uranium-molybdenum fuel plate development

    International Nuclear Information System (INIS)

    Wiencek, T.C.; Prokofiev, I.G.

    2000-01-01

    To examine the fabricability of low-enriched uranium-molybdenum powders, full-size 450 x 60 x 0.5-mm (17.7 x 2.4 x 0.020-in.) fuel zone test plates loaded to 6 g U/cm 3 were produced. U-10 wt.% Mo powders produced by two methods, centrifugal atomization and grinding, were tested. These powders were supplied at no cost to Argonne National Laboratory by the Korean Atomic Energy Research Institute and Atomic Energy of Canada Limited, respectively. Fuel homogeneity indicated that both of the powders produced acceptable fuel plates. Operator skill during loading of the powder into the compacting die and fuel powder morphology were found to be important when striving to achieve homogeneous fuel distribution. Smaller, 94 x 22 x 0.6-mm (3.7 x 0.87 x 0.025-in.) fuel zone, test plates were fabricated using U-10 wt.% Mo foil disks instead of a conventional powder metallurgy compact. Two fuel plates of this type are currently undergoing irradiation in the RERTR-4 high-density fuel experiment in the Advanced Test Reactor. (author)

  4. Demonstration of Subscale Cermet Fuel Specimen Fabrication Approach Using Spark Plasma Sintering and Diffusion Bonding

    Science.gov (United States)

    Barnes, Marvin W.; Tucker, Dennis S.; Benensky, Kelsa M.

    2018-01-01

    Nuclear thermal propulsion (NTP) has the potential to expand the limits of human space exploration by enabling crewed missions to Mars and beyond. The viability of NTP hinges on the development of a robust nuclear fuel material that can perform in the harsh operating environment (> or = 2500K, reactive hydrogen) of a nuclear thermal rocket (NTR) engine. Efforts are ongoing to develop fuel material and to assemble fuel elements that will be stable during the service life of an NTR. Ceramic-metal (cermet) fuels are being actively pursued by NASA Marshall Space Flight Center (MSFC) due to their demonstrated high-temperature stability and hydrogen compatibility. Building on past cermet fuel development research, experiments were conducted to investigate a modern fabrication approach for cermet fuel elements. The experiments used consolidated tungsten (W)-60vol%zirconia (ZrO2) compacts that were formed via spark plasma sintering (SPS). The consolidated compacts were stacked and diffusion bonded to assess the integrity of the bond lines and internal cooling channel cladding. The assessment included hot hydrogen testing of the manufactured surrogate fuel and pure W for 45 minutes at 2500 K in the compact fuel element environmental test (CFEET) system. Performance of bonded W-ZrO2 rods was compared to bonded pure W rods to access bond line integrity and composite stability. Bonded surrogate fuels retained structural integrity throughout testing and incurred minimal mass loss.

  5. Development and fabrication of superconducting hybrid Cable-In-Conduit-Conductor (CICC) for indigenous fusion programme

    International Nuclear Information System (INIS)

    Singh, A.K.; Hussain, M.M.; Abdulla, K.K.; Singh, R.P.

    2011-01-01

    The Atomic Fuels Division has initiated development and fabrication of Cable-In-Conduit-Conductor (CICC) of various configurations, for superconducting fusion grade magnets required for their indigenous Fusion Programme. The process involves development of high grade superconducting multifilamentary wire, multi stage cabling of superconducting as well as copper wires and, finally, jacketing of the cables in SS316LN tubes. The overview of the development and fabrication of CICC is presented in this article. (author)

  6. Glucose Fuel Cells with a MicroChannel Fabricated on Flexible Polyimide Film

    Science.gov (United States)

    Sano, Ryohei; Fukushi, Yudai; Sasaki, Tsubasa; Mogi, Hiroshi; Koide, Syohei; Ikoma, Ryuta; Akatsuka, Wataru; Tsujimura, Seiya; Nishioka, Yasushiro

    2013-12-01

    In this work, a glucose fuel cell was fabricated using microfabrication processes assigned for microelectromechanical systems. The fuel cell was equipped with a microchannel to flow an aqueous solution of glucose. The cell was fabricated on a flexible polyimide substrate, and its porous carbon-coated aluminum (Al) electrodes of 2.8 mm in width and 11 mm in length were formed using photolithography and screen printing techniques. Porous carbon was deposited by screen printing of carbon black ink on the Al electrode surfaces in order to increase the effective electrode surface area and to absorb more enzymes on the electrode surfaces. The microchannel with a depth of 200 μm was fabricated using a hot embossing technique. A maximum power of 0.45 μW at 0.5 V that corresponds to a power density of 1.45 μW/cm2 was realized by introducing a 200 mM concentrated glucose solution at room temperature.

  7. Glucose Fuel Cells with a MicroChannel Fabricated on Flexible Polyimide Film

    International Nuclear Information System (INIS)

    Sano, Ryohei; Fukushi, Yudai; Sasaki, Tsubasa; Mogi, Hiroshi; Koide, Syohei; Ikoma, Ryuta; Nishioka, Yasushiro; Akatsuka, Wataru; Tsujimura, Seiya

    2013-01-01

    In this work, a glucose fuel cell was fabricated using microfabrication processes assigned for microelectromechanical systems. The fuel cell was equipped with a microchannel to flow an aqueous solution of glucose. The cell was fabricated on a flexible polyimide substrate, and its porous carbon-coated aluminum (Al) electrodes of 2.8 mm in width and 11 mm in length were formed using photolithography and screen printing techniques. Porous carbon was deposited by screen printing of carbon black ink on the Al electrode surfaces in order to increase the effective electrode surface area and to absorb more enzymes on the electrode surfaces. The microchannel with a depth of 200 μm was fabricated using a hot embossing technique. A maximum power of 0.45 μW at 0.5 V that corresponds to a power density of 1.45 μW/cm 2 was realized by introducing a 200 mM concentrated glucose solution at room temperature

  8. Improvements in the consistency of fabrication and the reliability of nuclear fuels through quality assurance

    International Nuclear Information System (INIS)

    Sifferlen, R.

    1976-01-01

    By establishing correlations between rejection level and fabrication parameters, quality assurance can guide corrective action for improving the consistency of fabrication and the reliability of fuel elements. The author cites examples relating to the quality of the uranium in metallic fuels, the influence of the parent metal on the welding of zirconium alloys, the behaviour of the springs inside the cladding during the welding of plugs and the behaviour of uranium oxide pellets. (author)

  9. Supplemental information for a notice of construction for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    Energy Technology Data Exchange (ETDEWEB)

    1989-08-01

    This ''Notice of Construction'' has been submitted by the US Department of Energy-Richland Operations Office (P.O. Box 550, Richland, Washington 99352), pursuant to WAC 402-80-070, for three new sources of radionuclide emissions at the Hanford Site in Washington State (Figure 1). The three new sources, the Fueled Clad Fabrication System (FCFS) the Radioisotope Power Systems Facility (RPSF) and the Fuel Assembly Area (FAA) will be located in one facility, the Fuels and materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post- irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies to be used in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, to the extent possible, these systems will be dealt with separately. The FAA is a comparatively independent operation though it will share the FMEF complex.

  10. The miscibility and oxidation study of the simulated metallic spent fuel for the development of an advanced spent fuel management process

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Y. J.; You, G. S.; Ju, J. S.; Lee, E. P.; Seo, H. S.; Ahn, S. B. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    1999-03-01

    The simulated metallic spent fuel ingots were fabricated and evaluated the oxidation rates and the activation energies under several temperature conditions to develop an advanced spent fuel management process. It was also checked the immiscibility of the some elements with metal uranium. 2 refs., 45 figs. (Author)

  11. Concept and development status of fast breeder reactor fuels in the FaCT project

    International Nuclear Information System (INIS)

    Maeda, S.; Suzuki, M.; Kaito, T.; Tanaka, K.; Abe, T.

    2013-01-01

    The fuel development and the conceptual design study have been progressed in the first phase of the FaCT project in Japan. Significant outcomes of key technologies related to fuel design, fuel properties, core materials, fuel fabrication have been provided. The prospects of these technologies have been identified. After the Fukushima accident, the research and development for reducing the amount and toxic level of radioactive wastes will be promoted more than before. These outcomes will be reflected on the future development

  12. Method of fabricating zirconium metal for use in composite type fuel cans

    International Nuclear Information System (INIS)

    Imahashi, Hiromichi; Inagaki, Masatoshi; Akabori, Kimihiko; Tada, Naofumi; Yasuda, Tetsuro.

    1985-01-01

    Purpose: To mass produce zirconium metal for fuel cans with less radiation hardening. Method: Zirconium sponges as raw material are inserted in a hearth mold and a procedure of melting the zirconium sponges portionwise by using a melting furnace having electron beams as a heat source while moving the hearth is repeated at least for once. Then, the rod-like ingot after melting is melted again in a vacuum or inert gas atmosphere into an ingot of a low oxygen density capable of fabrication. A composite fuel can billet is formed by using the thus obtained zirconium ingot and a zircalloy, and a predetermined composite type fuel can is manufactured by way of hot extrusion and pipe drawing fabrication. The raw material usable herein is zirconium sponge with an oxygen density of 400 ppm or higher and the content of impurity other than oxygen is between 1000 - 5000 ppm in total, or the molten material thereof. (Kamimura, M.)

  13. Basic tendencies of restructured UO2 nuclear fuels fabrication industry for water-moderated reactors

    International Nuclear Information System (INIS)

    Makhova, V.A.; Bokshitskij, V.I.; Blinova, I.V.

    2002-01-01

    Processes of reformation and consolidation of firms and frontier nuclear fuels fabrication industry associated with processes of globalization and deregulation of electric power market are analyzed. Current state of nuclear fuel market and basic factors influenced on the market are presented. The role of nuclear fuel in increasing competition of NPP and fundamental directions of innovation action on the creation of perspective kinds of fuel were considered [ru

  14. Main trends and content of works on fabrication of fuel rods with MOX fuel for the WWER-1000 reactor

    International Nuclear Information System (INIS)

    Tsykanov, V.A.; Golovanov, V.N.; Mayorshin, A.A.; Yurchenko, A.D.; Ilyenko, S.A.; Syuzev, V.N.

    2000-01-01

    The main trends of production of pellet MOX-fuel for the WWER reactors using the trial-experimental equipment at SSC RF RIAR are set forth. The main realized parameters of fabrication of MOX-fuel pellets are presented. The content of the reactor tests program is considered with allowance for their licensing requirements for the WWER reactors. (author)

  15. Fabrication of small-orifice fuel injectors for diesel engines.

    Energy Technology Data Exchange (ETDEWEB)

    Woodford, J. B.; Fenske, G. R.

    2005-04-08

    Diesel fuel injector nozzles with spray hole diameters of 50-75 {micro}m have been fabricated via electroless nickel plating of conventionally made nozzles. Thick layers of nickel are deposited onto the orifice interior surfaces, reducing the diameter from {approx}200 {micro}m to the target diameter. The nickel plate is hard, smooth, and adherent, and covers the orifice interior surfaces uniformly.

  16. MOX fuel development: Experience in Argentina

    International Nuclear Information System (INIS)

    Marchi, D.E.; Adelfang, P.; Menghini, J.E.

    1999-01-01

    Since 1973, when a laboratory conceived for the safe manipulation of a few hundred grams of plutonium was built, the CNEA (Argentinean Atomic Energy Commission) has been involved in the small-scale development of MOX fuel technology. The plutonium laboratory consists in a glove box facility (α Facility) featuring the necessary equipment to prepare MOX fuel rods for experimental irradiations and to carry out studies on preparative processes development and chemical and physical characterization. The irradiation of the first prototypes of (U,Pu)O 2 fuels fabricated in Argentina began in 1986. These experiments were carried out in the HFR (High Flux Reactor)- Petten , Holland. The rods were prepared and controlled in the CNEA's a Facility. The post-irradiation examinations (PIE) were performed in the KFK (Kernforschungszentrum Karlsruhe), Germany and the JRC (Joint Research Center), Petten. In the period 1991-1995, the development of new laboratory methods of co-conversion of uranium and plutonium were carried out: reverse strike co-precipitation of ADU-Pu(OH) 4 and direct denitration using microwaves. The reverse strike process produced pellets with a high sintered density, excellent micro-homogeneity and good solubility in nitric acid. Liquid wastes showed a very low content of actinides and the process is easy to operate in a glove box environment. The microwave direct denitration was optimized with uranium alone and the conditions to obtain high density pellets, with a good microstructure, without using a milling step, have been developed. At present, new experiments are being carried out to improve the reverse strike co-precipitation process and direct microwave denitration. A new glove box is being installed at the plutonium laboratory, this glove box has process equipment designed to recover scrap from previous fabrication campaigns, and to co-convert mixed U-Pu solutions by direct microwave denitration. (author)

  17. Present status of uranium-plutonium mixed carbide fuel development for LMFBR

    International Nuclear Information System (INIS)

    Handa, Muneo; Suzuki, Yasufumi.

    One Oarai characteristic of a carbide fuel is that its doubling time is about 13 years which is only about half as long as that of an oxide fuel. The development of carbide fuels in the past ten years has been truly remarkable. Especially, through the new fuel development program initiated in 1974 in the United States, success has been achieved with respect to He- and Na-bond fuels in obtaining a 16 a/o burning rate without damage to cladding tubes. In 1984 at FFTF, a radiation of a fuel assembly consisting 91 fuel pins is contemplated. On the other hand, in Japan, in 1974, a Fuel Research Wing specializing in the study of carbide fuels was constructed in the Oarai Laboratory of the Atomic Energy Research Institute and in the fall of 1982, was successful in fabricating two carbide fuel pins having different chemical compositions

  18. Reproduction of the RA reactor fuel element fabrication; Reprodukcija izrade gorivnog elementa za reaktor RA

    Energy Technology Data Exchange (ETDEWEB)

    Novakovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals. Ovaj rad sadrzi devet priloga: 1. Zavrsni izvestaj o podzadatku 08/12, ispitivanje elementa goriva reaktora RA; 2. Koncepcija izrade gorivnog elementa reaktora RA; 3. Ispitivanje mikrostrukture gorivnog elementa reaktora RA; 4. Zavrsni izvestaj o podzadatku 08/13, dobijanje binarnih legura urana sa legirajucim komponentama Al, Mo, Zr, Nb i B; 5. Dobijanje legure U-Al; 6. Zavrsni izvestaj o podzadacima 08/14 i 08/16; 7. Zavrsni izvestaj o podzadatku 08/32, difuziona veza goriva i kosuljice gorivnog elementa reaktora RA; 8. Zavrsni izvestaj o podzadatku 08/33, izrada kosuljice gorivnog elementa reaktora RA; 9. Zavrsni izvestaj o podzadatku 08/36, difuzija kod metala u cvrstom stanju.

  19. Development of ISA procedure for uranium fuel fabrication and enrichment facilities: overview of ISA procedure and its application

    International Nuclear Information System (INIS)

    Yamate, Kazuki; Yamada, Takashi; Takanashi, Mitsuhiro; Sasaki, Noriaki

    2013-01-01

    Integrated Safety Analysis (ISA) procedure for uranium fuel fabrication and enrichment facilities has been developed for aiming at applying risk-informed regulation to these uranium facilities. The development has carried out referring to the ISA (NUREG-1520) by the Nuclear Regulatory Commission (NRC). The paper presents purpose, principles and activities for the development of the ISA procedure, including Risk Level (RL) matrix and grading evaluation method of IROFS (Items Relied on for Safety), as well as general description and features of the procedure. Also described in the paper is current status in application of risk information from the ISA. Japanese four licensees of the uranium facilities have been conducting ISA for their representative processes using the developed procedure as their voluntary safety activities. They have been accumulating experiences and knowledge on the ISA procedure and risk information through the field activities. NISA (Nuclear and Industrial Safety Agency) and JNES (Japan Nuclear Energy Safety Organization) are studying how to use such risk information for the safety regulation of the uranium facilities, taking into account the licensees' experiences and knowledge. (authors)

  20. Development of advanced mixed oxide fuels for plutonium management

    International Nuclear Information System (INIS)

    Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

    1997-01-01

    A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium. (author)

  1. Development of advanced mixed oxide fuels for plutonium management

    International Nuclear Information System (INIS)

    Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

    1997-06-01

    A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium

  2. Fuel performance, design and development

    International Nuclear Information System (INIS)

    Prasad, P.N.; Tripathi, Rahul Mani; Soni, Rakesh; Ravi, M.; Vijay Kumar, S.; Dwivedi, K.P.; Pandarinathan, P.R.; Neema, L.K.

    2006-01-01

    The normal fuel configurations for operating 220 MWe and 540 MWe PHWRs are natural uranium dioxide 19-element and 37- element fuel bundle types respectively. The fuel configuration for BWRs is 6 x 6 fuel. So far, about 330 thousand PHWR fuel bundles and 3500 number of BWR bundles have been irradiated in the 14 PHWRs and 2 BWRs. Improvements in fuel design, fabrication, quality control and operating practices are continuously carried out towards improving fuel utilization as well as reducing fuel failure rate. Efforts have been put to improve the fuel bundle utilization by increasing the fuel discharge burnup of the natural uranium bundles The overall fuel failure rate currently is less than 0.1 % . Presently the core discharge burnups in different reactors are around 7500 MWD/TeU. The paper gives the fuel performance experience over the years in the different power reactors and actions taken to improve fuel performance over the years. (author)

  3. DEVELOPMENT OF OTM SYNGAS PROCESS AND TESTING OF SYNGAS-DERIVED ULTRA-CLEAN FUELS IN DIESEL ENGINES AND FUEL CELLS; TOPICAL

    International Nuclear Information System (INIS)

    E.T. Robinson; James P. Meagher; Ravi Prasad

    2001-01-01

    This topical report summarizes work accomplished for the Program from January 1 through September 15, 2001 in the following task areas: Task 1--materials development; Task 2--composite element development; Task 3--tube fabrication; Task 4--reactor design and process optimization; Task 5--catalyst development; Task 6--P-1 operation; Task 8--fuels and engine testing; and Task 10--project management. OTM benchmark material, LCM1, exceeds the commercial oxygen flux target and was determined to be sufficiently robust to carry on process development activities. Work will continue on second-generation OTM materials that will satisfy commercial life targets. Three fabrication techniques for composite elements were determined to be technically feasible. These techniques will be studied and a lead manufacturing process for both small and large-scale elements will be selected in the next Budget Period. Experiments in six P-0 reactors, the long tube tester (LTT) and the P-1 pilot plant were conducted. Significant progress in process optimization was made through both the experimental program and modeling studies of alternate reactor designs and process configurations. Three tailored catalyst candidates for use in OTM process reactors were identified. Fuels for the International diesel engine and Nuvera fuel cell tests were ordered and delivered. Fuels testing and engine development work is now underway

  4. Establishing QC/QA system in the fabrication of nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Suh, K.S.; Choi, S.K.; Park, H.G.; Park, T.G.; Chung, J.S.

    1980-01-01

    Quality control instruction manuals and inspection methods for UO 2 powder and zircaloy materials as the material control, and for UO 2 pellets and nuclear fuel rods as the process control were established. And for the establishment of Q.A programme, the technical specifications of the purchased materials, the control regulation of the measuring and testing equipments, and traceability chart as a part of document control have also been provided and practically applied to the fuel fabrication process

  5. Radiological control aspects of the fabrication of the Light Water Breeder Reactor core (LWBR Development Program)

    International Nuclear Information System (INIS)

    Schultz, B.G.

    1979-05-01

    A description is presented of the radiological control aspects of the fabrication of the Light Water Breeder Reactor (LWBR) core. Included are the radiological control criteria applied for the design and use of fabrication facilities, the controls and limits imposed to minimize radiaion exposure to personnel, and an evaluation of the applied radiological program in meeting the program objectives. The goal of the LWBR program is to develop the technology to breed in light water reactors so that nuclear fuel may be used significantly more efficiently in these reactors. This technology is being developed by designing and fabricating a breeder reactor core, utilizing thoria (ThO 2 ) and binary thoria--urania (ThO 2 - 233 UO 2 ) fuel, to be operated in the existing pressurized water reactor plant owned by the Department of Energy at Shippingport, Pennsylvania

  6. Decision fundamentals for emergencies in fuel fabrication plants

    International Nuclear Information System (INIS)

    Thomas, W.; Pfeffer, W.; Wiesemes, J.

    1995-01-01

    This report is a compilation of fundamental physical and chemical data for emergencies in fuel element fabrication facilities. The release of uranium and plutonium and a criticality accident constitute the main hazards to be considered. In addition information related to the chemical risk of a release of toxic uranium hexafluoride is included in the report. This fundamental information is to be applied in planning emergency measures and could be useful as advisory material for the emergency staff. (orig.) [de

  7. Microencapsulation and fabrication of fuel pellets for inertial confinement fusion

    International Nuclear Information System (INIS)

    Nolen, R.L. Jr.; Kool, L.B.

    1981-01-01

    Various microencapsulation techniques were evaluated for fabrication of thermonuclear fuel pellets for use in existing experimental facilities studying inertial confinement fusion and in future fusion-power reactors. Coacervation, spray drying, in situ polymerization, and physical microencapsulation methods were employed. Highly spherical, hollow polymeric shells were fabricated ranging in size from 20 to 7000 micron. In situ polymerization microencapsulation with poly(methyl methacrylate) provided large shells, but problems with local wall defects still must be solved. Extension to other polymeric systems met with limited success. Requirements for inertial confinement fusion targets are described, as are the methods that were used

  8. Practical experience in the application of quality control in water-reactor fuel fabrication

    International Nuclear Information System (INIS)

    Vollath, D.

    1984-07-01

    Highly industrialized countries have gained vast experience in manufacturing water reactor fuel. Manufacturing is followed by a stringent system of quality assurance and quality control. The Seminar on Practical Experience in the Application of Quality Control in Water-Reactor Fuel Fabrication provided a forum for an exchange of information on methods and systems of quality assurance and quality control for reactor fuel. In addition, many developing countries which have started or intend to set up a nuclear fuel industry are interested in the application of quality assurance and quality control. This meeting has been preceded by two different series of conferences: the IAEA meetings 1976 in Oslo, 1978 in Prague and 1979 in Buenos Aires, and the Karlsruhe meetings on Characterization and Quality Control of Nuclear Fuel held in 1978 and 1981. Quality control and quality assurance has many different facets. Unlike the purely technical aspects, covered by the Karlsruhe conference series, the IAEA meetings always relate to a wider field of topics. They include governmental regulations and codes for practical quality assurance. This volume contains the papers presented at the seminar and a record of the discussions. (orig.)

  9. Development of packagings for 'MONJU' blanket fuel assemblies

    International Nuclear Information System (INIS)

    Shibata, Kan; Ouchi, Yuichiro; Matsuzaki, Masaaki; Okuda, Yoshihisa

    1995-01-01

    Blanket assemblies for prototype Fast Breeder Reactor 'MONJU' are made at commercial fuel fabrication plants capable of handling deplete Uranium in Japan. For the purpose of transport the assemblies are inserted into a packaging that is set horizontally at the fabrication plants because of compatibility with equipment installed at the plants. On the other hand, the assemblies must be taken out from the packaging set vertically at 'MONJU' due to compatibility. For this reason development of a new packaging, which makes it possible to take assemblies in and out both horizontally and vertically, is needed to carry out transport of assemblies for reload. The development and fabrication of the packagings, taking about two years, were completed in March 1995. The packagings were used in transport of assemblies in June 1995 for the first change. This report introduces the outline of the packaging and confirmation tests done in the process of development. (author)

  10. The fabrication and performance of Canadian silicide dispersion fuel for test reactors

    International Nuclear Information System (INIS)

    Sears, D.F.; Wood, J.C.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.

    1985-01-01

    Fuel fabrication effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer material accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been Irradiated successfully to burnups in the range 65-80 atomic percent. Irradiations have also been conducted on mini-elements having 1.2 mm diameter holes In their mid-sections, some drilled before irradiation and others after irradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated silicide aspersion fuels. Small segments of fuel cores released 85 Kr starting at about 520 deg. C and peaking at about 680 deg C. After a holding period of 1 hour at 720 deg. C a secondary 85 Kr peak occurred during cooling (at about 330 deg. C) probably due to thermal contraction cracking. Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally, with encouraging results. After about 0.25 h at 530 deg. C the aluminum cladding developed very localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. (author)

  11. The fabrication and performance of Canadian silicide dispersion fuel for test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sears, D F; Wood, J C; Berthiaume, L C; Herbert, L N; Schaefer, J D

    1985-07-01

    Fuel fabrication effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer material accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been Irradiated successfully to burnups in the range 65-80 atomic percent. Irradiations have also been conducted on mini-elements having 1.2 mm diameter holes In their mid-sections, some drilled before irradiation and others after irradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated silicide aspersion fuels. Small segments of fuel cores released {sup 85}Kr starting at about 520 deg. C and peaking at about 680 deg C. After a holding period of 1 hour at 720 deg. C a secondary {sup 85}Kr peak occurred during cooling (at about 330 deg. C) probably due to thermal contraction cracking. Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally, with encouraging results. After about 0.25 h at 530 deg. C the aluminum cladding developed very localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. (author)

  12. Probabilistic safety analysis for nuclear fuel cycle facilities, an exemplary application for a fuel fabrication plant

    International Nuclear Information System (INIS)

    Gmal, B.; Gaenssmantel, G.; Mayer, G.; Moser, E.F.

    2013-01-01

    In order to assess the risk of complex technical systems, the application of the Probabilistic Safety Assessment (PSA) in addition to the Deterministic Safety Analysis becomes of increasing interest. Besides nuclear installations this applies to e. g. chemical plants. A PSA is capable of expanding the basis for the risk assessment and of complementing the conventional deterministic analysis, by which means the existing safety standards of that facility can be improved if necessary. In the available paper, the differences between a PSA for a nuclear power plant and a nuclear fuel cycle facility (NFCF) are discussed in shortness and a basic concept for a PSA for a nuclear fuel cycle facility is described. Furthermore, an exemplary PSA for a partial process in a fuel assembly fabrication facility is described. The underlying data are partially taken from an older German facility, other parts are generic. Moreover, a selected set of reported events corresponding to this partial process is taken as auxiliary data. The investigation of this partial process from the fuel fabrication as an example application shows that PSA methods are in principle applicable to nuclear fuel cycle facilities. Here, the focus is on preventing an initiating event, so that the system analysis is directed to the modeling of fault trees for initiating events. The quantitative results of this exemplary study are given as point values for the average occurrence frequencies. They include large uncertainties because of the limited documentation and data basis available, and thus have only methodological character. While quantitative results are given, further detailed information on process components and process flow is strongly required for robust conclusions with respect to the real process. (authors)

  13. The development of lower enrichment fuels for Canadian research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A; Belanger, L; Grolway, C M [AECL, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Foo, M T [CRNL, Combustion Engineering Superheater Ltd., Moncton, NB (Canada)

    1983-08-01

    As part of the world wide move to proliferation resistant fuels, new fuels which use reduced enrichment uranium are being developed for use in the NRX and NRU reactors. A fuel consisting of particles of a USiAl alloy dispersed in an Al matrix has been selected for development along with Al-37 wt% U alloy and Al-U{sub 3}O{sub 8} cermet as backup fuels. This report outlines the progress made in the development of the Al-USiAl and Al-37 wt% U. Results show that good quality extruded rods containing either fuel can be made with techniques similar to those used to fabricate the current NRX and NRU fuels. However, the new fuels will be more expensive to make. Although the oxidation behaviour of the Al-USiAl is not as good as that of the Al-U alloys, its corrosion behaviour in high temperature water does not seem much worse. The oxidation and aqueous corrosion of A-37 wt% U are not much different from those of the Al-U alloys currently used. (author)

  14. Inert matrix fuel in dispersion type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Savchenko, A.M. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)]. E-mail: sav@bochvar.ru; Vatulin, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Morozov, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Sirotin, V.L. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Dobrikova, I.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kulakov, G.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Ershov, S.A. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kostomarov, V.P. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Stelyuk, Y.I. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)

    2006-06-30

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg{sup -1} (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  15. Inert matrix fuel in dispersion type fuel elements

    Science.gov (United States)

    Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

    2006-06-01

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  16. Facility safeguards at an LEU fuel fabrication facility in Japan

    International Nuclear Information System (INIS)

    Kuroi, H.; Osabe, T.

    1984-01-01

    A facility description of a Japanese LEU BWR-type fuel fabrication plant focusing on safeguards viewpoints is presented. Procedures and practices of MC and A plan, measurement program, inventory taking, and the report and record system are described. Procedures and practices of safeguards inspection are discussed and lessons learned from past experiences are reviewed

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  18. Development of fresh fuel packaging for ATR demonstration reactor

    International Nuclear Information System (INIS)

    Kurakami, J.; Kurita, I.

    1993-01-01

    Related to development of the demonstration advanced thermal reactor, it is necessary and important to develop transport packaging which is used for transporting fresh fuel assemblies. Therefore, the packaging is now being developed in Power Reactor and Nuclear Fuel Development Corporation (PNC). Currently, PNC is fabricating two prototype packagings based on the final design, and land cruising and vibration tests, handling performance tests and prototype packaging tests will be executed with prototype packagings in order to experimentally confirm the soundness of packaging and its contents and the propriety of design technique. This paper describes the summary of general specifications and structures of this packaging and the summary of preliminary safety analysis of package. (J.P.N.)

  19. Development of nuclear fuel for integrated reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  20. Development of nuclear fuel for integrated reactor

    International Nuclear Information System (INIS)

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

    1999-04-01

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

  1. Fabrication of AA6061-T6 Plate Type Fuel Assembly Using Electron Beam Welding Process

    International Nuclear Information System (INIS)

    Kim, Soosung; Seo, Kyoungseok; Lee, Donbae; Park, Jongman; Lee, Yoonsang; Lee, Chongtak

    2014-01-01

    AA6061-T6 aluminum alloy is easily welded by conventional GTAW (Gas Tungsten Arc Welding), LBW (Laser Beam Welding) and EBW. However, certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes possess the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the nuclear fuel plate fabrication and assembly, a fundamental EBW experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the welding process, and satisfy the requirements of the weld quality, EBW apparatus using an electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. The EB weld quality of AA6061-T6 aluminum alloy for the fuel plate assembly has been also studied by the shrinkage measurement and weld inspection using computed tomography. This study was carried out to determine the suitable welding parameters and to evaluate tensile strength of AA6061-T6 aluminum alloy. In the present experiment, satisfactory electron beam welding process of the full-sized sample was being developed. Based on this fundamental study, fabrication of the plate-type fuel assembly will be provided for the future Ki-Jang research reactor project

  2. A Flexible Ascorbic Acid Fuel Cell with a Microchannel Fabricated using MEMS Techniques

    Science.gov (United States)

    Mogi, Hiroshi; Fukushi, Yudai; Koide, Syohei; Sano, Ryohei; Sasaki, Tsubasa; Nishioka, Yasushiro

    2013-12-01

    We fabricated a miniature ascorbic acid fuel cells equipped with a microchannel for the circulation of ascorbic acid (AA) solution using micro electronic mechanical system techniques. The fuel cell was fabricated on a flexible polyimide substrate, and its porous carbon-coated aluminium (Al) electrodes of 2.8 mm in width and 11 mm in length were formed using photolithography and screen-printing techniques. The porous carbon was deposited by screen-printing of carbon-black ink on the Al electrode surfaces in order to increase the effective electrode surface area and to absorb more enzymes on the cathode surface. The microchannel with a depth of 200 μm was fabricated using a hot-embossing technique. A maximum power of 0.60 μW at 0.58 V that corresponds to a power density of 1.83 μW/cm2 was realized by introducing a 200 mM concentrated AA solution at room temperature.

  3. Progress of the Russian RERTR program: Development of new-type fuel elements for Russian-built research reactors

    International Nuclear Information System (INIS)

    Vatulin, A. V.; Stetskiy, Y.A.; Mishunin, V.A.; Suprun, V.B.; Dobrikova, I.V.

    2002-01-01

    The new design of pin-type fuel elements and fuel assembly on their basis for Russian research reactors has been developed. The number of following activities has been performed: computational and experimental substantiation of fuel element design; development of fabrication process of fuel elements; manufacturing of experimental assembly for lifetime in-pile tests. The relevant fuel assemblies are considered to be perspective for usage as low-enriched fuel for Russian research reactors. (author)

  4. MOX Fabrication Isolation Considerations

    Energy Technology Data Exchange (ETDEWEB)

    Eric L. Shaber; Bradley J Schrader

    2005-08-01

    This document provides a technical position on the preferred level of isolation to fabricate demonstration quantities of mixed oxide transmutation fuels. The Advanced Fuel Cycle Initiative should design and construct automated glovebox fabrication lines for this purpose. This level of isolation adequately protects the health and safety of workers and the general public for all mixed oxide (and other transmutation fuel) manufacturing efforts while retaining flexibility, allowing parallel development and setup, and minimizing capital expense. The basis regulations, issues, and advantages/disadvantages of five potential forms of isolation are summarized here as justification for selection of the preferred technical position.

  5. Fabrication of the fuel elements cladding for utilization in the fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Schaeffer, L.; Sefidvash, F.

    1986-01-01

    A method for the fabrication of cladding of the spherical fuel elements for the utilization in the fluidized bed nuclear reactor is presented. Some prelimminary experiments were performed to adopt a method which adapt itself to mass production with the desired high quality. Still methods for cladding fabrication are under study. (Author) [pt

  6. Standard model for safety analysis report of fuel fabrication plants

    International Nuclear Information System (INIS)

    1980-09-01

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

  7. Method to fabricate block fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1977-01-01

    The fabrication of block fuel elements for gas-cooled high temperature reactors can be improved upon by adding 0.2 to 2 wt.% of a hydrocarbon compound to the lubricating mixture prior to pressing. Hexanol or octanol are named as substances. The dimensional accuracy of the block is thus improved. 2 examples illustrate the method. (RW) [de

  8. Method to fabricate block fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1978-01-01

    The fabrication of block fuel elements for gas-cooled high temperature reactors can be improved upon by adding 0.2 to 2 wt.% of a hydrocarbon compound to the lubricating mixture prior to pressing. Hexanol or octanol are named as substances. The dimensional accuracy of the block is thus improved. 2 examples illustrate the method. (orig./PW)

  9. Development of Coated Particle Fuel Technology

    International Nuclear Information System (INIS)

    Cho, Moon Sung; Kim, B. G.; Kim, Y. K.

    2009-04-01

    UO 2 kernel fabrication technology was developed at the lab sacle(20∼30g-UO 2 /batch). The GSP technique, modified method of sol-gel process, was used in the preparation of spherical ADU gel particle and these particles were converted to UO 3 and UO 2 phases in calcination furnace and sintering furnace respectively. Based on the process variables optimized using simulant kernels in 1-2 inch beds, SiC TRISO-coated particles were fabricated using UO 2 kernel. Power densities of TRISO coated particle fuels and gamma heat of the tubes are calculated as functions of vertical location of the fuel specimen in the irradiation holes by using core physics codes, MCNP and Helios. A finite model was developed for the calculations of temperatures and stresses of the specimen and the irradiation tubes. Dimensions of the test tubes are determined based on the temperatures and stresses as well as the gamma heat generated at the given condition. 9 modules of the COPA code (MECH, FAIL, TEMTR, TEMBL, TEMPEB, FPREL, MPRO, BURN, ABAQ), the MECH, FAIL, TEMTR, TEMBL, TEMPEB, and FPREL were developed. The COPA-FPREL was verified through IAEA CRP-6 accident benchmarking problems. KAERI participated in the round robin test of IAEA CRP-6 program to characterize the diameter, sphericity, coating thickness, density and anisotropy of coated particles provided by Korea, USA and South Africa. The inspection and test plan describing specifications and inspection method of coated particles was developed to confirm the quality standard of coated particles. The quality inspection instructions were developed for the inspection of coated particles by particle size analyzer, density inspection of coating layers by density gradient column, coating thickness inspection by X-ray, and inspection of optical anistropy factor of PyC layer. The quality control system for the TRISO-coated particle fuel was derived based on the status of quality control systems of other countries

  10. Procedure for the fabrication of ceramic fuel pellets with an adjustable structure

    International Nuclear Information System (INIS)

    Henke, M.; Klemm, U.; Sobek, D.

    1986-01-01

    The invention concerns a procedure for the fabrication of ceramic fuel pellets of UO 2 , PuO 2 , ThO 2 and their mixtures with an adjustable structure. Before or during the milling the particle shaped fuel pellets have been added polyethylenglycol in a 20 - 60 % aqueous solution with an amount of 0.5 - 2.0 % in weight. This additive has an effect on a controlled pore formation and grain growth advancement

  11. DUPIC fuel fabrication using spent PWR fuels at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Dong; Yang, Myung Seung; Ko, Won Il and others

    2000-12-01

    This document contains DUPIC fuel cycle R and D activities to be carried out for 5 years beyond the scope described in the report KAERI/AR-510/98, which was attached to Joint Determination for Post-Irradiation Examination of irradiated nuclear fuel, by MOST and US Embassy in Korea, signed on April 8, 1999. This document is purposely prepared as early as possible to have ample time to review that the over-all DUPIC activities are within the scope and contents in compliance to Article 8(C) of ROK-U.S. cooperation agreement, and also maintain the current normal DUPIC project without interruption. Manufacturing Program of DUPIC Fuel in DFDF and Post Irradiation Examination of DUPIC Fuel are described in Chapter I and Chapter II, respectively. In Chapter III, safeguarding procedures in DFDF and on-going R and D on DUPIC safeguards such as development of nuclear material accounting system and development of containment/surveillance system are described in details.

  12. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1979-01-01

    The regulations are entirely revised under the law for the regulations of nuclear materials, nuclear fuel materials and reactors and provisions concerning the fabricating business in the order for execution of the law. Basic concepts and terms are defined, such as: exposure dose; accumulative dose; controlled area; inspected surrounding area; employee and radioactive waste. The application for permission of the fabricating business shall include: location of processing facilities; structure of building structure and equipment of chemical processing facilities; molding facilities; structure and equipment of covering and assembling facilities, storage facilities of nuclear fuel materials and disposal facilities of radioactive waste, etc. Records shall be made and kept for particular periods in each works and place of enterprise on inspection of processing facilities, control of dose, operation, maintenance, accident of processing facilities and weather. Specified measures shall be taken in controlled area and inspected surrounding area to restrict entrance. Measures shall be made not to exceed permissible exposure dose for employees defined by the Director General of Science and Technology Agency. Inspection and check up of processing facilities shall be carried on by employees more than once a day. Operation of processing facilities, transportation in the works and enterprise, storage, disposal, safety securing, report and measures in dangerous situations, etc. are in detail prescribed. (Okada, K.)

  13. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

    2008-01-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

  14. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    Science.gov (United States)

    Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

    2008-12-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

  15. Development of nuclear fuel materials for research reactor

    International Nuclear Information System (INIS)

    Kim, Chang Kyu; Park, H. D.; Kim, K. H.; Lee, J. T.; Ryu, W. S.; Hwang, W.; Kim, H. N.; Kim, H. I.; Kwon, H. I.; Park, C.; Lee, B. C.; Park, J. M.; Lee, C. S.; Chae, H. T.; Im, N. J.; Cho, M. S.; Im, I. C.; Nam, C.; Lee, D. B.; Goh, Y. M.; Kim, J. D.; Ahn, H. S.; Woo, Y. M.; Chang, S. J.; Cho, H. D.

    1997-09-01

    This project has aimed at the development of U 3 Si dispersion fuel for the localization of HANARO fuel and the application of atomization process to advanced RERTR fuel development. The design criteria were established through the modified computer codes. Design documents were prepared and issued. The acceptable co-extrusion cladding was achieved. The electron beam welding technology has been developed and the sealing of the end plug and cladding was accomplished without defects. The atomization fuel meats have about 200% higher elongation and about 20% higher than comminution fuel meats. The thermal compatibility test showed that atomization fuel have about 30% higher stability that the comminution fuel. The pressure drops of 18 rods fuel assembly and 36 rods fuel assembly were measured to have 213 kPa and 205 kPa respectively. Apparent wear was not found in endurance test. The irradiation fuel was designed and fabricated by using low enriched uranium metal following the developed Q/A system. The safety analysis of irradiation fuel assembly was performed through linear power calculation by using MCNP4A code and centerline temperature calculation by using DIFAIR code. The quality assurance system has been established. The quality inspection technologies were developed. By acquiring the license, low enriched uranium of 100 kg as well as depleted uranium can be used. U 3 Si 2 -Al fuel swelled less than comminution fuel irrespective of temperature and fuel fraction in a compatibility test. The atomized U-10wt.%Mo powder were found to have gamma phase of isotropic structure. Gamma structure remained with a little swelling without any structure change at 400 deg C for 100 hours. Irradiation miniplate and test rig were designed preliminary manufactured. Thermal hydraulic and linear power calculations were performed by using PLTEMP and MCNP4A computer codes respectively. The hydraulic test showed that the pressure drop met the HANARO requirement. The vibration

  16. Development of TVSA VVER-1000 fuel

    International Nuclear Information System (INIS)

    Samoilov, O.; Kaydalov, V.; Romanov, A.; Falkov, A.; Morozkin, O.; Sholin, E.

    2013-01-01

    The TVSA fuel assemblies with a rigid angle-piece skeleton operate at 21 VVER-1000 units of Kalinin NPP, and Ukrainian, and Czech and Bulgarian NPPs. The total of more than 6,000 TVSA fuel assemblies have been fabricated. High lifetime performance has been achieved, namely, the maximum FA burnup is 65 MW∙day/kgU; maximum fuel rod burnup is 72 MW∙day/kgU; the lifetime is 50,000 EFPH. The TVSA fuel assembly is being improved to enhance its technical and economic performance and competitiveness of the Russian fuel for the VVER-1000 reactor: 1) Reliability and safety are being enhanced; repairability is being ensured. 2) High burnup levels in fuel are being ensured. 3) The uranium content in FAs is being increased. 4) The operational life is being extended. 5) Thermal-technical characteristics of FAs are being improved. The basic TVSA fuel assembly design evolved into the TVSA-PLUS with the fuel column elongated by 150 mm. The TVSA-PLUS fuel assembly has been in operation since 2010 at Kalinin NPP power units; an eighteen-month cycle is implemented at the uprated power of 104%. The TVSA-12PLUS fuel assembly has been developed with an elongated fuel column, optimized spacer grid positions (the spacer grid pitch is 340 mm) and with ensuring higher rigidity for the skeleton. It is provided for that fuel rods with the elevated uranium content and mixing intensifier grids will be used. The TVSA-T is developed for VVER-1000 reactor cores at the Temelin NPP. The TVSA-T is characterized by a load-carrying skeleton formed with angle-pieces and combined spacer grids that incorporate mixer grids. The TVSA-T design won the international tender to supply fuel to the Temelin NPP in the Czech Republic, and currently Temelin NPP Unit 1 and 2 are operating with the cores fully loaded with TVSA-Ts

  17. Field characterization of plutonium aerosols in mixed-oxide fuel fabrication

    International Nuclear Information System (INIS)

    Newton, G.J.; Teague, S.V.; Yeh, H.C.

    1976-01-01

    Nuclear reactor fuel pellets of PuO 2 and UO 2 are fabricated within safety enclosures at Babcock and Wilcox's Parker Township Site near Apolla, Pa. Nineteen sample runs were taken from within glove boxes of aerosols formed during powder comminution and blending. Eight sampling runs were also taken of a centerless grinding operation during routine industrial operations. A small seven-stage cascade impactor and the Lovelace Aerosol Particle Separator (LAPS) were used to determine aerodynamic size distribution and gross alpha aerosol concentrations. The potential toxicity of inhaled plutonium originating in the nuclear fuel cycle following accidental releases of these aerosols and possible inhalation by industrial workers is considered

  18. State of the art of UO2 fuel fabrication processes

    International Nuclear Information System (INIS)

    Henke, M.; Klemm, U.

    1980-01-01

    Starting from the need of UO 2 for thermal power reactors in the period from 1980 to 1990 and the role of UF 6 conversion into UO 2 within the fuel cycle, the state-of-the-art of the three established industrial processes - ADU process, AUC process, IDR process - is assessed. The number of process stages and requirements on process management are discussed. In particular, the properties of the fabricated UO 2 powders, their influence on the following pellet production and on operational behaviour of the fuel elements under reactor conditions are described. Hence, an evaluation of the three essential conversion processes is derived. (author)

  19. Options for converting excess plutonium to feed for the MOX fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Watts, Joe A [Los Alamos National Laboratory; Smith, Paul H [Los Alamos National Laboratory; Psaras, John D [Los Alamos National Laboratory; Jarvinen, Gordon D [Los Alamos National Laboratory; Costa, David A [Los Alamos National Laboratory; Joyce, Jr., Edward L [Los Alamos National Laboratory

    2009-01-01

    The storage and safekeeping of excess plutonium in the United States represents a multibillion-dollar lifecycle cost to the taxpayers and poses challenges to National Security and Nuclear Non-Proliferation. Los Alamos National Laboratory is considering options for converting some portion of the 13 metric tons of excess plutonium that was previously destined for long-term waste disposition into feed for the MOX Fuel Fabrication Facility (MFFF). This approach could reduce storage costs and security ri sks, and produce fuel for nuclear energy at the same time. Over the course of 30 years of weapons related plutonium production, Los Alamos has developed a number of flow sheets aimed at separation and purification of plutonium. Flow sheets for converting metal to oxide and for removing chloride and fluoride from plutonium residues have been developed and withstood the test oftime. This presentation will address some potential options for utilizing processes and infrastructure developed by Defense Programs to transform a large variety of highly impure plutonium into feedstock for the MFFF.

  20. Development of very-high-density low-enriched-uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Trybus, C.L.; Wiencek, T.C.

    1997-01-01

    Following a hiatus of several years and following its successful development and qualification of 4.8 g U cm -3 U 3 Si 2 -Al dispersion fuel for application with low-enriched uranium in research and test reactors, the US Reduced Enrichment for Research and Test Reactors program has embarked on the development of even-higher-density fuels. Our goal is to achieve uranium densities of 8-9 g cm -3 in aluminum-based dispersion fuels. Achieving this goal will require the use of high-density, γ-stabilized uranium alloy powders in conjunction with the most-advanced fuel fabrication techniques. Key issues being addressed are the reaction of the fuel alloys with aluminum and the irradiation behavior of the fuel alloys and any reaction products. Test irradiations of candidate fuels in very-small (micro) plates are scheduled to begin in the Advanced Test Reactor during June, 1997. Initial results are expected to be available in early 1998. We are performing out-of-reactor studies on the phase structure of the candidate alloys on diffusion of the matrix material into the aluminum. In addition, we are modifying our current dispersion fuel irradiation behavior model to accommodate the new fuels. Several international partners are participating in various phases of this work. (orig.)

  1. Research and development projects, new processes binded at the stop of a fuel fabrication plant (Nukem-A)

    International Nuclear Information System (INIS)

    Wehner, E.; Sohnius, B.

    1992-01-01

    This research work is aimed at the assessment of new procedures in the framework of the decommissioning of the NUKEM-A facility, a plant used for over 30 years for the fabrication of Material Test Reactor (MTR) and Thorium High Temperature Reactor (THTR) fuel elements. Important issues in this work are the preparation of detailed uranium and thorium contamination distribution maps in walls and floors, the execution of various dismantling and decontamination operations under health physics control, the large-scale treatment of arising primary waste and the minimization of secondary waste

  2. Case study application of the IAEA safeguards assessment methodology to a mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Swartz, J.; McDaniel, T.

    1981-01-01

    Science Applications, Inc. has prepared a case study illustrating the application of an assessment methodology to an international system for safeguarding mixed oxide (MOX) fuel fabrication facilities. This study is the second in a series of case studies which support an effort by the International Atomic Energy Agency (IAEA) and an international Consultant Group to develop a methodology for assessing the effectiveness of IAEA safeguards. 3 refs

  3. Evaluation of methods for seismic analysis of nuclear fuel reprocessing and fabrication facilities

    International Nuclear Information System (INIS)

    Arthur, D.F.; Dong, R.G.; Murray, R.C.; Nelson, T.A.; Smith, P.D.; Wight, L.H.

    1978-01-01

    Methods of seismic analysis for critical structures and equipment in nuclear fuel reprocessing plants (NFRPs) and mixed oxide fuel fabrication plants (MOFFPs) are evaluated. The purpose of this series of reports is to provide the NRC with a technical basis for assessing seismic analysis methods and for writing regulatory guides in which methods ensuring the safe design of nuclear fuel cycle facilities are recommended. The present report evaluates methods of analyzing buried pipes and wells, sloshing effects in large pools, earth dams, multiply supported equipment, pile foundations, and soil-structure interactions

  4. Development of a data base system for quality control of MOX fuels

    International Nuclear Information System (INIS)

    Takahashi, Kuniaki; Yamaguchi, Toshihiro; Mishima, Takeshi

    1988-01-01

    For the purpose of improvement and speed up of work concerning quality control for mixed oxide fuel fabrication, we have been developing a data base system having a data base consisting of data as to fabrication conditions and inspects, We have aimed to develop a data base system having capability of analysis and function giving informations as to quality control. The data base system is full interactive system on real time basis, consequently it makes analyzing and editing data easy. It has capability of relational research, numerical analysis, correlation analysis, drawing control charts, histograms, and other figures, and expressing status of fabrication processes using control charts. (author)

  5. Development of compaction technique for spent fuel skeletons

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Sup; Kim, Young Hwan; Jung, Jae Hoo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    To increase the utilization of uranium resources contained in the spent fuel, the spent fuel is reused. For this, the spent fuel is dismantled or spent fuel rod is extracted from the spent fuel assembly. When the rod is extracted, the remaining components of spent fuel assembly, so called a NFBC(Non-Fuel Bearing Components), should be compacted for the final disposal. To this end, several companies developed the NFBC compactors. German company, named as GNS has developed the direct compression devices of the NFBCs for the rod consolidation and installed it at the PKA(2) of pilot conditioning plant. B and W (Babcock and Wilcox) in USA adopted cutting method rather than the compression method and developed the special cutting devices of NFBC which can be applied underwater environment. In this study the characteristics of these two methods was investigated, in terms of fabrication cost of devices, maintainability in a high radioactive environment, required power and work volume for operation. Also, the optimal power source is selected by comparing the maximum power versus the work volume for operation. In addition to these, the reduction ratio of the bulk volume is obtained while varying the cutting length of the NFBC through a series of experiments. Based on the results of analysis and experiments, the cutting method after compression is selected as an optimal volume reduction method and its design specification is obtained. 8 refs., 62 figs., 32 tabs. (Author)

  6. Development of materials for fuel cell application by radiation technology

    International Nuclear Information System (INIS)

    Rhee, Chang Kyu; Lee, Min Ku; Park, Junju; Lee, Gyoungja; Lee, Byung Cheol; Shin, Junhwa; Nho, Youngchang; Kang, Philhyun; Sohn, Joon Yong; Rang, Uhm Young

    2012-06-01

    The development of the single cell of SOFC with low operation temperature at and below 650 .deg. C(above 400 mW/cm 2 ) Ο The development of fabrication method for the single cell of solid oxide fuel cell (SOFC) by dip-coating of nanoparticles such as NiO, YSZ, Ag, and Ag/C, etc. Ο The optimization of the preparation and performance of SOFC by using nanoparticles. Ο The preparation of samples for SOFC with large dimension. The development of fluoropolymer-based fuel cell membranes with crosslinked structure by radiation grafting technique Ο The development of fuel cell membranes with low methanol permeability via the introduction of novel monomers (e. g. vinylbenzyl chloride and vinylether chloride) by radiation grafting technique Ο The development of hydrocarbon fuel cell membrane by radiation crosslinking technique Ο The structure analysis and the evaluations of the property, performance, and radiation effect of the prepared membranes Ο The optimization of the preparation and performance of DMFC fuel cell membrane via the structure-property analysis (power: above 130 mW/cm 2 /50 cm 2 at 5M methanol) Ο The preparation of samples for MEA stack assembly

  7. UO2 fuel pellets fabrication via Spark Plasma Sintering using non-standard molybdenum die

    Science.gov (United States)

    Papynov, E. K.; Shichalin, O. O.; Mironenko, A. Yu; Tananaev, I. G.; Avramenko, V. A.; Sergienko, V. I.

    2018-02-01

    The article investigates spark plasma sintering (SPS) of commercial uranium dioxide (UO2) powder of ceramic origin into highly dense fuel pellets using non-standard die instead of usual graphite die. An alternative and formerly unknown method has been suggested to fabricate UO2 fuel pellets by SPS for excluding of typical problems related to undesirable carbon diffusion. Influence of SPS parameters on chemical composition and quality of UO2 pellets has been studied. Also main advantages and drawbacks have been revealed for SPS consolidation of UO2 in non-standard molybdenum die. The method is very promising due to high quality of the final product (density 97.5-98.4% from theoretical, absence of carbon traces, mean grain size below 3 μm) and mild sintering conditions (temperature 1100 ºC, pressure 141.5 MPa, sintering time 25 min). The results are interesting for development and probable application of SPS in large-scale production of nuclear ceramic fuel.

  8. Development of fast reactor metal fuels containing minor actinides

    International Nuclear Information System (INIS)

    Ohta, Hirokazu; Ogata, Takanari; Kurata, Masaki; Koyama, Tadafumi; Papaioannou, Dimitrios; Glatz, Jean-Paul; Rondinella, Vincenzo V.

    2011-01-01

    Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phenix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU. (author)

  9. Simulated physical inventory verification exercise at a mixed-oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Reilly, D.; Augustson, R.

    1985-01-01

    A physical inventory verification (PIV) was simulated at a mixed-oxide fuel fabrication facility. Safeguards inspectors from the International Atomic Energy Agency (IAEA) conducted the PIV exercise to test inspection procedures under ''realistic but relaxed'' conditions. Nondestructive assay instrumentation was used to verify the plutonium content of samples covering the range of material types from input powders to final fuel assemblies. This paper describes the activities included in the exercise and discusses the results obtained. 5 refs., 1 fig., 6 tabs

  10. ThO2-based pellet fuels - their properties, methods of fabrication, and irradiation performance: a critical assessment of the state of the technology and recommendations for further work

    Energy Technology Data Exchange (ETDEWEB)

    Hart, P.E.; Griffin, C.W.; Hsieh, K.A.; Matthews, R.B.; White, G.D.

    1979-09-01

    This critical assessment of the ThO/sub 2/-UO/sub 2/ pellet fuel technology was conducted in support of the Fuels Refabrication and Development Program (FRAD). Included in this critical review are the following areas: powder preparation; pellet fabrication; fuel chemical, physical, and mechanical properties; and fuel irradiation performance. The authors identify (1) areas where data are either deficient or lacking and (2) requirements for additional development and experimental work.

  11. ThO2-based pellet fuels - their properties, methods of fabrication, and irradiation performance: a critical assessment of the state of the technology and recommendations for further work

    International Nuclear Information System (INIS)

    Hart, P.E.; Griffin, C.W.; Hsieh, K.A.; Matthews, R.B.; White, G.D.

    1979-09-01

    This critical assessment of the ThO 2 -UO 2 pellet fuel technology was conducted in support of the Fuels Refabrication and Development Program (FRAD). Included in this critical review are the following areas: powder preparation; pellet fabrication; fuel chemical, physical, and mechanical properties; and fuel irradiation performance. The authors identify (1) areas where data are either deficient or lacking and (2) requirements for additional development and experimental work

  12. Uranium accountability for ATR fuel fabrication. Part I. A description of the existing system

    International Nuclear Information System (INIS)

    Dolan, C.A.; Nieschmidt, E.B.; Vegors, S.H. Jr.; Wagner, E.P. Jr.

    1977-06-01

    An evaluation of the materials accountability program at the Atomics International fuel fabrication facility in Canoga Park, California, with regard to the fabrication of highly enriched uranium fuel for the Advanced Test Reactor is presented. An analysis is given of the existing standards program, the existing measurements program and the existing statistical analysis procedures. In addition a short discussion is given of our evaluation of the safeguards procedures at Atomics International together with suggestions for possible modifications and improvements. Appendices of this report contain a rather complete description of the Atomics International plant and the flow of highly enriched uranium through the plant as well as the principal documents used for material accountability records

  13. Romanian concern for advanced fuels development

    International Nuclear Information System (INIS)

    Ohai, Dumitru

    2001-01-01

    The Institute for Nuclear Research (ICN), a subsidiary of Romanian Authority for Nuclear Activities, at Pitesti - Romania, has developed a preliminary design of a fuel bundle with 43 elements named SEU 43 for high burnup in CANDU Reactor. A very high experience in nuclear fuels manufacturing and control has also been accumulated. Additionally, on the nuclear site Pitesti there is the Nuclear Fuel Plant (NFP) qualified to manufacturing CANDU 6 type fuel, the main fuel supplier for NPP Cernavoda. A very good collaboration of ICN with NFP can lead to a low cost upgrading the facilities which ensure at present the CANDU standard fuel fabrication to be able of manufacturing also SEU 43 fuel for extended burnup. The financial founds are allocated by Romanian Authority for Nuclear Activities of the Ministry of Industry and Resources to sustain the departmental R and D program 'Nuclear Fuel'. This Program has the main objective to establish a technology for manufacturing a new CANDU fuel type destined for extended burnup. It is studied the possibility to use the Recovered Uranium (RU) resulted from LWR spent fuel reprocessing facility existing in stockpiles. The International Agency for Atomic Energy (IAEA) sustains also this program. By ROM/4/025/ Model Project, IAEA helps ICN to solve the problems regarding materials (RU, Zircaloy 4 tubes) purchasing, devices' upgrading and personnel training. The paper presents the main actions needing to be create the technical base for SEU 43 fuel bundle manufacturing. First step, the technological experiments and experimental fuel element manufacturing, will be accomplished in ICN installations. Second step, the industrial scale, need thorough studies for each installation from NFP to determine tools and technology modification imposed by the new CANDU fuel bundle manufacturing. All modifications must be done such as to the NFP, standard CANDU and SEU fuel bundles to be manufactured alternatively. (author)

  14. Romanian nuclear fuel fabrication and in-reactor fuel operational experience

    International Nuclear Information System (INIS)

    Budan, O.

    2003-01-01

    A review of the Romanian nuclear program since mid 60's is made. After 1990, the new Romanian nuclear power authority, RENEL-GEN, elaborated a realistic Nuclear Fuel Program. This program went through the Romanian nuclear fuel plant qualification with the Canadian (AECL and ZPI) support, restarting in January 1995 of the industrial nuclear fuel production, quality evaluation of the fuel produced before 1990 and the recovery of this fuel. This new policy produced good results. FCN is since 1995 the only CANDU fuel supplier from outside Canada recognised by AECL as an authorised CANDU fuel manufacturer. The in-reactor performances and behaviour of the fuel manufactured by FCN after its qualification have been excellent. Very low - more then five times lesser than the design value - fuel defect rate has been recorded up to now and the average discharge of this fuel was with about 9% greater than the design value. Since mid 1998 when SNN took charge of the production of nuclear generated electricity, FCN made significant progresses in development and procurement of new and more efficient equipment and is now very close to double its fuel production capacity. After the completion of the recovery of the fuel produced before June 1990, FCN is already prepared to shift its fuel production to the so-called 'heavy' bundle containing about 19.3 kg of Uranium per bundle

  15. Development of a Microchannel High Temperature Recuperator for Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    Lukas, Michael [Fuelcell Energy, Inc., Danbury, CT (United States)

    2014-03-24

    This report summarizes the progress made in development of microchannel recuperators for high temperature fuel cell/turbine hybrid systems for generation of clean power at very high efficiencies. Both Solid Oxide Fuel Cell/Turbine (SOFC/T) and Direct FuelCell/Turbine (DFC/T) systems employ an indirectly heated Turbine Generator to supplement fuel cell generated power. The concept extends the high efficiency of the fuel cell by utilizing the fuel cell’s byproduct heat in a Brayton cycle. Features of the SOFC/T and DFC/T systems include: electrical efficiencies of up to 65% on natural gas, minimal emissions, reduced carbon dioxide release to the environment, simplicity in design, and potential cost competitiveness with existing combined cycle power plants. Project work consisted of candidate material selection from FuelCell Energy (FCE) and Pacific Northwest National Laboratory (PNNL) institutional databases as well as from industrial and academic literature. Candidate materials were then downselected and actual samples were tested under representative environmental conditions resulting in further downselection. A microchannel thermal-mechanical model was developed to calculate overall device cost to be later used in developing a final Tier 1 material candidate list. Specifications and operating conditions were developed for both SOFC/T and DFC/T systems. This development included system conceptualization and progression to process flow diagrams (PFD’s) including all major equipment. Material and energy balances were then developed for the two types of systems which were then used for extensive sensitivity studies that used high temperature recuperator (HTR) design parameters (e.g., operating temperature) as inputs and calculated overall system parameters (e.g., system efficiency). The results of the sensitivity studies determined the final HTR design temperatures, pressure drops, and gas compositions. The results also established operating conditions and

  16. Lessons learned from MELOX plant operation and support to design of new MOX fuel fabrication plants

    International Nuclear Information System (INIS)

    Tourre, Joel; Gattegno, Robert; Guay, Philippe; Bariteau, Jean-Pierre

    2005-01-01

    AREVA is participating in the design of the US MOX Fuel Fabrication Facility (MFFF). To support this project and allow the U.S. Department of Energy (DOE) client to reap full benefit from the MELOX operating experience, AREVA, through COGEMA and its engineering subsidiary SGN have implemented a rigorous process to prudently apply MELOX Lessons Learned to the MFFF design. This paper describes the Lessons Learned process, how the process supports the advancement of fuel fabrication technology and, how the results of the process are benefiting the client. (author)

  17. Cost effectiveness of robotics and remote tooling for occupational risk reduction at a nuclear fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Lochard, Jacques

    1989-08-01

    This case study, related to the design stage of a fuel fabrication facility, presents the evaluation of alternative options to manipulate mixed oxide fuel rods in a quality control shop. It is based on a study performed in the framework of the 'MELOX project' developed by COGEMA in France. The methodology for evaluating robotic actions is resulting from a research work part funded by the IAEA under the co-ordinated research programme on 'Comparison of cost-effectiveness of risk reduction among different energy systems', and by the commission of the European Communities under the research and training programme on radiation protection.

  18. Cost effectiveness of robotics and remote tooling for occupational risk reduction at a nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Lochard, Jacques

    1989-01-01

    This case study, related to the design stage of a fuel fabrication facility, presents the evaluation of alternative options to manipulate mixed oxide fuel rods in a quality control shop. It is based on a study performed in the framework of the 'MELOX project' developed by COGEMA in France. The methodology for evaluating robotic actions is resulting from a research work part funded by the IAEA under the co-ordinated research programme on 'Comparison of cost-effectiveness of risk reduction among different energy systems', and by the commission of the European Communities under the research and training programme on radiation protection

  19. Facility effluent monitoring plan for the 300 Area Fuels Fabrication Facility

    International Nuclear Information System (INIS)

    Nickels, J.M.; Brendel, D.F.

    1991-11-01

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP- 0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the first annual report. It shall ensure long-range integrity of the effluent monitoring system by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years. The Fuel Fabrication Facility in the Hanford 300 Area supported the production reactors from the 1940's until they were shut down in 1987. Prior to 1987 the Fuel Fabrication Facility released both airborne and liquid radioactive effluents. In January 1987 the emission of airborne radioactive effluents ceased with the shutdown of the fuels facility. The release of liquid radioactive effluents have continued although decreasing significantly from 1987 to 1990

  20. Constrained Sintering in Fabrication of Solid Oxide Fuel Cells.

    Science.gov (United States)

    Lee, Hae-Weon; Park, Mansoo; Hong, Jongsup; Kim, Hyoungchul; Yoon, Kyung Joong; Son, Ji-Won; Lee, Jong-Ho; Kim, Byung-Kook

    2016-08-09

    Solid oxide fuel cells (SOFCs) are inevitably affected by the tensile stress field imposed by the rigid substrate during constrained sintering, which strongly affects microstructural evolution and flaw generation in the fabrication process and subsequent operation. In the case of sintering a composite cathode, one component acts as a continuous matrix phase while the other acts as a dispersed phase depending upon the initial composition and packing structure. The clustering of dispersed particles in the matrix has significant effects on the final microstructure, and strong rigidity of the clusters covering the entire cathode volume is desirable to obtain stable pore structure. The local constraints developed around the dispersed particles and their clusters effectively suppress generation of major process flaws, and microstructural features such as triple phase boundary and porosity could be readily controlled by adjusting the content and size of the dispersed particles. However, in the fabrication of the dense electrolyte layer via the chemical solution deposition route using slow-sintering nanoparticles dispersed in a sol matrix, the rigidity of the cluster should be minimized for the fine matrix to continuously densify, and special care should be taken in selecting the size of the dispersed particles to optimize the thermodynamic stability criteria of the grain size and film thickness. The principles of constrained sintering presented in this paper could be used as basic guidelines for realizing the ideal microstructure of SOFCs.