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Sample records for htgr nuclear fuel

  1. HTGR fuel reprocessing technology

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  2. HTGR fuel cycle

    International Nuclear Information System (INIS)

    1987-08-01

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

  3. HTGR Fuel performance basis

    International Nuclear Information System (INIS)

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

    1982-05-01

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

  4. HTGR fuel performance basis

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Mack, J.E.

    1977-01-01

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

  6. Overview of HTGR fuel recycle

    International Nuclear Information System (INIS)

    Notz, K.J.

    1976-01-01

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

  7. HTGR fuel element structural design consideration

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  8. HTGR fuel element structural design considerations

    International Nuclear Information System (INIS)

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

    1986-09-01

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

  9. HTGR spent fuel storage study

    International Nuclear Information System (INIS)

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

    1979-04-01

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

  10. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  11. Advances in HTGR fuel performance models

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  12. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

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

    1976-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-04-01

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

  14. The investigation of HTGR fuel regeneration process

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-07-01

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

  15. The investigation of HTGR fuel regeneration process

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  16. HTGR fuel particle crusher design evaluation

    International Nuclear Information System (INIS)

    Johanson, N.W.

    1978-10-01

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

  17. Fission-product retention in HTGR fuels

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  18. Irradiation performance of HTGR recycle fissile fuel

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  19. Flowsheet development for HTGR fuel reprocessing

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  20. HTGR fuel particle crusher: Mark 2 design

    International Nuclear Information System (INIS)

    Baer, J.W.

    1979-06-01

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

  1. HTGR fuel particle crusher: Mark 2 design

    Energy Technology Data Exchange (ETDEWEB)

    Baer, J.W.

    1979-06-01

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

  2. Nondestructive assay of HTGR fuel rods

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1974-01-01

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

  3. Conceptual design of the special nuclear material nondestructive assay and accountability system for the HTGR fuel refabrication pilot plant

    International Nuclear Information System (INIS)

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

    1975-07-01

    The conceptual design of the fissile material assay and accountability system for the HTGR refabrication pilot plant has been established. The primary feature affecting the design is the high, time varying, gamma activity of the process material due to the unavoidable presence of uranium-232. This imposes stringent requirements for remote operation and remote maintainability of system components. At the same time, the remote operation lends itself to implementation of an automated data collection and processing system for real-time accountability. The high time-varying gamma activity of the material also precludes application of a number of techniques presently employed for light-water reactor fuel assay. The techniques selected for application in the refabrication facility are (1) active thermal neutron interrogation with fast-fission or delayed-neutron counting for fuel-rod and small-sample assay, (2) calorimetry for high-level waste assay, and (3) passive gamma scanning for low-level waste assay, and rapid on-line relative rod-loading measurements. The principal nondestructive assay subsystems are identified as (1) on-line devices for 100 percent product fuel rod assay and quality control, (2) a multipurpose device in the sample inspection laboratory for small- sample assay and secondary standards calibration, and (3) equipment for assay of high- and low-uranium content scrap and waste materials. A data processing system, which coordinates data from these subsystems with information from other process control sensors, is included to provide real-time material balance information. (U.S.)

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

    International Nuclear Information System (INIS)

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

    1974-01-01

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

  5. Calorimetric assay of HTGR fuel samples

    International Nuclear Information System (INIS)

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

    1979-04-01

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

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

    International Nuclear Information System (INIS)

    Lin, K.H.

    1976-01-01

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

  7. Quality control procedures for HTGR fuel element components

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  8. Safety aspects of solvent nitration in HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Wilbourn, R.G.

    1977-06-01

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

  9. Nuclear heat source design for an advanced HTGR process heat plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; O'Hanlon, T.W.

    1983-01-01

    A high-temperature gas-cooled reactor (HTGR) coupled with a chemical process facility could produce synthetic fuels (i.e., oil, gasoline, aviation fuel, methanol, hydrogen, etc.) in the long term using low-grade carbon sources (e.g., coal, oil shale, etc.). The ultimate high-temperature capability of an advanced HTGR variant is being studied for nuclear process heat. This paper discusses a process heat plant with a 2240-MW(t) nuclear heat source, a reactor outlet temperature of 950 0 C, and a direct reforming process. The nuclear heat source outputs principally hydrogen-rich synthesis gas that can be used as a feedstock for synthetic fuel production. This paper emphasizes the design of the nuclear heat source and discusses the major components and a deployment strategy to realize an advanced HTGR process heat plant concept

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  11. Information exchange on HTGR and nuclear hydrogen technology between JAEA and INET in 2008

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Tachibana, Yukio; Sun Yuliang

    2009-07-01

    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2008. (author)

  12. Information exchange on HTGR and nuclear hydrogen technology between JAEA and INET in 2009

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Wang Hong

    2010-07-01

    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2009. (author)

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

    International Nuclear Information System (INIS)

    Wallroth, C.F.

    1975-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-07-01

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

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

    International Nuclear Information System (INIS)

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

    1978-09-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  18. A 1500-MW(e) HTGR nuclear generating station

    International Nuclear Information System (INIS)

    Stinson, R.C.; Hornbuckle, J.D.; Wilson, W.H.

    1976-01-01

    A conceptual design of a 1500-MW(e) HTGR nuclear generating station is described. The design concept was developed under a three-party arrangement among General Atomic Company as nuclear steam supply system (NSSS) supplier, Bechtel Power Corporation as engineer-constructors of the balance of plant (BOP), and Southern California Edison Company as a potential utility user. A typical site in the lower Mojave Desert in southeastern California was assumed for the purpose of establishing the basic site criteria. Various alternative steam cycles, prestressed concrete reactor vessel (PCRV) and component arrangements, fuel-handling concepts, and BOP layouts were developed and investigated in a programme designed to lead to an economic plant design. The paper describes the NSSS and BOP designs, the general plant arrangement and a description of the site and its unique characteristics. The elements of the design are: the use of four steam generators that are twice the capacity of GA's steam generators for its 770-MW(e) and 1100-MW(e) units; the rearrangement of steam and feedwater piping and support within the PCRV; the elimination of the PCRV star foundation to reduce the overall height of the containment building as well as of the PCRV; a revised fuel-handling concept which permits the use of a simplified, grade-level fuel storage pool; a plant arrangement that permits a substantial reduction in the penetration structure around the containment while still minimizing the lengths of cable and piping runs; and the use of two tandem-compound turbine generators. Plant design bases are discussed, and events leading to the changes in concept from the reference 8-loop PCRV 1500-MW(e) HTGR unit are described. (author)

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

    International Nuclear Information System (INIS)

    Strand, J.B.

    1978-06-01

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

  20. 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

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

    International Nuclear Information System (INIS)

    Taswanda Taryo; Geni Rina Sunaryo; Ridwan; Meniek Rachmawati

    2018-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. Study on the HTGR axial fuel loading

    International Nuclear Information System (INIS)

    Tanaka, Ryokichi

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-10-01

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

  6. HTGR fuel behavior at very high temperature

    International Nuclear Information System (INIS)

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

    1986-03-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-10-01

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

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

    International Nuclear Information System (INIS)

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

    1977-08-01

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

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

    International Nuclear Information System (INIS)

    Rode, J.S.

    1976-10-01

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

  12. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  13. Optimization of MOX fuel cycles in pebble bed HTGR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  17. Status of reprocessing technology in the HTGR fuel cycle

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  18. CHAP: a composite nuclear plant simulation program applied to the 3000 MW(t) HTGR

    International Nuclear Information System (INIS)

    Secker, P.A.; Bailey, P.G.; Gilbert, J.S.; Willcutt, G.J.E. Jr.; Vigil, J.C.

    1977-01-01

    The Composite HTGR Analysis Program (CHAP) is a general systems analysis program which has been developed at LASL. The program is being used for simulating large HTGR nuclear power plant operation and accident transients. The general features and analytical methods of the CHAP program are discussed. Features of the large HTGR model and results of model transients are also presented

  19. HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

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

    1978-06-01

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

  20. Advances in HTGR spent fuel treatment technology

    International Nuclear Information System (INIS)

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

    1984-08-01

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

  1. Evaluation of a blender for HTGR fuel particles

    International Nuclear Information System (INIS)

    Johnson, D.R.

    1977-03-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  3. Automatic particle-size analysis of HTGR recycle fuel

    International Nuclear Information System (INIS)

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

    1977-09-01

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

  4. Experimental design for HTGR fuel rods

    International Nuclear Information System (INIS)

    Bayne, C.K.

    1975-01-01

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

  5. Device for sampling HTGR recycle fuel particles

    International Nuclear Information System (INIS)

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

    1977-03-01

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

  6. HTGR nuclear heat source component design and experience

    International Nuclear Information System (INIS)

    Peinado, C.O.; Wunderlich, R.G.; Simon, W.A.

    1982-05-01

    The high-temperature gas-cooled reactor (HTGR) nuclear heat source components have been under design and development since the mid-1950's. Two power plants have been designed, constructed, and operated: the Peach Bottom Atomic Power Station and the Fort St. Vrain Nuclear Generating Station. Recently, development has focused on the primary system components for a 2240-MW(t) steam cycle HTGR capable of generating about 900 MW(e) electric power or alternately producing high-grade steam and cogenerating electric power. These components include the steam generators, core auxiliary heat exchangers, primary and auxiliary circulators, reactor internals, and thermal barrier system. A discussion of the design and operating experience of these components is included

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

    International Nuclear Information System (INIS)

    Lindemer, T.B.

    1982-09-01

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

  8. Crushing strength of HTGR fuel particles

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    International Nuclear Information System (INIS)

    Bao Borong; Wang Gaodong; Qian Jun

    1992-05-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  11. Very small HTGR nuclear power plant concepts for special terrestrial applications

    International Nuclear Information System (INIS)

    McDonald, C.F.; Goodjohn, A.J.

    1983-01-01

    The role of the very small nuclear power plant, of a few megawatts capacity, is perceived to be for special applications where an energy source as required but the following prevail: 1) no indigenous fossil fuel source, in long transport distances that add substantially to the cost of oil, coal in gas, and 3) secure long-term power production for defense applications with freedom from fuel supply lines. A small High Temperature Gas-Cooled reactor (HTGR) plant could provide the total energy needs for 1) a military installation, 2) an island base of strategic significance, 3) an industrial community or 4) an urban area. The small HTGR is regarded as a fixed-base installation (as opposed to a mobile system). All of the major components would be factory fabricated and transported to the site where emphasis would be placed on minimizing the construction time. The very small HTGR plant, currently in an early stage of design definition, has the potential for meeting the unique needs of the small energy user in both the military and private sectors. The plant may find acceptance for specialized applications in the industrialized nations and to meet the energy needs of developing nations. Emphasis in the design has been placed on safety, simplicity and compactness

  12. Fission-product SiC reaction in HTGR fuel

    International Nuclear Information System (INIS)

    Montgomery, F.

    1981-01-01

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

  13. The choice of equipment mix and parameters for HTGR-based nuclear cogeneration plants

    Energy Technology Data Exchange (ETDEWEB)

    Malevski, A L; Stoliarevski, A Ya; Vladimirov, V T; Larin, E A; Lesnykh, V V; Naumov, Yu V; Fedotov, I L

    1990-07-01

    Improvement of heat and electricity supply systems based on cogeneration is one of the high-priority problems in energy development of the USSR. Fossil fuel consumption for heat supply exceeds now its use for electricity production and amounts to about 30% of the total demands. District heating provides about 80 million t.c.e. of energy resources conserved annually and meets about 50% of heat consumption of the country, including about 30% due to cogeneration. The share of natural gas and liquid fuel in the fuel consumption for district heating is about 70%. The analysis of heat consumption dynamics in individual regions and industrial-urban agglomerations shows the necessity of constructing cogeneration plants with the total capacity of about 60 million kW till the year 2000. However, their construction causes some serious problems. The most important of them are provision of environmentally clean fuels for cogeneration plants and provision of clear air. The limited reserves of oil and natural gas and the growing expenditures on their production require more intensive introduction of nuclear energy in the national energy balance. Possible use of nuclear energy based on light-water reactors for substitution of deficient hydrocarbon fuels is limited by the physical, technical and economic factors and requirements of safety. Further development of nuclear energy in the USSR can be realized on a new technological base with construction of domestic reactors of increased and ultimate safety. The most promising reactors under design are high-temperature gas-cooled reactors (HTGR) of low and medium capacity with the intrinsic property of safety. HTGR of low (about 200-250 MW(th) in a steel vessel), medium (about 500 MW(th) in a steel-concrete vessel) and high (about 1000-2500 MW(th) in a prestressed concrete vessel) are now designed and studied in the country. At outlet helium temperature of 920-1020 K it is possible to create steam turbine installations producing both

  14. The choice of equipment mix and parameters for HTGR-based nuclear cogeneration plants

    International Nuclear Information System (INIS)

    Malevski, A.L.; Stoliarevski, A.Ya.; Vladimirov, V.T.; Larin, E.A.; Lesnykh, V.V.; Naumov, Yu.V.; Fedotov, I.L.

    1990-01-01

    Improvement of heat and electricity supply systems based on cogeneration is one of the high-priority problems in energy development of the USSR. Fossil fuel consumption for heat supply exceeds now its use for electricity production and amounts to about 30% of the total demands. District heating provides about 80 million t.c.e. of energy resources conserved annually and meets about 50% of heat consumption of the country, including about 30% due to cogeneration. The share of natural gas and liquid fuel in the fuel consumption for district heating is about 70%. The analysis of heat consumption dynamics in individual regions and industrial-urban agglomerations shows the necessity of constructing cogeneration plants with the total capacity of about 60 million kW till the year 2000. However, their construction causes some serious problems. The most important of them are provision of environmentally clean fuels for cogeneration plants and provision of clear air. The limited reserves of oil and natural gas and the growing expenditures on their production require more intensive introduction of nuclear energy in the national energy balance. Possible use of nuclear energy based on light-water reactors for substitution of deficient hydrocarbon fuels is limited by the physical, technical and economic factors and requirements of safety. Further development of nuclear energy in the USSR can be realized on a new technological base with construction of domestic reactors of increased and ultimate safety. The most promising reactors under design are high-temperature gas-cooled reactors (HTGR) of low and medium capacity with the intrinsic property of safety. HTGR of low (about 200-250 MW(th) in a steel vessel), medium (about 500 MW(th) in a steel-concrete vessel) and high (about 1000-2500 MW(th) in a prestressed concrete vessel) are now designed and studied in the country. At outlet helium temperature of 920-1020 K it is possible to create steam turbine installations producing both

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-08-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    1976-06-01

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

  18. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-01-01

    Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

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

    International Nuclear Information System (INIS)

    Pierce, V.H.

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Tiegs, T.N.

    1982-03-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    1981-05-01

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

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

    International Nuclear Information System (INIS)

    1982-11-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-05-01

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

  6. Nuclear fuels

    International Nuclear Information System (INIS)

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

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

    International Nuclear Information System (INIS)

    Hedrick, C.E.

    1976-01-01

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

  8. On transient irradiation behavior of HTGR fuel particles

    International Nuclear Information System (INIS)

    Mortenson, S.C.; Okrent, D.

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-06-01

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

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

    International Nuclear Information System (INIS)

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

    1976-12-01

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

  12. Nuclear fuel

    International Nuclear Information System (INIS)

    D Hondt, P.

    1998-01-01

    The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

  13. Oxidation parameters of nuclear graphite for HTGR air-ingress

    International Nuclear Information System (INIS)

    Kim, E.S.; No, H.C.

    2004-01-01

    In order to investigate chemical behaviors of the graphite during an air-ingress accident in HTGR, the kinetic tests on nuclear graphite IG-110 were performed in chemical reaction dominant regime. In the present experiment, inlet gas flow rate ranged between 8 and 18 SLPM, graphite temperatures and oxygen mole fraction ranged from 540 to 630degC and from 3 to 30% respectively. The test section was made of a quartz tube having 75 mm diameter and 750 mm length and the test specimen machined to the size of 21 mm diameter and 30 mm length was supported at the center of it by the alumina rod. The 15 kW induction heater was installed around the outside of test section to heat the specimen and its temperature was measured by 2 infrared thermometers. The oxidation rate was calculated from the gas concentration analysis between inlet and outlet using NDIR (non-dispersive infrared) gas analyzer. As a result the activation energy (Ea) and the order of reaction (n) were determined within 95% confidence level and the qualitative characteristics of the two parameters were also widely investigated by experimental and analytical methods. (author)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    Llewellyn, G.H.

    1977-06-01

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

  16. Nuclear fuel reprocessing expansion strategies

    International Nuclear Information System (INIS)

    Gallagher, J.M.

    1975-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  18. Bibliographical survey of heat exchangers for nuclear power plants and problems of HTGR

    International Nuclear Information System (INIS)

    Yamao, Hiroyuki; Okamoto, Yoshizo; Sanokawa, Konomo

    1977-04-01

    The problems in development of heat exchangers for nuclear reactors have been examined in literature survey through Annual Index Subjects of NSA (Nuclear Science Abstracts) for the past ten years. R and D on heat exchangers for LMFBR, HTGR, LWR and HWR are on the increase. In the case of HTGRs, R and D on heat resisting materials including the corrosion and on hydrogen permeation of heat exchanger walls in high temperature pressure helium environment are important. Future R and D subjects for HTGR heat exchangers in showing the high temperature endurance are presented. (auth.)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-05-01

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

  1. Nuclear fuels

    International Nuclear Information System (INIS)

    2008-01-01

    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  2. Nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, H [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1976-10-01

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts.

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

    International Nuclear Information System (INIS)

    Fukaya, Y.; Goto, M.

    2017-01-01

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

  4. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  5. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  6. Nuclear fuel

    International Nuclear Information System (INIS)

    Azevedo, J.B.L. de.

    1980-01-01

    All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.) [pt

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

    International Nuclear Information System (INIS)

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

    2010-06-01

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

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

    International Nuclear Information System (INIS)

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

    1994-12-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  10. LABORATORY-SCALE PRODUCTION OF ADU GELS BY EXTERNAL GELATION FOR AN INTERMEDIATE HTGR NUCLEAR

    Directory of Open Access Journals (Sweden)

    S Simbolon

    2015-03-01

    Full Text Available LABORATORY-SCALE PRODUCTION OF ADU GELS BY EXTERNAL GELATION FOR AN INTERMEDIATE HTGR NUCLEAR. The The aim of this research is to produce thousands of microsphere ADU (Ammonium Diuranate gels by using external gelation for an intermediate HTGR (High Temperature Gas-cooled Reactor nuclear fuel in laboratory-scale. Microsphere ADU gels were based on sol-solution which was made from a homogeneous mixture of ADUN (Acid Deficient Uranyl Nitrate which was containing uranyl ion in high concentration, a water soluble organic compound PVA (Polyvinyl Alcohol and THFA (Tetrahydrofurfuryl Alcohol. The simple unified home made laboratory experimental machine was developed to replace test tube experiment method which was once used due to a tiny amount of microsphere ADU gels produced. It consists of four main parts: tank filled sol-solution connecting to peristaltic pump and vibrating nozzle, preliminary gelation and gelation column. The machine has successfully converted 150 mL sol-solution into thousands of drops which produced 120 - 130 drops in each minute in steady state in ammonia gas free sector. Preliminary gelation reaction was carried out in ammonia gas sector where drops react with ammonia gas in a bat an eye followed by gelation reaction in column containing ammonia solution 7 M. In ageing process, ADU gels were collected and submerged into a vessel containing ammonia solution which was shaken for 1 hour in a shaker device. Isopropyl alcohol (90% solution was used to wash ADU gels and a digital camera was used to measured spherical form of ADU gels. Diameters in spherical spheroid form were found between 1.8 mm until 2.2 mm. The spherical purity of ADU gels were 10% - 20% others were oblate, prolate spheroid and microsphere which have hugetiny of dimples on the surface.   PRODUKSI GEL ADU SKALA LABORATORIUM DENGAN MENGGUNAKAN GELASI EKSTERNAL UNTUK BAHAN BAKAR ANTARA HTGR. Penelitian ini bertujuan untuk membuat ribuan gel bulat ADU (Ammonium

  11. Significance and prospects of the energo-technological usage of HTGR for nuclear power development in the beginning of the XXI century

    International Nuclear Information System (INIS)

    Breger, A.Kh.; Putilov, A.V.; Bogoyavlensky, R.G.; Glebov, V.P.

    1993-01-01

    This report describes the economic efficiency of atomic stations (HTGR plants). The realization of the complex (energy radiation-technological) using of nuclear fuel leads to the economically effective mastering of nuclear energy sources instead of organic ones for supplying inductry and municipal economy. It is necessary to include in the power engineering development programme, under the circumstances of fulfillment of requirements of safety and reliability, research and development by the end of the century of the pattern of complex unit on the base of the HTGR with spherical fuel elements and (by 2010-15), mastering the energy-technological plants in high-energy branches of industry and municipal economy. Solving the mentioned problems will make a perceptible contribution into scientific progress, will allow to fulfill the conversion of war industry, attract highly qualified specialists to solving the tasks of national economy

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-10-15

    This paper describes temperature-dependent k-infinity measurements conducted using an assembly of loose HTGR coated particles in the BR-2 reactor by means of null reactivity oscillating method comparing the effect of poisoned and unpoisoned lattices like tests performed in the Physical Constants Test Reactor (PCTR) at Hanford. The RB-2 reactor was the property of the Italian firm AGIP NUCLEARE and operated at the Montecuccolino Center in Bologna.

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

    International Nuclear Information System (INIS)

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

    1978-12-01

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

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

    International Nuclear Information System (INIS)

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

    1977-02-01

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

  15. Proceedings of the 1st JAEA/KAERI information exchange meeting on HTGR and nuclear hydrogen technology

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Sakaba, Nariaki; Nishihara, Tetsuo; Yan, Xing L.; Hino, Ryutaro

    2007-03-01

    Japan Atomic Energy Agency (JAEA) has completed an implementation with Korea Atomic Energy Research Institute (KAERI) on HTGR and nuclear hydrogen technology, 'The Implementation of Cooperative Program in the Field of Peaceful Uses of Nuclear Energy between KAERI and JAEA. 'To facilitate efficient technology development on HTGR and nuclear hydrogen by the IS process, an information exchange meeting was held at the Oarai Research and Development Center of JAEA on August 28-30, 2006 under Program 13th of the JAEA/KAERI Implementation, 'Development of HTGR and Nuclear Hydrogen Technology'. JAEA and KAERI mutually showed the status and future plan of the HTTR (High-Temperature Engineering Test Reactor) project in Japan and of the NHDD (Nuclear Hydrogen Development and Demonstration) project in Korea, respectively, and discussed collaboration items. This proceedings summarizes all materials of presented technical discussions on HTGR and hydrogen production technology as well as the meeting briefing including collaboration items. (author)

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

    International Nuclear Information System (INIS)

    Verfondern, K.; Nabielek, H.

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Jiao Rongzhou; He Peijun; Liu Bingren; Zhu Yongjun

    1992-08-01

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

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

    International Nuclear Information System (INIS)

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

    1978-02-01

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

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Weicht, U.; Mueller, W.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1976-04-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-05-01

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

  5. Nuclear closed-cycle gas turbine (HTGR-GT): dry cooled commercial power plant studies

    International Nuclear Information System (INIS)

    McDonald, C.F.; Boland, C.R.

    1979-11-01

    Combining the modern and proven power conversion system of the closed-cycle gas turbine (CCGT) with an advanced high-temperature gas-cooled reactor (HTGR) results in a power plant well suited to projected utility needs into the 21st century. The gas turbine HTGR (HTGR-GT) power plant benefits are consistent with national energy goals, and the high power conversion efficiency potential satisfies increasingly important resource conservation demands. Established technology bases for the HTGR-GT are outlined, together with the extensive design and development program necessary to commercialize the nuclear CCGT plant for utility service in the 1990s. This paper outlines the most recent design studies by General Atomic for a dry-cooled commercial plant of 800 to 1200 MW(e) power, based on both non-intercooled and intercooled cycles, and discusses various primary system aspects. Details are given of the reactor turbine system (RTS) and on integrating the major power conversion components in the prestressed concrete reactor vessel

  6. Design and main characteristics of HTGR fuel elements

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  7. Development of a fissile particle for HTGR fuel recycle

    International Nuclear Information System (INIS)

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

    1976-12-01

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

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

    International Nuclear Information System (INIS)

    Young, D.T.

    1977-03-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Carpenter, J.A. Jr.

    1977-11-01

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

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

    International Nuclear Information System (INIS)

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

    1979-04-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-06-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-05-01

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

  16. Formation of actinides in irradiated HTGR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    dos Santos, A. M.

    1976-03-15

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-05-01

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

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

    International Nuclear Information System (INIS)

    Haas, P.A.

    1975-09-01

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

  20. Formation of actinides in irradiated HTGR fuel elements

    International Nuclear Information System (INIS)

    Santos, A.M. dos.

    1976-03-01

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

  1. Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site

    International Nuclear Information System (INIS)

    Demick, L.E.

    2011-01-01

    This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

  2. Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2011-10-01

    This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

  3. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

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

    International Nuclear Information System (INIS)

    Lee, D.A.

    1979-02-01

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

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

    International Nuclear Information System (INIS)

    Ugajin, M.; Shiba, K.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-11-01

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

  7. Safety analysis of coupling system of hybrid (MED-RO) nuclear desalination system utilising waste heat from HTGR

    International Nuclear Information System (INIS)

    Raha, Abhijit; Kishore, G.; Rao, I.S.; Adak, A.K.; Srivastava, V.K.; Prabhakar, S.; Tewari, P.K.

    2010-01-01

    To meet the generation IV goals, High Temperature Gas Cooled Reactors (HTGRs) are designed to have relatively higher thermal efficiency and enhanced safety and environmental characteristics. It can provide energy for combined production of hydrogen, electricity and other industrial applications. The waste heat available in the HTGR power cycle can also be utilized for the desalination of seawater for producing potable water. Desalination is an energy intensive process, so use of waste heat from HTGR certainly makes desalination process more affordable to create fresh water resources. So design of the coupling system, as per the safety design requirement of nuclear desalination plant, of desalination plant with HTGR is very crucial. In the first part of this paper, design of the coupling system between hybrid Multi Effect Desalination-Reverse Osmosis (MED-RO) nuclear desalination plant and HTGR to utilize the waste heat in HTGR are discussed. In the next part deterministic safety analysis of the designed coupling system of are presented in detail. It was found that all the coupling system meets the acceptance criteria for all the Postulated Initiating Events (PIE's) limited to DBA. (author)

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

    International Nuclear Information System (INIS)

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

    1976-02-01

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

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

    International Nuclear Information System (INIS)

    Bullock, R.E.

    1977-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

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

    International Nuclear Information System (INIS)

    1980-08-01

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

  12. Utilization of Plutonium and Higher Actinides in the HTGR as Possibility to Maintain Long-Term Operation on One Fuel Loading

    International Nuclear Information System (INIS)

    Tsvetkova, Galina V.; Peddicord, Kenneth L.

    2002-01-01

    Promising existing nuclear reactor concepts together with new ideas are being discussed worldwide. Many new studies are underway in order to identify prototypes that will be analyzed and developed further as systems for Generation IV. The focus is on designs demonstrating full inherent safety, competitive economics and proliferation resistance. The work discussed here is centered on a modularized small-size High Temperature Gas-cooled Reactor (HTGR) concept. This paper discusses the possibility of maintaining long-term operation on one fuel loading through utilization of plutonium and higher actinides in the small-size pebble-bed reactor (PBR). Acknowledging the well-known flexibility of the PBR design with respect to fuel composition, the principal limitations of the long-term burning of plutonium and higher actinides are considered. The technological challenges and further research are outlined. The results allow the identification of physical features of the PBR that significantly influence flexibility of the design and its applications. (authors)

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

    International Nuclear Information System (INIS)

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

    2008-03-01

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

  14. Nuclear fuel

    International Nuclear Information System (INIS)

    Quinauk, J.P.

    1990-01-01

    Since 1985, Fragema has been marketing and selling the Advanced Fuel Assemby AFA whose main features are its zircaloy grids and removable top and bottom nozzles. It is this product, which exists for several different fuel assembly arrays and heights, that will be employed in the reactors at Daya Bay. Fragema employs gadolinium as the consumable poison to enable highperformance fuel management. More recently, the company has supplied fuel assemblies of the mixed-oxide(MOX) and enriched reprocessed uranium type. The reliability level of the fuel sold by Fragema is one of the highest in the world, thanks in particular to the excellence of the quality assurance and quality control programs that have been implemented at all stages of its design and manufacture

  15. Conceptual design of primary coolant purification system using cylindrical membrane for nuclear energy system base on HTGR

    International Nuclear Information System (INIS)

    Piping Supriatna

    2011-01-01

    The recent progress of reactor technology design for next generation reactor will be implemented on cogeneration reactor, which the aim of reactor operation not only for generating electrical energy, but also for other application like desalination, industrial manufacturing process, hydrogen production, Enhanced Oil Recovery (EOR), etc. The cogeneration reactor concept developed for generate energy effectively, efficiently and sustainable, which reserve of uranium and thorium nuclear fuel for cogeneration reactor is supply able for world energy demand until next thousand years. The cogeneration reactor produce temperature output higher than commonly Nuclear Power Plant (NPP), and need special Heat Exchanger with helium gas as coolant. In order to preserve heat transfer with high efficiency, constant purity of the gas must be maintained as well as possible, especially contamination from its impurities. In this research has been designed modeling and assessment of primary coolant gas purification system with purify and fill up helium gas continuously, by using Cylindrical Helium Splitting Membrane and helium gas inventory system. The result of flow rate helium assessment for the purification system is 0.844x10 -3 kg/sec, where helium flow rate of reactor primary coolant is 120 kg/sec. The result of study show that the Primary Coolant Gas Purification System is enable to be implemented on Cogeneration Reactor HTGR200C. (author)

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

    International Nuclear Information System (INIS)

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

    1980-02-01

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

  17. Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project - Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Saurwein, John

    2011-07-15

    This report is the Final Technical Report for the Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project conducted by a team led by General Atomics under DOE Award DE-NE0000245. The primary overall objective of the project was to develop and document a conceptual design for the Steam Cycle Modular Helium Reactor (SC-MHR), which is the reactor concept proposed by General Atomics for the NGNP Demonstration Plant. The report summarizes the project activities over the entire funding period, compares the accomplishments with the goals and objectives of the project, and discusses the benefits of the work. The report provides complete listings of the products developed under the award and the key documents delivered to the DOE.

  18. Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project - Final Technical Report

    International Nuclear Information System (INIS)

    Saurwein, J.

    2011-01-01

    This report is the Final Technical Report for the Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project conducted by a team led by General Atomics under DOE Award DE-NE0000245. The primary overall objective of the project was to develop and document a conceptual design for the Steam Cycle Modular Helium Reactor (SC-MHR), which is the reactor concept proposed by General Atomics for the NGNP Demonstration Plant. The report summarizes the project activities over the entire funding period, compares the accomplishments with the goals and objectives of the project, and discusses the benefits of the work. The report provides complete listings of the products developed under the award and the key documents delivered to the DOE.

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

    International Nuclear Information System (INIS)

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

    1984-05-01

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

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

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

    International Nuclear Information System (INIS)

    Han Zandong; Luo Sheng; Liu Jiguo

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-11-01

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

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

    International Nuclear Information System (INIS)

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

    2002-05-01

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

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

    International Nuclear Information System (INIS)

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

    2001-05-01

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

  5. USNRC HTGR safety research program overview

    International Nuclear Information System (INIS)

    Foulds, R.B.

    1982-01-01

    An overview is given of current activities and planned research efforts of the US Nuclear Regulatory Commission (NRC) HTGR Safety Program. On-going research at Brookhaven National Laboratory, Oak Ridge National Laboratory, Los Alamos National Laboratory, and Pacific Northwest Laboratory are outlined. Tables include: HTGR Safety Issues, Program Tasks, HTGR Computer Code Library, and Milestones for Long Range Research Plan

  6. HTGR analytical methods and design verification

    International Nuclear Information System (INIS)

    Neylan, A.J.; Northup, T.E.

    1982-05-01

    Analytical methods for the high-temperature gas-cooled reactor (HTGR) include development, update, verification, documentation, and maintenance of all computer codes for HTGR design and analysis. This paper presents selected nuclear, structural mechanics, seismic, and systems analytical methods related to the HTGR core. This paper also reviews design verification tests in the reactor core, reactor internals, steam generator, and thermal barrier

  7. Nuclear fuel preheating system

    International Nuclear Information System (INIS)

    Andrea, C.

    1975-01-01

    A nuclear reactor new fuel handling system which conveys new fuel from a fuel preparation room into the reactor containment boundary is described. The handling system is provided with a fuel preheating station which is adaptd to heat the new fuel to reactor refueling temperatures in such a way that the fuel is heated from the top down so that fuel element cladding failure due to thermal expansions is avoided. (U.S.)

  8. Romanian nuclear fuel program

    International Nuclear Information System (INIS)

    Budan, O.

    1999-01-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

  9. Nuclear fuel lease accounting

    International Nuclear Information System (INIS)

    Danielson, A.H.

    1986-01-01

    The subject of nuclear fuel lease accounting is a controversial one that has received much attention over the years. This has occurred during a period when increasing numbers of utilities, seeking alternatives to traditional financing methods, have turned to leasing their nuclear fuel inventories. The purpose of this paper is to examine the current accounting treatment of nuclear fuel leases as prescribed by the Financial Accounting Standards Board (FASB) and the Federal Energy Regulatory Commission's (FERC's) Uniform System of Accounts. Cost accounting for leased nuclear fuel during the fuel cycle is also discussed

  10. The Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    2011-08-01

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

  11. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Nakai, Keiichi

    1983-01-01

    Purpose: To decrease the tensile stresses resulted in a fuel can as well as prevent decladding of fuel pellets into the bore holes by decreasing the inner pressure within the nuclear fuel element. Constitution: A fuel can is filled with hollow fuel pellets, inserted with a spring for retaining the hollow fuel pellets with an appropriate force and, thereafter, closely sealed at the both ends with end plugs. A cylindrical body is disposed into the bore holes of the hollow fuel pellets. Since initial sealing gases and/or gaseous nuclear fission products can thus be excluded from the bore holes where the temperature is at the highest level, the inner pressure of the nuclear fuel element can be reduced to decrease the tensile strength resulted to the fuel can. Furthermore, decladding of fuel pellets into the bore holes can be prevented. (Moriyama, K.)

  12. Nuclear fuel replacement device

    International Nuclear Information System (INIS)

    Ritz, W.C.; Robey, R.M.; Wett, J.F.

    1984-01-01

    A fuel handling arrangement for a liquid metal cooled nuclear reactor having a single rotating plug eccentric to the fuel core and a fuel handling machine radially movable along a slot in the plug with a transfer station disposed outside the fuel core but covered by the eccentric plug and within range of movement of said fuel handling machine to permit transfer of fuel assemblies between the core and the transfer station. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-08

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

  14. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1981-01-01

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

  15. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

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

  16. Nuclear fuel accounting

    International Nuclear Information System (INIS)

    Aisch, D.E.

    1977-01-01

    After a nuclear power plant has started commercial operation the actual nuclear fuel costs have to be demonstrated in the rate making procedure. For this purpose an accounting system has to be developed which comprises the following features: 1) All costs associated with nuclear fuel shall be correctly recorded; 2) it shall be sufficiently flexible to cover also deviations from proposed core loading patterns; 3) it shall be applicable to different fuel cycle schemes. (orig./RW) [de

  17. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    1998-05-01

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

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

    International Nuclear Information System (INIS)

    Krohn, H.; Finken, R.

    1983-06-01

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

  19. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  20. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  1. Future perspective of nuclear energy utilization and expected role of HTGR. JAERI's energy systems analysis research

    International Nuclear Information System (INIS)

    Sato, Osamu

    1996-01-01

    Studies have been made in JAERI in order to assess the possibility of using nuclear energy symbiotically with fossil and biomass fuels, and to evaluate its implications for the environment. The application system of high temperature nuclear heat has been designed for this purpose with various technology options. The core of the system is a set of technologies for hydrogen production and its application to produce clean and convenient fuels from fossil or biomass sources. The results of analytical studies using the MARKAL model have indicated sufficient possibilities of combining nuclear energy effectively with fossil or biomass fuels via hydrogen produced by high temperature nuclear heat. In addition to providing clean and convenient liquid fuels on a large scale, the combined system will contribute to the substantial reduction of long-term CO 2 emissions. The relatively high cost of this system will be well justified when CO 2 emission penalties are taken into account. (J.P.N.)

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

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  3. Nuclear fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding, and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  4. Nuclear fuel element

    International Nuclear Information System (INIS)

    Thompson, J.R.; Rowland, T.C.

    1976-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  5. Nuclear fuel production

    International Nuclear Information System (INIS)

    Randol, A.G.

    1985-01-01

    The production of new fuel for a power plant reactor and its disposition following discharge from the power plant is usually referred to as the ''nuclear fuel cycle.'' The processing of fuel is cyclic in nature since sometime during a power plant's operation old or ''depleted'' fuel must be removed and new fuel inserted. For light water reactors this step typically occurs once every 12-18 months. Since the time required for mining of the raw ore to recovery of reusable fuel materials from discharged materials can span up to 8 years, the management of fuel to assure continuous power plant operation requires simultaneous handling of various aspects of several fuel cycles, for example, material is being mined for fuel to be inserted in a power plant 2 years into the future at the same time fuel is being reprocessed from a discharge 5 years prior. Important aspects of each step in the fuel production process are discussed

  6. Nuclear fuel element

    International Nuclear Information System (INIS)

    Mogard, J.H.

    1977-01-01

    A nuclear fuel element is disclosed for use in power producing nuclear reactors, comprising a plurality of axially aligned ceramic cylindrical fuel bodies of the sintered type, and a cladding tube of metal or metal alloys, wherein said cladding tube on its cylindrical inner surface is provided with a plurality of slightly protruding spacing elements distributed over said inner surface

  7. National HTGR safety program

    International Nuclear Information System (INIS)

    Davis, D.E.; Kelley, A.P. Jr.

    1982-01-01

    This paper presents an overview of the National HTGR Program in the US with emphasis on the safety and licensing strategy being pursued. This strategy centers upon the development of an integrated approach to organizing and classifying the functions needed to produce safe and economical nuclear power production. At the highest level, four plant goals are defined - Normal Operation, Core and Plant Protection, Containment Integrity and Emergency Preparedness. The HTGR features which support the attainment of each goal are described and finally a brief summary is provided of the current status of the principal safety development program supporting the validation of the four plant goals

  8. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-01-01

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

  9. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sasaki, Y.; Tashima, J.

    1975-01-01

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

  10. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1980-01-01

    A bimetallic spacer means is cooperatively associated with a nuclear fuel assembly and operative to resist the occurrence of in-reactor bowing of the nuclear fuel assembly. The bimetallic spacer means in one embodiment of the invention includes a space grid formed, at least principally, of zircaloy to the external surface of which are attached a plurality of stainless steel strips. In another embodiment the strips are attached to fuel pins. In each of the embodiments, the stainless steel strips during power production expand outwardly to a greater extent than do the members to which the stainless steel strips are attached, thereby forming stiff springs which abut against like bimetallic spacer means with which the other nuclear fuel assemblies are provided in a given nuclear reactor core to thus prevent the occurrence of in-reactor bowing of the nuclear fuel assemblies. (author)

  11. Nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Matsumoto, Takashi; Isaka, Shinji.

    1987-01-01

    Purpose: To increase the spent fuel storage capacity and reduce the installation cost in a nuclear fuel storage facility. Constitution: Fuels handled in the nuclear fuel storage device of the present invention include the following four types: (1) fresh fuels, (2) 100 % reactor core charged fuels, (3) spent fuels just after taking out and (4) fuels after a certain period (for example one half-year) from taking out of the reactor. Reactivity is high for the fuels (1), and some of fuels (2), while low in the fuels (3) (4), Source intensity is strong for the fuels (3) and some of the fuels (2), while it is low for the fuels (1) and (4). Taking notice of the fact that the reactivity, radioactive source intensity and generated after heat are different in the respective fuels, the size of the pool and the storage capacity are increased by the divided storage control. While on the other hand, since the division is made in one identical pool, the control method becomes important, and the working range is restricted by means of a template, interlock, etc., the operation mode of the handling machine is divided into four, etc. for preventing errors. (Kamimura, M.)

  12. Study on commercial HTGR hydrogen production system

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo

    2000-07-01

    The Japanese energy demand in 2030 will increase up to 117% in comparison with one in 2000. We have to avoid a large consumption of fossil fuel that induces a large CO 2 emission from viewpoint of global warming. Furthermore new energy resources expected to resolve global warming have difficulty to be introduced more because of their low energy density. As a result, nuclear power still has a possibility of large introduction to meet the increasing energy demand. On the other hand, in Japan, 40% of fossil fuels in the primary energy are utilized for power generation, and the remaining are utilized as a heat source. New clean energy is required to reduce the consumption of fossil fuels and hydrogen is expected as a alternative energy resource. Prediction of potential hydrogen demand in Japan is carried out and it is clarified that the demand will potentially increase up to 4% of total primary energy in 2050. In present, steam reforming method is the most economical among hydrogen generation processes and the cost of hydrogen production is about 7 to 8 yen/m 3 in Europe and the United States and about 13 yen/m 3 in Japan. JAERI has proposed for using the HTGR whose maximum core outlet temperature is at 950degC as a heat source in the steam reforming to reduced the consumption of fossil fuels and resulting CO 2 emission. Based on the survey of the production rate and the required thermal energy in conventional industry, it is clarified that a hydrogen production system by the steam reforming is the best process for the commercial HTGR nuclear heat utilization. The HTGR steam reforming system and other candidate nuclear heat utilization systems are considered from viewpoint of system layout and economy. From the results, the hydrogen production cost in the HTGR stream reforming system is expected to be about 13.5 yen/m 3 if the cost of nuclear heat of the HTGR is the same as one of the LWR. (author)

  13. Waste management considerations in HTGR recycle operations

    International Nuclear Information System (INIS)

    Pence, D.T.; Shefcik, J.J.; Heath, C.A.

    1975-01-01

    Waste management considerations in the recycle of HTGR fuel are different from those encountered in the recycle of LWR fuel. The types of waste associated with HTGR recycle operations are discussed, and treatment methods for some of the wastes are described

  14. Nuclear fuel activities in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Bairiot, H

    1997-12-01

    In his presentation on nuclear fuel activities in belgium the author considers the following directions of this work: fuel fabrication, NPP operation, fuel performance, research and development programmes.

  15. Boosting nuclear fuels

    International Nuclear Information System (INIS)

    Demarthon, F.; Donnars, O.; Dupuy-Maury, F.

    2002-01-01

    This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UO x and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)

  16. STAT, GAPS, STRAIN, DRWDIM: a system of computer codes for analyzing HTGR fuel test element metrology data. User's manual

    Energy Technology Data Exchange (ETDEWEB)

    Saurwein, J.J.

    1977-08-01

    A system of computer codes has been developed to statistically reduce Peach Bottom fuel test element metrology data and to compare the material strains and fuel rod-fuel hole gaps computed from these data with HTGR design code predictions. The codes included in this system are STAT, STRAIN, GAPS, and DRWDIM. STAT statistically evaluates test element metrology data yielding fuel rod, fuel body, and sleeve irradiation-induced strains; fuel rod anisotropy; and additional data characterizing each analyzed fuel element. STRAIN compares test element fuel rod and fuel body irradiation-induced strains computed from metrology data with the corresponding design code predictions. GAPS compares test element fuel rod, fuel hole heat transfer gaps computed from metrology data with the corresponding design code predictions. DRWDIM plots the measured and predicted gaps and strains. Although specifically developed to expedite the analysis of Peach Bottom fuel test elements, this system can be applied, without extensive modification, to the analysis of Fort St. Vrain or other HTGR-type fuel test elements.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-04-15

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

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  19. Improved nuclear fuel element

    International Nuclear Information System (INIS)

    Klepfer, H.H.

    1974-01-01

    A nuclear fuel element is described which comprises: 1) an elongated clad container, 2) a layer of high lubricity material being disposed in and adjacent to the clad container, 3) a low neutron capture cross section metal liner being disposed in the clad container and adjacent to the layer, 4) a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, 5) an enclosure integrally secured and sealed at each end of the container, and a nuclear fuel material retaining means positioned in the cavity. (author)

  20. Nuclear Fuel Reprocessing

    International Nuclear Information System (INIS)

    Simpson, Michael F.; Law, Jack D.

    2010-01-01

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

  1. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2005-01-01

    When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

  2. Nuclear fuel element

    International Nuclear Information System (INIS)

    Yamamoto, Seigoro.

    1994-01-01

    Ultrafine particles of a thermal neutron absorber showing ultraplasticity is dispersed in oxide ceramic fuels by more than 1% to 10% or lower. The ultrafine particles of the thermal neutron absorber showing ultrafine plasticity is selected from any one of ZrGd, HfEu, HfY, HfGd, ZrEu, and ZrY. The thermal neutron absorber is converted into ultrafine particles and solid-solubilized in a nuclear fuel pellet, so that the dispersion thereof into nuclear fuels is made uniform and an absorbing performance of the thermal neutrons is also made uniform. Moreover, the characteristics thereof, for example, physical properties such as expansion coefficient and thermal conductivity of the nuclear fuels are also improved. The neutron absorber, such as ZrGd or the like, can provide plasticity of nuclear fuels, if it is mixed into the nuclear fuels for showing the plasticity. The nuclear fuel pellets are deformed like an hour glass as burning, but, since the end portion thereof is deformed plastically within a range of a repulsive force of the cladding tube, there is no worry of damaging a portion of the cladding tube. (N.H.)

  3. Transportation of nuclear fuel

    International Nuclear Information System (INIS)

    Prowse, D.R.

    1979-01-01

    Shipment of used fuel from nuclear reactors to a central fuel management facility is discussed with particular emphasis on the assessment of the risk to the public due to these shipments. The methods of transporting used fuel in large shipping containers is reviewed. In terms of an accident scenario, it is demonstrated that the primary risk of transport of used fuel is due to injury and death in common road accidents. The radiological nature of the used fuel cargo is, for all practical purposes, an insignificant factor in the total risk to the public. (author)

  4. Nuclear fuel banks

    International Nuclear Information System (INIS)

    Anon.

    2010-01-01

    In december 2010 IAEA gave its agreement for the creation of a nuclear fuel bank. This bank will allow IAEA to help member countries that renounce to their own uranium enrichment capacities. This bank located on one or several member countries will belong to IAEA and will be managed by IAEA and its reserve of low enriched uranium will be sufficient to fabricate the fuel for the first load of a 1000 MW PWR. Fund raising has been successful and the running of the bank will have no financial impact on the regular budget of the IAEA. Russia has announced the creation of the first nuclear fuel bank. This bank will be located on the Angarsk site (Siberia) and will be managed by IAEA and will own 120 tonnes of low-enriched uranium fuel (between 2 and 4.95%), this kind of fuel is used in most Russian nuclear power plants. (A.C.)

  5. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

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

  6. Nuclear fuel waste disposal

    International Nuclear Information System (INIS)

    Merrett, G.J.; Gillespie, P.A.

    1983-07-01

    This report discusses events and processes that could adversely affect the long-term stability of a nuclear fuel waste disposal vault or the regions of the geosphere and the biosphere to which radionuclides might migrate from such a vault

  7. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Patarin, L.

    2002-01-01

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

  8. Nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    White, D.

    1981-01-01

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

  9. Information exchange mainly on HTGR operation and maintenance technique between JAEA and INET in 2005

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Hino, Ryutaro; Yu Suyuan

    2006-06-01

    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation with emphasis on HTGR operation and maintenance techniques between JAEA and INET and outlines cooperation activities during the fiscal year 2005. (author)

  10. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Hayashi, Hiroshi; Watari, Yoshio; Hizahara, Hiroshi; Masuoka, Ryuzo.

    1970-01-01

    When exchanging nuclear fuel assemblies during the operation of a nuclear reactor, melting of fuel bodies, and severence of tubular claddings is halted at the time of insertion by furnishing a neutron absorbing material such as B 10 , Cd, Gd or the like at the forward end of the fuel assembly to thereby lower the power peak at the forward ends of the fuel elements to within tolerable levels and thus prevent both fuel liquification and excessive expansion. The neutron absorbing material may be attached in the form of a plate to the fuel assembly forward tie plate, or may be inserted as a pellet into the front end of the tubular cladding. (Owens, K.J.)

  11. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Ainsworth, K.F.

    1979-01-01

    A nuclear fuel element is described having a cluster of nuclear fuel pins supported in parallel, spaced apart relationship by transverse cellular braces within coaxial, inner and outer sleeves, the inner sleeve being in at least two separate axial lengths, each of the transverse braces having a peripheral portion which is clamped peripherally between the ends of the axial lengths of the inner sleeve. (author)

  12. Nuclear fuel manufacture

    International Nuclear Information System (INIS)

    Costello, J.M.

    1980-09-01

    The technologies used to manufacture nuclear fuel from uranium ore are outlined, with particular reference to the light water reactor fuel cycle. Capital and operating cost estimates for the processing stages are given, and the relevance to a developing uranium industry in Australia is discussed

  13. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Butterfield, C.E.; Waite, E.

    1982-01-01

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

  14. Nuclear fuel pellet loading apparatus

    International Nuclear Information System (INIS)

    Gerkey, K.S.

    1979-01-01

    An automatic apparatus for loading a predetermined amount of nuclear fuel pellets into a nuclear fuel element to be used in a nuclear reactor is described. The apparatus consists of a vibratory bed capable of supporting corrugated trays containing rows of nuclear fuel pellets and arranged in alignment with the open ends of several nuclear fuel elements. A sweep mechanism is arranged above the trays and serves to sweep the rows of fuel pellets onto the vibratory bed and into the fuel element. A length detecting system, in conjunction with a pellet stopping mechanism, is also provided to assure that a predetermined amount of nuclear fuel pellets are loaded into each fuel element

  15. Nuclear Fuel Cycle Objectives

    International Nuclear Information System (INIS)

    2013-01-01

    . The four Objectives publications include Nuclear General Objectives, Nuclear Power Objectives, Nuclear Fuel Cycle Objectives, and Radioactive Waste management and Decommissioning Objectives. This publication sets out the objectives that need to be achieved in the area of the nuclear fuel cycle to ensure that the Nuclear Energy Basic Principles are satisfied. Within each of these four Objectives publications, the individual topics that make up each area are addressed. The five topics included in this publication are: resources; fuel engineering and performance; spent fuel management and reprocessing; fuel cycles; and the research reactor nuclear fuel cycle

  16. Milling Behavior of Matrix Graphite Powders with Different Binder Materials in HTGR Fuel Element Fabrication: I. Variation in Particle Size Distribution

    International Nuclear Information System (INIS)

    Lee, Young Woo; Cho, Moon Sung

    2011-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a matrix graphite powder properly prepared and pressed into a spherical shape or a cylindrical compact finally heat-treated at about 1900 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, overcoating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. In order to develop a fuel compact fabrication technology, it is important to develop a technology to prepare the matrix graphite powder (MGP) with proper characteristics, which has a strong influence on further steps and the material properties of fuel element. In this work, the milling behavior of matrix graphite powder mixture with different binder materials and their contents was investigated by analyzing the change in particle size distribution with different milling time

  17. Nuclear power and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Hardy, C.J.; Silver, J.M.

    1985-09-01

    The report provides data and assessments of the status and prospects of nuclear power and the nuclear fuel cycle. The report discusses the economic competitiveness of nuclear electricity generation, the extent of world uranium resources, production and requirements, uranium conversion and enrichment, fuel fabrication, spent fuel treatment and radioactive waste management. A review is given of the status of nuclear fusion research

  18. HTGR safety philosophy

    Energy Technology Data Exchange (ETDEWEB)

    Joksimovic, V.; Fisher, C. R. [General Atomic Co., San Diego, CA (USA)

    1981-01-15

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity.

  19. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joksimovic, V.; Fisher, C.R.

    1981-01-01

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity. (author)

  20. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joskimovic, V.; Fisher, C.R.

    1980-08-01

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the US. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity

  1. Improved nuclear fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a metal liner disposed between the cladding and the nuclear fuel material and a high lubricity material in the form of a coating disposed between the liner and the cladding. The liner preferably has a thickness greater than the longest fission product recoil distance and is composed of a low neutron capture cross-section material. The liner is preferably composed of zirconium, an alloy of zirconium, niobium or an alloy of niobium. The liner serves as a preferential reaction site for volatile impurities and fission products and protects the cladding from contact and reaction with such impurities and fission products. The high lubricity material acts as an interface between the liner and the cladding and reduces localized stresses on the cladding due to fuel expansion and cracking of the fuel

  2. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Wakamatsu, Mitsuo.

    1974-01-01

    Object: To improve a circulating flow passage of coolant so as to be able to accurately detect the temperature of coolant, rare gases contained, and the like. Structure: A fuel assembly comprising a flow regulating lattice provided with a plurality of communication holes in an axial direction, said lattice being positioned at the upper end of an outer tube in which nuclear fuel elements are received, and a neutron shielding body having a plurality of spiral coolant flow passages disposed between the lattice and the nuclear fuel elements, whereby a coolant comprised of liquid sodium or the like, which moves up passing through the coolant flow passages and the flow regulating passage, is regulated and passed through a detector mounted at the upper part of the flow regulating lattice to detect coolant temperature, flow rate, and rare gases or the like as the origin of nuclear fission contained in the coolant due to breakage of fuel elements. (Kamimura, M.)

  3. Nuclear fuel quality assurance

    International Nuclear Information System (INIS)

    1976-01-01

    Full text: Quality assurance is used extensively in the design, construction and operation of nuclear power plants. This methodology is applied to all activities affecting the quality of a nuclear power plant in order to obtain confidence that an item or a facility will perform satisfactorily in service. Although the achievement of quality is the responsibility of all parties participating in a nuclear power project, establishment and implementation of the quality assurance programme for the whole plant is a main responsibility of the plant owner. For the plant owner, the main concern is to achieve control over the quality of purchased products or services through contractual arrangements with the vendors. In the case of purchase of nuclear fuel, the application of quality assurance might be faced with several difficulties because of the lack of standardization in nuclear fuel and the proprietary information of the fuel manufacturers on fuel design specifications and fuel manufacturing procedures. The problems of quality assurance for purchase of nuclear fuel were discussed in detail during the seminar. Due to the lack of generally acceptable standards, the successful application of the quality assurance concept to the procurement of fuel depends on how much information can be provided by the fuel manufacturer to the utility which is purchasing fuel, and in what form and how early this information can be provided. The extent of information transfer is basically set out in the individual vendor-utility contracts, with some indirect influence from the requirements of regulatory bodies. Any conflict that exists appears to come from utilities which desire more extensive control over the product they are buying. There is a reluctance on the part of vendors to permit close insight of the purchasers into their design and manufacturing procedures, but there nevertheless seems to be an increasing trend towards release of more information to the purchasers. It appears that

  4. Development of a surveillance robot for dimensional and visual inspection of fuel and reflector elements from the Fort St. Vrain HTGR

    International Nuclear Information System (INIS)

    Wallroth, C.F.; Marsh, N.I.; Miller, C.M.; Saurwein, J.J.; Smith, T.L.

    1979-11-01

    A robotic device has been developed for dimensional and visual inspection of irradiated HTGR core components. The robot consists of a rotary table and a two-finger probe, driven by stepping motors, and four remotely controlled television cameras. Automated operation is accomplished via minicomputer control. A total of 51 irradiated fuel and reflector elements were inspected at a fraction of the time and cost required for conventional methods

  5. Fractional release of short-lived noble gases and iodine from HTGR fuel compact containing a fraction of coated fuel particles with through-coating defects

    International Nuclear Information System (INIS)

    Ogawa, Toru; Fukuda, Kosaku; Kobayashi, Fumiaki; Kikuchi, Teruo; Tobita, Tsutomu; Kashimura, Satoru; Kikuchi, Hironobu; Yamamoto, Katsumune.

    1986-10-01

    Fractional release (R/B) data of short-lived noble gases and iodine from sweep-gas irradiated HTGR fuel compacts were analyzed. Empirical formulas to predict R/B of 88 Kr as a function of temperature and fraction through-coating defects, and to calculate ratios of R/B's of other shortlived gases to that of 88 Kr were proposed. A method to predict R/B of iodine was also proposed. As for 131 I, a fission product of major safety concern, (R/B) I 131 ≅ (R/B) Xe 133 was predicted. Applying those methods, R/B from OGL-1 fuel element (5th and 6th) was predicted to show a good agreement with observation. (author)

  6. Reprocessing of nuclear fuels

    International Nuclear Information System (INIS)

    Hatfield, G.W.

    1960-11-01

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

  7. Nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Butterfield, R.S.; Garner, D.L.M.

    1977-01-01

    Reference is made to nuclear fuel assemblies designed for cooling on the 'tube-in-shell' principle in which the fuel is contained by a shell and is cooled by coolant passed through tubes extending through the shell. It has been proposed to employ coated particle fuel as a porous bed on the tube side and the bleed coolant from the tubes into direct contact with the fuel particles. In this way heat is extracted both by direct contact with the fuel and by heat transfer through the coolant tube walls. The system described aims to provide an improved structure of tube and shell for a fuel assembly of this kind and is particularly suitable for use in a gas cooled fast reactor, being able to withstand the neutron flux and high temperature conditions in these reactors. Constructional details are given. (U.K.)

  8. Nuclear fuel element

    International Nuclear Information System (INIS)

    Hirayama, Satoshi; Kawada, Toshiyuki; Matsuzaki, Masayoshi.

    1980-01-01

    Purpose: To provide a fuel element for reducing the mechanical interactions between a fuel-cladding tube and the fuel element and for alleviating the limits of the operating conditions of a reactor. Constitution: A fuel element having mainly uranium dioxide consists of a cylindrical outer pellet and cylindrical inner pellet inserted into the outer pellet. The outer pellet contains two or more additives selected from aluminium oxide, beryllium oxide, magnesium oxide, silicon oxide, sodium oxide, phosphorus oxide, calcium oxide and iron oxide, and the inner pellet contains nuclear fuel substance solely or one additive selected from calcium oxide, silicon oxide, aluminium oxide, magnesium oxide, zirconium oxide and iron oxide. The outer pellet of the fuel thus constituted is reduced in mechanical strength and also in the mechanical interactions with the cladding tube, and the plastic fluidity of the entire pellet is prevented by the inner pellet increased in the mechanical strength. (Kamimura, M.)

  9. Nuclear fuel deformation phenomena

    International Nuclear Information System (INIS)

    Van Brutzel, L.; Dingreville, R.; Bartel, T.J.

    2015-01-01

    Nuclear fuel encounters severe thermomechanical environments. Its mechanical response is profoundly influenced by an underlying heterogeneous microstructure but also inherently dependent on the temperature and stress level histories. The ability to adequately simulate the response of such microstructures, to elucidate the associated macroscopic response in such extreme environments is crucial for predicting both performance and transient fuel mechanical responses. This chapter discusses key physical phenomena and the status of current modelling techniques to evaluate and predict fuel deformations: creep, swelling, cracking and pellet-clad interaction. This chapter only deals with nuclear fuel; deformations of cladding materials are discussed elsewhere. An obvious need for a multi-physics and multi-scale approach to develop a fundamental understanding of properties of complex nuclear fuel materials is presented. The development of such advanced multi-scale mechanistic frameworks should include either an explicit (domain decomposition, homogenisation, etc.) or implicit (scaling laws, hand-shaking,...) linkage between the different time and length scales involved, in order to accurately predict the fuel thermomechanical response for a wide range of operating conditions and fuel types (including Gen-IV and TRU). (authors)

  10. Determination of end-of-life-failure fractions of HTGR-fuel particles by postirradiation annealing and beta autoradiography

    International Nuclear Information System (INIS)

    Thiele, B.A.; Herren, M.

    1978-11-01

    Fission-product contamination of the helium coolant of High-Temperature Gas-Cooled Reactors (HTGR) is strongly influenced by the end-of-life (EOL) failed-particle fraction. Knowledge of the EOL-failure fraction is the basis for model calculations to predict the total fission product release from the reactor core. After disintegration of irradiation fuel rods, fuel particles are placed in individual holes of a graphite tray. During a 5-h heat treatment at 1000 0 C in a helium atmosphere failed particles leak fission products, especially the volatile cesium, into the graphite. After unloading a β-autoradiograph of the tray is made. Holes that housed defective particles are identified from black spots on the β-sensitive film. The EOL-failure fraction is the ratio of defective particles to the total number of particles tested. The technique is called PIAA, PostIrradiation Annealing and Autoradiography. The PIAA technique was applied to particles of a Trisocoated highly-enriched UO 2 fissile batch irradiated to a burnup of 35% FIMA at an irradiation temperature of 1250 0 C. Visual examination showed all particles to be intact. From 11 to 47% of the particles had failed, as determined by PIAA. Further, postirradiation examination showed that localized corrosion of the silicon carbide coating by fission-product rare-earth chlorides had occurred

  11. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1981-01-01

    A nuclear fuel storage apparatus for use in a water-filled pool is fabricated of a material such as stainless steel in the form of an egg crate structure having vertically extending openings. Fuel may be stored in this basic structure in a checkerboard pattern with high enrichment fuel, or in all openings when the fuel is of low effective enrichment. Inserts of a material such as stainless steel are adapted to fit within these openings so that a water gap and, therefore, a flux trap is formed between adjacent fuel storage locations. These inserts may be added at a later time and fuel of a higher enrichment may be stored in each opening. When it is desired to store fuel of still greater enrichment, poison plates may be added to the water gap formed by the installed insert plates, or substituted for the insert plates. Alternately, or in addition, fuel may be installed in high neutron absorption poison boxes which surround the fuel assembly. The stainless steel inserts and the poison plates are each not required until the capacity of the basic egg crate structure is approached. Purchase of these items can, therefore, be deferred for many years. Should the fuel to be stored be of higher enrichment than initially forecast, the deferred decision on the poison plates makes it possible to obtain increased poison in the plates to satisfy the newly discovered requirement

  12. Nuclear fuel element

    International Nuclear Information System (INIS)

    Knowles, A.N.

    1979-01-01

    A nuclear fuel-containing body for a high temperature gas cooled nuclear reactor is described which comprises a flat plate in which the nuclear fuel is contained as a dispersion of fission product-retaining coated fuel particles in a flat sheet of graphitic or carbonaceous matrix material. The flat sheet is clad with a relatively thin layer of unfuelled graphite bonded to the sheet by being formed initially from a number of separate preformed graphitic artefacts and then platen-pressed on to the exterior surfaces of the flat sheet, both the matrix material and the artefacts being in a green state, to enclose the sheet. A number of such flat plates are supported edge-on to the coolant flow in the bore of a tube made of neutron moderating material. Where a number of tiers of plates are superimposed on one another, the abutting edges are chamfered to reduce vibration. (author)

  13. Nuclear fuel strategies

    International Nuclear Information System (INIS)

    Rippon, S.

    1989-01-01

    The paper reports on two international meetings on nuclear fuel strategies, one organised by the World Nuclear Fuel Market in Seville (Spain) October 1988, and the other organised by the American and European nuclear societies in Washington (U.S.A.) November 1988. At the Washington meeting a description was given of the uranium supply and demand market, whereas free trade in uranium was considered in Seville. Considerable concern was expressed at both meetings on the effect on the uranium and enrichment services market of very low prices for spot deals being offered by China and the Soviet Union. Excess enrichment capacity, the procurement policies of the USA and other countries, and fuel cycle strategies, were also discussed. (U.K.)

  14. HTGR Economic / Business Analysis and Trade Studies Market Analysis for HTGR Technologies and Applications

    Energy Technology Data Exchange (ETDEWEB)

    Richards, Matt [Ultra Safe Nuclear Corporation, Los Alamos, NM (United States); Hamilton, Chris [Ultra Safe Nuclear Corporation, Los Alamos, NM (United States)

    2013-11-01

    This report provides supplemental information to the assessment of target markets provided in Appendix A of the 2012 Next Generation Nuclear Plant (NGNP) Industry Alliance (NIA) business plan [NIA 2012] for deployment of High Temperature Gas-Cooled Reactors (HTGRs) in the 2025 – 2050 time frame. This report largely reiterates the [NIA 2012] assessment for potential deployment of 400 to 800 HTGR modules (100 to 200 HTGR plants with 4 reactor modules) in the 600-MWt class in North America by 2050 for electricity generation, co-generation of steam and electricity, oil sands operations, hydrogen production, and synthetic fuels production (e.g., coal to liquids). As the result of increased natural gas supply from hydraulic fracturing, the current and historically low prices of natural gas remain a significant barrier to deployment of HTGRs and other nuclear reactor concepts in the U.S. However, based on U.S. Department of Energy (DOE) Energy Information Agency (EIA) data, U.S. natural gas prices are expected to increase by the 2030 – 2040 timeframe when a significant number of HTGR modules could be deployed. An evaluation of more recent EIA 2013 data confirms the assumptions in [NIA 2012] of future natural gas prices in the range of approximately $7/MMBtu to $10/MMBtu during the 2030 – 2040 timeframe. Natural gas prices in this range will make HTGR energy prices competitive with natural gas, even in the absence of carbon-emissions penalties. Exhibit ES-1 presents the North American projections in each market segment including a characterization of the market penetration logic. Adjustments made to the 2012 data (and reflected in Exhibit ES-1) include normalization to the slightly larger 625MWt reactor module, segregation between steam cycle and more advanced (higher outlet temperature) modules, and characterization of U.S. synthetic fuel process applications as a separate market segment.

  15. Nuclear fuel element

    International Nuclear Information System (INIS)

    Penrose, R.T.; Thompson, J.R.

    1976-01-01

    A method of protecting the cladding of a nuclear fuel element from internal attack and a nuclear fuel element for use in the core of a nuclear reactor are disclosed. The nuclear fuel element has disposed therein an additive of a barium-containing material and the barium-containing material collects reactive gases through chemical reaction or adsorption at temperatures ranging from room temperature up to fuel element plenum temperatures. The additive is located in the plenum of the fuel element and preferably in the form of particles in a hollow container having a multiplicity of gas permeable openings in one portion of the container with the openings being of a size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles. The additive is comprised of elemental barium or a barium alloy containing one or more metals in addition to barium such as aluminum, zirconium, nickel, titanium and combinations thereof. 6 claims, 3 drawing figures

  16. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    Niedrig, T.

    1987-01-01

    Nuclear fuel supply is viewed as a buyer's market of assured medium-term stability. Even on a long-term basis, no shortage is envisaged for all conceivable expansion schedules. The conversion and enrichment facilities developed since the mid-seventies have done much to stabilize the market, owing to the fact that one-sided political decisions by the USA can be counteracted efficiently. In view of the uncertainties concerning realistic nuclear waste management strategies, thermal recycling and mixed oxide fuel elements might increase their market share in the future. Capacities are being planned accordingly. (orig.) [de

  17. Nuclear fuel element

    International Nuclear Information System (INIS)

    Grossman, L.N.; Levin, H.A.

    1975-01-01

    A nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smallr than the size of the particles. The container is preferably held in the spring in the plenum of the fuel element. (E.C.B.)

  18. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Kawada, Toshiyuki; Hirayama, Satoshi; Yoneya, Katsutoshi.

    1980-01-01

    Purpose: To enable load-depending operation as well as moderation for the restriction of operation conditions in the present nuclear reactors, by specifying the essential ingredients and the total weight of the additives to UO 2 fuel substances. Constitution: Two or more additives selected from Al 2 O 3 , B 2 O, CaO, MgO, SiO 2 , Na 2 O and P 2 O 5 are added by the total weight of 2 - 5% to fuel substances consisting of UO 2 or a mixture of UO 2 and PuO 2 . When the mixture is sintered, the strength of the fuel elements is decreased and the fuel-cladding interactions due to the difference in the heat expansion coefficients between the ceramic fuel elements and the metal claddings are decreased to a substantially harmless degree. (Horiuchi, T.)

  19. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Domoto, Noboru; Masuda, Hiroyuki

    1989-01-01

    In a nuclear fuel assembly loaded with a plurality of fuel rods, the inside of a fuel rod disposed at a high neutron flux region is divided into an inner region and an outer region, and more burnable poisons are mixed in the inner region than in the outer region. Alternatively, the central portion of a pellet disposed in a high neutron flux region is made hollow, in which burnable poisons are charged. This can prevent neutron infinite multiplication factor from decreasing extremely at the initial burning stage. Further, the burnable poisons are not rapidly burnt completely and local peaking coefficient can be controlled. Accordingly, in a case of suppressing a predetermined excess reactivity by using a fuel rod incorporated with the burnable poison, the fuel economy can be improved more and the reactor core controllability can also be improved as compared with the usual case. (T.M.)

  20. Present status of research on hydrogen energy and perspective of HTGR hydrogen production system

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yoshiaki; Ogawa, Masuro; Akino, Norio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2001-03-01

    A study was performed to make a clear positioning of research and development on hydrogen production systems with a High Temperature Gas-cooled Reactor (HTGR) under currently promoting at the Japan Atomic Energy Research Institute through a grasp of the present status of hydrogen energy, focussing on its production and utilization as an energy in future. The study made clear that introduction of safe distance concept for hydrogen fire and explosion was practicable for a HTGR hydrogen production system, including hydrogen properties and need to provide regulations applying to handle hydrogen. And also generalization of hydrogen production processes showed technical issues of the HTGR system. Hydrogen with HTGR was competitive to one with fossil fired system due to evaluation of production cost. Hydrogen is expected to be used as promising fuel of fuel cell cars in future. In addition, the study indicated that there were a large amount of energy demand alternative to high efficiency power generation and fossil fuel with nuclear energy through the structure of energy demand and supply in Japan. Assuming that hydrogen with HTGR meets all demand of fuel cell cars, an estimation would show introduction of the maximum number of about 30 HTGRs with capacity of 100 MWt from 2020 to 2030. (author)

  1. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    The papers presented at the International Conference on The Nuclear Fuel Cycle, held at Stockholm, 28 to 31 October 1975, are reviewed. The meeting, organised by the U.S. Atomic Industrial Forum, and the Swedish Nuclear Forum, was concerned more particularly with economic, political, social and commercial aspects than with tecnology. The papers discussed were considered under the subject heading of current status, uranium resources, enrichment, and reprocessing. (U.K.)

  2. Nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    1975-12-01

    The papers presented at the International Conference on The Nuclear Fuel Cycle, held at Stockholm, 28 to 31 October 1975, are reviewed. The meeting, organised by the U.S. Atomic Industrial Forum, and the Swedish Nuclear Forum, was concerned more particularly with economic, political, social and commercial aspects than with tecnology. The papers discussed were considered under the subject heading of current status, uranium resources, enrichment, and reprocessing.

  3. Encapsulating spent nuclear fuel

    International Nuclear Information System (INIS)

    Fleischer, L.R.; Gunasekaran, M.

    1979-01-01

    A system is described for encapsulating spent nuclear fuel discharged from nuclear reactors in the form of rods or multi-rod assemblies. The rods are completely and contiguously enclosed in concrete in which metallic fibres are incorporated to increase thermal conductivity and polymers to decrease fluid permeability. This technique provides the advantage of acceptable long-term stability for storage over the conventional underwater storage method. Examples are given of suitable concrete compositions. (UK)

  4. HTGR Cost Model Users' Manual

    International Nuclear Information System (INIS)

    Gandrik, A.M.

    2012-01-01

    The High Temperature Gas-Cooler Reactor (HTGR) Cost Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Cost Model calculates an estimate of the capital costs, annual operating and maintenance costs, and decommissioning costs for a high-temperature gas-cooled reactor. The user can generate these costs for multiple reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for a single or four-pack configuration; and for a reactor size of 350 or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Cost Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Cost Model. This model was design for users who are familiar with the HTGR design and Excel. Modification of the HTGR Cost Model should only be performed by users familiar with Excel and Visual Basic.

  5. Nuclear fuel cycle information workshop

    International Nuclear Information System (INIS)

    1983-01-01

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

  6. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Hirano, Yasushi; Hirukawa, Koji; Sakurada, Koichi.

    1994-01-01

    A bundle of fuel rods is divided into four fuel rod group regions of small fuel rod bundles by a cross-shaped partitioning structure consisting of paired plate-like structures which connect two opposing surfaces of a channel box. A water removing material with less neutron absorption (for example, Zr or a Zr alloy) or a solid moderator is inserted and secured to a portion of a non-boiling water region interposed between the paired plate-like structure. It has a structure that light water flows to the region in the plate-like structure. The volume, density or composition of the water removing material is controlled depending on the composition of the fuels, to change the moderating characteristics of neutrons in the non-boiling water region. This can easily moderate the difference of nuclear characteristics between each of fuel assemblies using fuel materials of different fuel compositions. Further, the reactivity control effect of the burnable poisons can be enhanced without worsening fuel economy or linear power density. (I.N.)

  7. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1977-01-01

    This invention relates to a nuclear fuel assembly for a light or heavy water reactor, or for a fast reactor of the kind with a bundle of cladded pins, maintained parallel to each other in a regular network by an assembly of separate supporting grids, fitted with elastic bearing surfaces on these pins [fr

  8. Nuclear fuel pellets

    International Nuclear Information System (INIS)

    Larson, R.I.; Brassfield, H.C.

    1981-01-01

    Increased strength and physical durability in green bodies or pellets formed of particulate nuclear fuel oxides is achieved by inclusion of a fugitive binder which is ammonium bicarbonate, bicarbonate carbomate, carbomate, sesquicarbonate or mixtures thereof. Ammonium oxadate may be included as pore former. (author)

  9. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Ito, Arata; Wakamatsu, Mitsuo.

    1976-01-01

    Object: To permit the coolant in an FBR type reactor to enter from the entrance nozzle into a nuclear fuel assembly without causing cavitation. Structure: In a nuclear fuel assembly, which comprises a number of thin fuel pines bundled together at a uniform spacing and enclosed within an outer cylinder, with a handling head connected to an upper portion of the outer cylinder and an entrance nozzle connected to a lower portion of the cylinder, the inner surface of the entrance nozzle is provided with a buffer member and an orifice successively in the direction of flow of the coolant. The coolant entering from a low pressure coolant chamber into the entrance nozzle strikes the buffer member and is attenuated, and thereafter flows through an orifice into the outer cylinder. (Horiuchi, T.)

  10. Nuclear fuel element

    International Nuclear Information System (INIS)

    Hirama, H.

    1978-01-01

    A nuclear fuel element comprises an elongated tube having upper and lower end plugs fixed to both ends thereof and nuclear fuel pellets contained within the tube. The fuel pellets are held against the lower end plug by a spring which is supported by a setting structure. The setting structure is maintained at a proper position at the middle of the tube by a wedge effect caused by spring force exerted by the spring against a set of balls coacting with a tapered member of the setting structure thereby wedging the balls against the inner wall of the tube, and the setting structure is moved free by pushing with a push bar against the spring force so as to release the wedge effect

  11. High temperature heat exchange: nuclear process heat applications

    International Nuclear Information System (INIS)

    Vrable, D.L.

    1980-09-01

    The unique element of the HTGR system is the high-temperature operation and the need for heat exchanger equipment to transfer nuclear heat from the reactor to the process application. This paper discusses the potential applications of the HTGR in both synthetic fuel production and nuclear steel making and presents the design considerations for the high-temperature heat exchanger equipment

  12. Integral nuclear fuel element assembly

    International Nuclear Information System (INIS)

    Schluderberg, D. C.

    1985-01-01

    An integral nuclear fuel element assembly utilizes longitudinally finned fuel pins. The continuous or interrupted fins of the fuel pins are brazed to fins of juxtaposed fuel pins or directly to the juxtaposed fuel pins or both. The integrally brazed fuel assembly is designed to satisfy the thermal and hydraulic requirements of a fuel assembly lattice having moderator to fuel atom ratios required to achieve high conversion and breeding ratios

  13. Nuclear fuel cycle techniques

    International Nuclear Information System (INIS)

    Pecqueur, Michel; Taranger, Pierre

    1975-01-01

    The production of fuels for nuclear power plants involves five principal stages: prospecting of uranium deposits (on the ground, aerial, geochemical, geophysical, etc...); extraction and production of natural uranium from the deposits (U content of ores is not generally high and a chemical processing is necessary to obtain U concentrates); production of 235 U enriched uranium for plants utilizing this type of fuel (a description is given of the gaseous diffusion process widely used throughout the world and particularly in France); manufacture of suitable fuel elements for the different plants; reprocessing of spent fuels for the purpose of not only recovering the fissile materials but also disposing safely of the fission products and other wastes [fr

  14. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Vikhorev, Yu.V.; Biryukov, G.I.; Kirilyuk, N.A.; Lobanov, V.N.

    1977-01-01

    A fuel assembly is proposed for nuclear reactors allowing remote replacement of control rod bundles or their shifting from one assembly to another, i.e., their multipurpose use. This leads to a significant increase in fuel assembly usability. In the fuel assembly the control rod bundle is placed in guide tube channels to which baffles are attached for fuel element spacing. The remote handling of control rods is provided by a hollow cylinder with openings in its lower bottom through which the control rods pass. All control rods in a bundle are mounted to a cross beam which in turn is mounted in the cylinder and is designed for grasping the whole rod bundle by a remotely controlled telescopic mechanism in bundle replacement or shifting. (Z.M.)

  15. Experience with nuclear fuel utilization in Bulgaria

    Energy Technology Data Exchange (ETDEWEB)

    Harizanov, Y [Committee on the Use of Atomic Energy for Peaceful Purposes, Sofia (Bulgaria)

    1997-12-01

    The presentation on experience with nuclear fuel utilization in Bulgaria briefly reviews the situation with nuclear energy in Bulgaria and then discusses nuclear fuel performance (amount of fuel loaded, type of fuel, burnup, fuel failures, assemblies deformation). 2 tabs.

  16. HTGR nuclear power plants: features of the VGR-50 high temperature reactor

    International Nuclear Information System (INIS)

    Glebov, V.P.; Bogoyavlenskii, R.G.; Glushkov, E.S.; Grebennik, V.N.; Ponomarev-Stepnoi, N.N.; Vinogradov, V.P.

    1983-01-01

    Current developmental trends in the power industry are guided to an appreciable extent by the increasing shortages of fossil fuels (coal, petroleum, natural gas) and by ecological problems. Assuming a continuing trend in worldwide consumption of energy resources, we see the electric power industry using up 20%, the other 80% (petroleum, coal, natural gas) going into generating industrial process heat and space heat, transportation, the chemical processing industry, the metallurgical industry, and other branches of industry. In the future, nuclear power will have the job of not only meeting the needs of the electric power industry, but also generating process heat. The most promising type of nuclear power plant available for solving complex problems in generation of electric power and heat for technological processes in the metallurgical processing industry and chemical processing industry is the one based around high-temperature reactors

  17. Nuclear fuel cycle system analysis

    International Nuclear Information System (INIS)

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

    2012-04-01

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

  18. HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

    1998-03-01

    As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

  19. Nuclear fuel waste disposal

    International Nuclear Information System (INIS)

    1982-01-01

    This film for a general audience deals with nuclear fuel waste management in Canada, where research is concentrating on land based geologic disposal of wastes rather than on reprocessing of fuel. The waste management programme is based on cooperation of the AECL, various universities and Ontario Hydro. Findings of research institutes in other countries are taken into account as well. The long-term effects of buried radioactive wastes on humans (ground water, food chain etc.) are carefully studied with the help of computer models. Animated sequences illustrate the behaviour of radionuclides and explain the idea of a multiple barrier system to minimize the danger of radiation hazards

  20. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E. D.

    1984-01-01

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

  1. Nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E. D.

    1984-10-16

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

  2. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    1975-01-01

    The nuclear fuel assembly described includes a cluster of fuel elements supported at a distance from each other so that their axes are parallel in order to establish secondary channels between them reserved for the coolant. Several ducts for an auxiliary cooling fluid are arranged in the cluster. The wall of each duct is pierced with coolant ejection holes which are placed circumferentially to a pre-determined pattern established according to the position of the duct in the cluster and by the axial distance of the ejection hole along the duct. This assembly is intended for reactors cooled by light or heavy water [fr

  3. Nuclear fuel activities in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Cox, D S [Fuel Development Branch, Chalk River Labs., AECL (Canada)

    1997-12-01

    Nuclear fuel activities in Canada are considered in the presentation on the following directions: Canadian utility fuel performance; CANDU owner`s group fuel programs; AECL advanced fuel program (high burnup fuel behaviour and development); Pu dispositioning (MOX) activities. 1 tab.

  4. Nuclear fuel brokerage

    International Nuclear Information System (INIS)

    Hoffman, J.; Schreiber, K.

    1985-01-01

    Making available nuclear fuels on the spot market, especially uranium in various compounds and processing stages, has become an important service rendered nuclear power plant operators. A secondary market has grown, both for natural uranium and for separative work, the conditions and transactions of which require a comprehensive overview of what is going on, especially also in connection with possibilities to terminate in a profitable manner existing contracts. This situation has favored the activity of brokers with excellent knowledge of the market, who are able to handle the complicated terms and conditions in an optimum way. (orig.) [de

  5. Postirradiation examination of Peach Bottom HTGR Driver Fuel Element E06-01

    International Nuclear Information System (INIS)

    Dyer, F.F.; Wichner, R.P.; Martin, W.J.; Fairchild, L.L.; Kedl, R.J.; de Nordwall, H.J.

    1976-04-01

    The report presented describes the postirradiation examinations of driver fuel element E06-01, which had been irradiated an equivalent of 384 full-power days in Peach Bottom, Unit 1. The fuel element is described in detail and its temperature and irradiation service history briefly outlined. Results presented include: (1) visual observations; (2) critical dimensions of fuel compacts, sleeve, and spine; (3) axial distributions of gamma-emitting nuclides plus 3 H and 90 Sr; (4) radial distributions of these nuclides in the sleeve and spine at three axial locations in the fueled regions and three locations in the upper reflector; (5) metallographic examination of samples of fuel compact material; and (6) burnup determinations via radiochemical analyses at two compact locations

  6. Results of Koo measurements of HTGR lattice by oscillated zero reactivity technique using the AGIP-NUCLEARE RB-2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, F; Brighenti, G.; Chiodi, P. L.; Ghilardotti, G.; Giuliani, C.

    1974-10-15

    This paper describes k-infinity measurements conducted using an assembly of loose HTGR coated particles in the BR-2 reactor by means of null reactivity oscillating method comparing the effect of poisoned and unpoisoned lattices like tests performed in the Physical Constants Test Reactor (PCTR) at Hanford. The RB-2 reactor was the property of the Italian firm AGIP NUCLEARE and operated at the Montecuccolino Center in Bologna.

  7. Compact nuclear fuel storage

    International Nuclear Information System (INIS)

    Kiselev, V.V.; Churakov, Yu.A.; Danchenko, Yu.V.; Bylkin, B.K.; Tsvetkov, S.V.

    1983-01-01

    Different constructions of racks for compact storage of spent fuel assemblies (FA) in ''coolin''g pools (CP) of NPPs with the BWR and PWR type reactors are described. Problems concerning nuclear and radiation safety and provision of necessary thermal conditions arising in such rack design are discussed. It is concluded that the problem of prolonged fuel storage at NPPs became Very actual for many countries because of retapdation of the rates of fuel reprocessing centers building. Application of compact storage racks is a promising solution of the problem of intermediate FA storage at NPPs. Such racks of stainless boron steel and with neutron absorbers in the from of boron carbide panels enable to increase the capacity of the present CP 2-2.6 times, and the period of FA storage in them up to 5-10 years

  8. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    1975-01-01

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

  9. Nuclear fuel pin

    International Nuclear Information System (INIS)

    Hartley, Kenneth; Moulding, T.L.J.; Rostron, Norman.

    1979-01-01

    Fuel pin for use in fast breeder nuclear reactors containing fissile and fertile areas of which the fissile and fertile materials do not mix. The fissile material takes the shape of large and small diameter microspheres (the small diameter microspheres can pass through the interstices between the large microspheres). The barrier layers being composed of microspheres with a diameter situated between those of the large and small microspheres ensure that the materials do not mix [fr

  10. Alternative nuclear fuel cycles

    International Nuclear Information System (INIS)

    Till, C.E.

    1979-01-01

    This diffuse subject involves value judgments that are political as well as technical, and is best understood in that context. The four questions raised here, however, are mostly from the technical viewpoints: (1) what are alternative nuclear fuel cycles; (2) what generalizations are possible about their characteristics; (3) what are the major practical considerations; and (4) what is the present situation and what can be said about the outlook for the future

  11. Vented nuclear fuel element

    International Nuclear Information System (INIS)

    Oguma, M.; Hirose, Y.

    1976-01-01

    A description is given of a vented nuclear fuel element having a plenum for accumulation of fission product gases and plug means for delaying the release of the fission product gases from the plenum, the plug means comprising a first porous body wettable with a liquid metal and a second porous body non-wettable with the liquid metal, the first porous body being impregnated with the liquid metal and in contact with the liquid metal

  12. Nuclear reactor fuel element splitter

    International Nuclear Information System (INIS)

    Yeo, D.

    1976-01-01

    A method and apparatus are disclosed for removing nuclear fuel from a clad fuel element. The fuel element is power driven past laser beams which simultaneously cut the cladding lengthwise into at least two longitudinal pieces. The axially cut lengths of cladding are then separated, causing the nuclear fuel contained therein to drop into a receptacle for later disposition. The cut lengths of cladding comprise nuclear waste which is disposed of in a suitable manner. 6 claims, 10 drawing figures

  13. HTGR R and D programs

    International Nuclear Information System (INIS)

    Neylan, A.J.; Brisbois, J.

    1979-01-01

    A significant R and D program (including in certain cases full-scale prototype tests) formed the basis for the design and key elements in the foregoing projects and is continuing to provide a basis for generic design development. HTGR R and D programs are both privately and government sponsored. This paper provides an overview of the background, current status and outstanding design issues/problems remaining in the area of NSS Plant, Materials and Fuel. The specific objectives and scope of all recently completed, ongoing and planned major HTGR R and D programs are presented

  14. Use of spikants in HTGR fuel and their effect on costs

    International Nuclear Information System (INIS)

    Brooks, L.H.

    1979-05-01

    The costs of fresh fuel fabrication have been estimated for flowsheets that have spikant added (Co-60) at various points. The costs are compared with refabricated and fresh fuel fabrication costs. It is shown that the costs increase as the spikant is added nearer to the plant feed point. The least cost is achieved by adding a detachable spikant to the finished fuel element. The highest cost is incurred when the spikant is fed in with the plant feed. This cost is greater than that for refabricated fuel because extra process cells are necessary to prepare the spikant and, in addition, the gamma flux from the Co-60 is much greater than from U-232 and greater precautions have to be taken

  15. Nuclear fuel handling apparatus

    International Nuclear Information System (INIS)

    Andrea, C.; Dupen, C.F.G.; Noyes, R.C.

    1977-01-01

    A fuel handling machine for a liquid metal cooled nuclear reactor in which a retractable handling tube and gripper are lowered into the reactor to withdraw a spent fuel assembly into the handling tube. The handling tube containing the fuel assembly immersed in liquid sodium is then withdrawn completely from the reactor into the outer barrel of the handling machine. The machine is then used to transport the spent fuel assembly directly to a remotely located decay tank. The fuel handling machine includes a decay heat removal system which continuously removes heat from the interior of the handling tube and which is capable of operating at its full cooling capacity at all times. The handling tube is supported in the machine from an articulated joint which enables it to readily align itself with the correct position in the core. An emergency sodium supply is carried directly by the machine to provide make up in the event of a loss of sodium from the handling tube during transport to the decay tank. 5 claims, 32 drawing figures

  16. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    International Nuclear Information System (INIS)

    Radulescu, H.

    2001-01-01

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report

  17. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    Energy Technology Data Exchange (ETDEWEB)

    H. radulescu

    2001-09-28

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report.

  18. Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

  19. Distribution of fission products in Peach Bottom HTGR fuel element E01-01

    International Nuclear Information System (INIS)

    Wichner, R.P.; Dyer, F.F.; Martin, W.J.; Fairchild, L.L.

    1978-10-01

    The fifth in a projected series of six postirradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements is described. The element analyzed received an equivalent of 897 full-power days of irradiation prior to the scheduled termination of Core 2 operation. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a 137 Cs inventory of 20.3 Ci in the graphite sleeve and 8.1 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides 134 Cs, /sup 110 m/Ag, 60 Co, and 154 Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the beta-emitters 3 H, 14 C, and 90 Sr were obtained at four axial locations of the fueled region of the element sleeve and two axial locations of the element spine. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. In addition to fission product distributions, the appearance of the component parts of the element was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed

  20. South Korea's nuclear fuel industry

    International Nuclear Information System (INIS)

    Clark, R.G.

    1990-01-01

    March 1990 marked a major milestone for South Korea's nuclear power program, as the country became self-sufficient in nuclear fuel fabrication. The reconversion line (UF 6 to UO 2 ) came into full operation at the Korea Nuclear Fuel Company's fabrication plant, as the last step in South Korea's program, initiated in the mid-1970s, to localize fuel fabrication. Thus, South Korea now has the capability to produce both CANDU and pressurized water reactor (PWR) fuel assemblies. This article covers the nuclear fuel industry in South Korea-how it is structures, its current capabilities, and its outlook for the future

  1. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.

    1981-01-01

    A nuclear fuel loading apparatus, incorporating a microprocessor control unit, is described which automatically loads nuclear fuel pellets into dual fuel rods with a minimum of manual involvement and in a manner and sequence to ensure quality control and accuracy. (U.K.)

  2. Behaviour of HTGR coated fuel particles at high-temperature tests

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Lyutikov, R.A.; Kurbakov, S.D.; Repnikov, V.M.; Khromonozhkin, V.V.; Soloviyov, G.I.

    1990-01-01

    At the temperature range 1200-2600 deg. C prereactor tests of TRISO fuel particles on the base of UO 2 , UC x O y and UO 2 +2Al 2 O 3 . SiO 2 kernels, and also fuel particle models with ZrC kernels were performed. Isothermal annealings carried out at temperatures of 1400-2600 deg. C, thermogradient ones at 1200-2200 deg. C (Δ T = 200-1200 deg. C/cm). It is shown that at heating to 2200 deg. C integrity of fuel particles is limited by different thermal expansion of PyC and SiC coatings, and also by thermal dissociation of SiC. At higher temperatures the failure is caused by development of high pressures within weakened fuel particles. It is found that uranium migration from alloyed fuel (UC x O y , UO 2 +2Al 2 O 3 .SiO 2 ) in the process of annealing is higher than that from UO 2 . (author)

  3. Future trends in nuclear fuels

    International Nuclear Information System (INIS)

    Guitierrez, J.E.

    2006-01-01

    This series of transparencies presents: the fuel management cycle and key areas (security of supplies, strategies and core management, reliability, spent fuel management), the world nuclear generating capacity, concentrate capacity, enrichment capacity, and manufacturing capacity forecasts, the fuel cycle strategies and core management (longer cycles, higher burnups, power up-rates, higher enrichments), the Spanish nuclear generation cost, the fuel reliability (no defects, robust designs, operational margins, integrated fuel and core design), spent fuel storage (design and safety criteria, fuel performance and integrity). (J.S.)

  4. Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

    International Nuclear Information System (INIS)

    Foerthmann, R.; Groos, E.; Gruebmeier, H.

    1975-08-01

    Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention of the kernel, which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl 2 Si 2 O 8 , BaAl 2 Si 2 O 8 and CsAlSi 2 O 6 in the additional aluminasilica phase of the kernel. (orig.) [de

  5. Nuclear fuel element

    International Nuclear Information System (INIS)

    Iwano, Yoshihiko.

    1993-01-01

    Microfine cracks having a depth of less than 10% of a pipe thickness are disposed radially from a central axis each at an interval of less than 100 micron over the entire inner circumferential surface of a zirconium alloy fuel cladding tube. For manufacturing such a nuclear fuel element, the inside of the cladding tube is at first filled with an electrolyte solution of potassium chloride. Then, electrolysis is conducted using the cladding tube as an anode and the electrolyte solution as a cathode, and the inner surface of the cladding tube with a zirconium dioxide layer having a predetermined thickness. Subsequently, the cladding tube is laid on a smooth steel plate and lightly compressed by other smooth steel plate to form microfine cracks in the zirconium dioxide layer on the inner surface of the cladding tube. Such a compressing operation is continuously applied to the cladding tube while rotating the cladding tube. This can inhibit progress of cracks on the inner surface of the cladding tube, thereby enabling to prevent failure of the cladding tube even if a pellet/cladding tube mechanical interaction is applied. Accordingly, reliability of the nuclear fuel elements is improved. (I.N.)

  6. Demonstration tests for HTGR fuel elements and core components with test sections in HENDEL

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yoshiaki; Hino, Ryutaro; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1995-03-01

    In the fuel stack test section (T{sub 1}) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T{sub 2}). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs.

  7. Fault diagnosis of generation IV nuclear HTGR components – Part II: The area error enthalpy–entropy graph approach

    International Nuclear Information System (INIS)

    Rand, C.P. du; Schoor, G. van

    2012-01-01

    Highlights: ► Different uncorrelated fault signatures are derived for HTGR component faults. ► A multiple classifier ensemble increases confidence in classification accuracy. ► Detailed simulation model of system is not required for fault diagnosis. - Abstract: The second paper in a two part series presents the area error method for generation of representative enthalpy–entropy (h–s) fault signatures to classify malfunctions in generation IV nuclear high temperature gas-cooled reactor (HTGR) components. The second classifier is devised to ultimately address the fault diagnosis (FD) problem via the proposed methods in a multiple classifier (MC) ensemble. FD is realized by way of different input feature sets to the classification algorithm based on the area and trajectory of the residual shift between the fault-free and the actual operating h–s graph models. The application of the proposed technique is specifically demonstrated for 24 single fault transients considered in the main power system (MPS) of the Pebble Bed Modular Reactor (PBMR). The results show that the area error technique produces different fault signatures with low correlation for all the examined component faults. A brief evaluation of the two fault signature generation techniques is presented and the performance of the area error method is documented using the fault classification index (FCI) presented in Part I of the series. The final part of this work reports the application of the proposed approach for classification of an emulated fault transient in data from the prototype Pebble Bed Micro Model (PBMM) plant. Reference data values are calculated for the plant via a thermo-hydraulic simulation model of the MPS. The results show that the correspondence between the fault signatures, generated via experimental plant data and simulated reference values, are generally good. The work presented in the two part series, related to the classification of component faults in the MPS of different

  8. Nuclear fuel element

    International Nuclear Information System (INIS)

    Armijo, J.S.

    1977-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed which has a composite cladding having a substrate, a metal barrier metallurgically bonded to the inside surface of the substrate and an inner layer metallurgically bonded to the inside surface of the metal barrier. In this composite cladding, the inner layer and the metal barrier shield the substrate from any impurities or fission products from the nuclear fuel material held within the composite cladding. The metal barrier forms about 1 to about 4 percent of the thickness of the cladding and is comprised of a metal selected from the group consisting of niobium, aluminum, copper, nickel, stainless steel, and iron. The inner layer and then the metal barrier serve as reaction sites for volatile impurities and fission products and protect the substrate from contact and reaction with such impurities and fission products. The substrate and the inner layer of the composite cladding are selected from conventional cladding materials and preferably are a zirconium alloy. Also in a preferred embodiment the substrate and the inner layer are comprised of the same material, preferably a zirconium alloy. 19 claims, 2 figures

  9. Improved nuclear fuel element

    International Nuclear Information System (INIS)

    1980-01-01

    The invention is of a nuclear fuel element which comprises a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium and mixtures thereof, and an elongated composite cladding container comprising a zirconium alloy tube containing constituents other than zirconium in an amount greater than about 5000 parts per million by weight and an undeformed metal barrier of moderate purity zirconium bonded to the inside surface of the alloy tube. The container encloses the core so as to leave a gap between the container and the core during use in a nuclear reactor. The metal barrier is of moderate purity zirconium with an impurity level on a weight basis of at least 1000ppm and less than 5000ppm. Impurity levels of specific elements are given. Variations of the invention are also specified. The composite cladding reduces chemical interaction, minimizes localized stress and strain corrosion and reduces the likelihood of a splitting failure in the zirconium alloy tube. Other benefits are claimed. (U.K.)

  10. Burning of spent fuel of an accelerator-driven modular HTGR in sub-critical condition

    International Nuclear Information System (INIS)

    Jing Xingqing; Yang Yongwei; Chang Hong; Wu Zongxin; Gu Yuxiang

    2002-01-01

    The modular high temperature gas cooled reactor (MHTGR) has good safety characteristics because of the use of coated particles in the fuel element. After the particles cool outside of the reactor for some time, the spent fuel can be re-utilized. The author describes a physics feasibility study for the burning of spent fuel from a 350 MW ring-shaped modular high temperature gas cooled reactor in an accelerator-driven sub-critical reactor. A conceptual design is given for the 30 MW accelerator-driven sub-critical reactor. The neutron transport in the sub-critical reactor was simulated using the MCNP code, and the burnup was calculated using the ORIGEN2 code. The results show that the accelerator-driven sub-critical gas-cooled reactor has reliable sub-criticality and low power density and that the spent fuel from a 350 MW ring-shaped modular high temperature gas cooled reactor can be burned to provide 20% more energy

  11. 1976 scientific progress report. [Fuel and coating materials for HTGR]; Wissenschaftlicher Ergebnisberict 1976

    Energy Technology Data Exchange (ETDEWEB)

    Nickel, H.

    1976-07-01

    Activities at the Institute for Reactor Materials in the production and properties of high temperature gas cooled reactor fuel and coating materials are summarized. Major emphasis was placed on investigations of pyrocarbon, BISO and TRISO coatings, uranium and thorium oxides and carbides, and graphite and matrix materials. A list of publications is included. (HDR)

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

    International Nuclear Information System (INIS)

    1978-09-01

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

  13. 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

  14. Quality management of nuclear fuel

    International Nuclear Information System (INIS)

    2006-01-01

    The Guide presents the quality management requirements to be complied with in the procurement, design, manufacture, transport, receipt, storage, handling and operation of nuclear fuel. The Guide also applies to control rods and shield elements to be placed in the reactor. The Guide is mainly aimed for the licensee responsible for the procurement and operation of fuel, for the fuel designer and manufacturer and for other organisations, whose activities affect fuel quality and the safety of fuel transport, storage and operation. General requirements for nuclear fuel are presented in Section 114 of the Finnish Nuclear Energy Decree and in Section 15 of the Government Decision (395/1991). Regulatory control of the safety of fuel is described in Guides YVL6.1, YVL6.2 and YVL6.3. An overview of the regulatory control of nuclear power plants carried out by STUK (Radiation and Nuclear Safety Authority, Finland) is clarified in Guide YVL1.1

  15. Nuclear power and the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1976-07-01

    The IAEA is organizing a major conference on nuclear power and the nuclear fuel cycle, which is to be held from 2 to 13 May 1977 in Salzburg, Austria. The programme for the conference was published in the preceding issue of the IAEA Bulletin (Vol.18, No. 3/4). Topics to be covered at the conference include: world energy supply and demand, supply of nuclear fuel and fuel cycle services, radioactivity management (including transport), nuclear safety, public acceptance of nuclear power, safeguarding of nuclear materials, and nuclear power prospects in developing countries. The articles in the section that follows are intended to serve as an introduction to the topics to be discussed at the Salzburg Conference. They deal with the demand for uranium and nuclear fuel cycle services, uranium supplies, a computer simulation of regional fuel cycle centres, nuclear safety codes, management of radioactive wastes, and a pioneering research project on factors that determine public attitudes toward nuclear power. It is planned to present additional background articles, including a review of the world nuclear fuel reprocessing situation and developments in the uranium enrichment industry, in future issues of the Bulletin. (author)

  16. Nuclear fuel supplies

    International Nuclear Information System (INIS)

    1960-01-01

    When the International Atomic Energy Agency was set up nearly three years ago, it was widely believed that it would soon become a world bank or broker for the supply of nuclear fuel. Some observers now seem to feel that this promise has been rather slow to come to fruition. A little closer analysis would, however, show that the promise can be fulfilled only in a certain objective context, and to the extent that this context exists, the development of the Agency's role has been commensurate with the actual needs of the situation

  17. Nuclear fuel waste disposal

    International Nuclear Information System (INIS)

    Allan, C.J.

    1993-01-01

    The Canadian concept for nuclear fuel waste disposal is based on disposing of the waste in a vault excavated 500-1000 m deep in intrusive igneous rock of the Canadian Shield. The author believes that, if the concept is accepted following review by a federal environmental assessment panel (probably in 1995), then it is important that implementation should begin without delay. His reasons are listed under the following headings: Environmental leadership and reducing the burden on future generations; Fostering public confidence in nuclear energy; Forestalling inaction by default; Preserving the knowledge base. Although disposal of reprocessing waste is a possible future alternative option, it will still almost certainly include a requirement for geologic disposal

  18. Development of a FE Model for the Stress Analysis of HTGR TRISO-coated particle fuel

    International Nuclear Information System (INIS)

    Cho, Moon Sung; Lee, Y. W.; Jeong, K. C.; Kim, Y. K.; Oh, S. C.; Chang, J. H.

    2005-12-01

    Finite element modelling of the stresses in TRISO-coated fuel particle under normal operating conditions was carried out with use of the structural analysis computer code ABAQUS. The FE model took into account the irradiation induced swelling and the creep of the PyC layers, the internal fission gas pressure that builds up during irradiation and the constant external ambient pressure. All of the inputs such as particle dimensions, swelling rates and creep rates of PyC layers and other mechanical properties used in these calculations were adopted from Miller's publication published in 1993. The FE model was verified against Miller's solution. Results of this model were found to be in good agreement with Miller's results. With use of the FE model, the static behavior of the TRISO-coated fuel particle, such as load shares, stress contours, stress variations as a function of fluence and shape changes of the TRISO -coated layers were investigated

  19. 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

  20. 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

  1. Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

    Energy Technology Data Exchange (ETDEWEB)

    Foerthmann, R.; Groos, E.; Gruebmeier, H.

    1975-08-15

    Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl2Si2O8, BaAl2Si2O8and CsAlSi2O6 in the additional alumina-silica phase of the kernel.

  2. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    International Nuclear Information System (INIS)

    Shropshire, D.E.; Herring, J.S.

    2004-01-01

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  3. Regulation at nuclear fuel cycle

    International Nuclear Information System (INIS)

    2002-01-01

    This bulletin contains information about activities of the Nuclear Regulatory Authority of the Slovak Republic (UJD). In this leaflet the role of the UJD in regulation at nuclear fuel cycle is presented. The Nuclear Fuel Cycle (NFC) is a complex of activities linked with production of nuclear fuel for nuclear reactors as a source of energy used for production of electricity and heat, and of activities linked with spent nuclear fuel handling. Activities linked with nuclear fuel (NF) production, known as the Front-End of Nuclear Fuel Cycle, include (production of nuclear fuel from uranium as the most frequently used element). After discharging spent nuclear fuel (SNF) from nuclear reactor the activities follow linked with its storage, reprocessing and disposal known as the Back-End of Nuclear Fuel Cycle. Individual activity, which penetrates throughout the NFC, is transport of nuclear materials various forms during NF production and transport of NF and SNF. Nuclear reactors are installed in the Slovak Republic only in commercial nuclear power plants and the NFC is of the open type is imported from abroad and SNF is long-term supposed without reprocessing. The main mission of the area of NFC is supervision over: - assurance of nuclear safety throughout all NFC activities; - observance of provisions of the Treaty on Non-Proliferation of Nuclear Weapons during nuclear material handling; with an aim to prevent leakage of radioactive substances into environment (including deliberated danage of NFC sensitive facilities and misuse of nuclear materials to production of nuclear weapons. The UJD carries out this mission through: - assessment of safety documentation submitted by operators of nuclear installations at which nuclear material, NF and SNF is handled; - inspections concentrated on assurance of compliance of real conditions in NFC, i.e. storage and transport of NF and SNF; storage, transport and disposal of wastes from processing of SNF; with assumptions of the safety

  4. Nuclear power generation and nuclear fuel

    International Nuclear Information System (INIS)

    Okajima, Yasujiro

    1985-01-01

    As of June 30, 1984, in 25 countries, 311 nuclear power plants of about 209 million kW were in operation. In Japan, 27 plants of about 19 million kW were in operation, and Japan ranks fourth in the world. The present state of nuclear power generation and nuclear fuel cycle is explained. The total uranium resources in the free world which can be mined at the cost below $130/kgU are about 3.67 million t, and it was estimated that the demand up to about 2015 would be able to be met. But it is considered also that the demand and supply of uranium in the world may become tight at the end of 1980s. The supply of uranium to Japan is ensured up to about 1995, and the yearly supply of 3000 st U 3 O 8 is expected in the latter half of 1990s. The refining, conversion and enrichment of uranium are described. In Japan, a pilot enrichment plant consisting of 7000 centrifuges has the capacity of about 50 t SWU/year. UO 2 fuel assemblies for LWRs, the working of Zircaloy, the fabrication of fuel assemblies, the quality assurance of nuclear fuel, the behavior of UO 2 fuel, the grading-up of LWRs and nuclear fuel, and the nuclear fuel business in Japan are reported. The reprocessing of spent fuel and plutonium fuel are described. (Kako, I.)

  5. Financing the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Stephany, M.

    1975-01-01

    While conventional power stations usually have fossil fuel reserves for only a few weeks, nuclear power stations, because of the relatively long time required for uranium processing from ore extraction to the delivery of the fuel elements and their prolonged in-pile time, require fuel reserves for a period of several years. Although the specific fuel costs of nuclear power stations are much lower than those of conventional power stations, this results in consistently higher financial requirements. But the problems involved in financing the nuclear fuel do not only include the aspect of financing the requirements of reactor operators, but also of financing the facilities of the nuclear fuel cycle. As far as the fuel supply is concerned, the true financial requirements greatly exceed the mere purchasing costs because the costs of financing are rather high as a consequence of the long lead times. (orig./UA) [de

  6. Comparison of US/FRG accident condition models for HTGR fuel failure and radionuclide release

    International Nuclear Information System (INIS)

    Verfondern, K.

    1991-03-01

    The objective was to compare calculation models used in safety analyses in the US and FRG which describe fission product release behavior from TRISO coated fuel particles under core heatup accident conditions. The frist step performed is the qualitative comparison of both sides' fuel failure and release models in order to identify differences and similarities in modeling assumptions and inputs. Assumptions of possible particle failure mechanisms under accident conditions (SiC degradation, pressure vessel) are principally the same on both sides though they are used in different modeling approaches. The characterization of a standard (= intact) coated particle to be of non-releasing (GA) or possibly releasing (KFA/ISF) type is one of the major qualitative differences. Similar models are used regarding radionuclide release from exposed particle kernels. In a second step, a quantitative comparison of the calculation models was made by assessing a benchmark problem predicting particle failure and radionuclide release under MHTGR conduction cooldown accident conditions. Calculations with each side's reference method have come to almost the same failure fractions after 250 hours for the core region with maximum core heatup temperature despite the different modeling approaches of SORS and PANAMA-I. The comparison of the results of particle failure obtained with the Integrated Failure and Release Model for Standard Particles and its revision provides a 'verification' of these models in this sense that the codes (SORS and PANAMA-II, and -III, respectively) which were independently developed lead to very good agreement in the predictions. (orig./HP) [de

  7. IAEA activities on nuclear fuel

    International Nuclear Information System (INIS)

    Basak, U.

    2011-01-01

    In this paper a brief description and the main objectives of IAEA Programme B on Nuclear fuel cycle are given. The following Coordinated Research Projects: 1) FUel performance at high burn-up and in ageing plant by management and optimisation of WAter Chemistry Technologies (FUWAC ); 2) Near Term and Promising Long Term Options for Deployment of Thorium Based Nuclear Energy; 3) Fuel Modelling (FUMEX-III) are shortly described. The data collected by the IAEA Expert Group of Fuel Failures in Water Cooled Reactors including information about fuel failure cause for PWR (1994-2006) and failure mechanisms for BWR fuel (1994-2006) are shown. The just published Fuel Failure Handbook as well as preparation of a Monograph on Zirconium including an overview of Zirconium for nuclear applications are presented. The current projects in Sub-programme B2 - Power Reactor Fuel Engineering are also listed

  8. Nuclear fuel pellet charging device

    International Nuclear Information System (INIS)

    Komuro, Kojiro.

    1990-01-01

    The present invention concerns a nuclear fuel pellet loading device, in which nuclear fuel pellets are successively charged from an open end of a fuel can while rotating the can. That is, a fuel can sealed at one end with an end plug and opened at the other end is rotated around its pipe axis as the center on a rotationally diriving table. During rotation of the fuel can, nuclear fuel pellets are successively charged by means of a feed rod of a feeding device to the inside of the fuel can. The fuel can is rotated while being supported horizontally and the fuel pellets are charged from the open end thereof. Alternatively, the fuel can is rotated while being supported obliquely and the fuel pellets are charged gravitationally into the fuel can. In this way, the damages to the barrier of the fuel can can be reduce. Further, since the fuel pellets can be charged gravitationally by rotating the fuel can while being supported obliquely, the damages to the barrier can be reduced remarkably. (I.S.)

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

    International Nuclear Information System (INIS)

    Holder, N.; Abraham, L.

    1977-08-01

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

  10. Nuclear power fuel cycle

    International Nuclear Information System (INIS)

    Havelka, S.; Jakesova, L.

    1982-01-01

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

  11. HTGR strategy for reduced proliferation potential

    International Nuclear Information System (INIS)

    Stewart, H.B.; Dahlberg, R.C.

    1978-01-01

    The HTGR stratregy for reduced proliferation potential is one aspect of a potential broader nuclear strategy aimed primarily toward a transition nuclear period between today's uranium-consumption reactors and the long-range balanced system of breeder and advanced near-breeder reactors. In particular, the normal commerce of U-233 could be made acceptable by: (a) dependence on the gamma radiation from U-232 daughter products, (b) enhancement of that radioactivity by incomplete fission-product decontamination of the bred-fuel, or (c) denaturing of the U-233 with U-238. These approaches would, of course, supplement institutional initiatives to improve proliferation resistance such as the collocation of facilities and the establishment of secure energy centers. 6 refs

  12. Nuclear power and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Scurr, I.F.; Silver, J.M.

    1990-01-01

    Australian Nuclear Science and Technology Organization maintains an ongoing assessment of the world's nuclear technology developments, as a core activity of its Strategic Plan. This publication reviews the current status of the nuclear power and the nuclear fuel cycle in Australia and around the world. Main issues discussed include: performances and economics of various types of nuclear reactors, uranium resources and requirements, fuel fabrication and technology, radioactive waste management. A brief account of the large international effort to demonstrate the feasibility of fusion power is also given. 11 tabs., ills

  13. Effect of the Heat Treatment on the Graphite Matrix of Fuel Element for HTGR

    International Nuclear Information System (INIS)

    Lee, Chungyong; Lee, Seungjae; Suh, Jungmin; Jo, Youngho; Lee, Youngwoo; Cho, Moonsung

    2013-01-01

    In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength for the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and Phenol as a binder were chosen and mixed with each other, formed and heated for the compressive strength test. The objective of this research is to optimize the kinds and composition of the mixed graphite and the forming process by evaluating the compressive strength before/after heat treatment (carbonization of binder). In this study, the effect of heat treatment on graphite matrix was studied in terms of the density and the compressive strength. The size (diameter and length) of pellet is increased by heat treatment. Due to additional weight reduction and swelling (length and diameter) of samples the density of graphite pellet is decreased from about 2.0 to about 1.7g/cm 3 . From the mechanical test results, the compressive strength of graphite pellets was related to the various conditions such as the contents of binder, the kinds of graphite and the heat treatment. Both the green pellet and the heat treated pellet, the compressive strength of G+S+P pellets is relatively higher than that of R+S+P pellets. To optimize fuel element matrix, the effect of Phenol and other binders, graphite composition and the heat treatment on the mechanical properties will be deeply investigated for further study

  14. Cesium transport data for HTGR systems

    International Nuclear Information System (INIS)

    Myers, B.F.; Bell, W.E.

    1979-09-01

    Cesium transport data on the release of cesium from HTGR fuel elements are reviewed and discussed. The data available through 1976 are treated. Equations, parameters, and associated variances describing the data are presented. The equations and parameters are in forms suitable for use in computer codes used to calculate the release of metallic fission products from HTGR fuel elements into the primary circuit. The data cover the following processes: (1) diffusion of cesium in fuel kernels and pyrocarbon, (2) sorption of cesium on fuel rod matrix material and on graphite, and (3) migration of cesium in graphite. The data are being confirmed and extended through work in progress

  15. Operation and postirradiation examination of ORR capsule OF-2: accelerated testing of HTGR fuel

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Thoms, K.R.

    1979-03-01

    Irradiation capsule OF-2 was a test of High-Temperature Gas-Cooled Reactor fuel types under accelerated irradiation conditions in the Oak Ridge Research Reactor. The results showed good irradiation performance of Triso-coated weak-acid-resin fissile particles and Biso-coated fertile particles. These particles had been coated by a fritted gas distributor in the 0.13-m-diam furnace. Fast-neutron damage (E > 0.18 MeV) and matrix-particle interaction caused the outer pyrocarbon coating on the Triso-coated particles to fail. Such failure depended on the optical anisotropy, density, and open porosity of the outer pyrocarbon coating, as well as on the coke yield of the matrix. Irradiation of specimens with values outside prescribed limits for these properties increased the failure rate of their outer pyrocarbon coating. Good irradiation performance was observed for weak-acid-resin particles with conversions in the range from 15 to 75% UC 2

  16. HTGR Generic Technology Program. Semiannual report for the period ending September 30, 1979

    International Nuclear Information System (INIS)

    1979-11-01

    The technical accomplishments on the HTGR Generic Technology Program at General Atomic during the second half of FY-79 are reported. The report covers a period when the major design direction of the National HTGR Program is in the process of changing from the HTGR-SC emphasis to an HTGR-GT emphasis in the near term. The HTGR Generic Technology Program activities have been redirected to ensure that the tasks covered are supportive of this changing emphasis in HTGR applications. The activities include the need to develop an MEU fuel, and the need to qualify materials and components for the higher temperatures of the gas turbine plant

  17. HTGR Generic Technology Program. Semiannual report for the period ending March 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1979-06-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the first half of FY-79. It covers a period when the major design direction of the National HTGR Program is in the process of changing from the HTGR-SC emphasis to an HTGR-GT emphasis in the near term. The HTGR Generic Technology Program activities have been redirected to ensure that the tasks covered are supportive of this changing emphasis in HTGR applications. The activities include the need to develop a medium enriched uranium (MEU) fuel, and the need to qualify materials and components for the higher temperatures of the gas turbine plant.

  18. HTGR Generic Technology Program. Semiannual report for the period ending March 31, 1979

    International Nuclear Information System (INIS)

    1979-06-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the first half of FY-79. It covers a period when the major design direction of the National HTGR Program is in the process of changing from the HTGR-SC emphasis to an HTGR-GT emphasis in the near term. The HTGR Generic Technology Program activities have been redirected to ensure that the tasks covered are supportive of this changing emphasis in HTGR applications. The activities include the need to develop a medium enriched uranium (MEU) fuel, and the need to qualify materials and components for the higher temperatures of the gas turbine plant

  19. Romanian nuclear fuel cycle development

    International Nuclear Information System (INIS)

    Rapeanu, S.N.; Comsa, Olivia

    1998-01-01

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

  20. Reactor Structure Materials: Nuclear Fuel

    International Nuclear Information System (INIS)

    Sannen, L.; Verwerft, M.

    2000-01-01

    Progress and achievements in 1999 in SCK-CEN's programme on applied and fundamental nuclear fuel research in 1999 are reported. Particular emphasis is on thermochemical fuel research, the modelling of fission gas release in LWR fuel as well as on integral experiments

  1. Burnable absorber coated nuclear fuel

    International Nuclear Information System (INIS)

    Chubb, W.; Radford, K.C.; Parks, B.H.

    1984-01-01

    A nuclear fuel body which is at least partially covered by a burnable neutron absorber layer is provided with a hydrophobic overcoat generally covering the burnable absorber layer and bonded directly to it. In a method for providing a UO 2 fuel pellet with a zirconium diboride burnable poison layer, the fuel body is provided with an intermediate niobium layer. (author)

  2. The fuel of nuclear reactors

    International Nuclear Information System (INIS)

    1995-03-01

    This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

  3. The evolving nuclear fuel cycle

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  4. Nuclear Fuel elements

    International Nuclear Information System (INIS)

    Hirakawa, Hiromasa.

    1979-01-01

    Purpose: To reduce the stress gradient resulted in the fuel can in fuel rods adapted to control the axial power distribution by the combination of fuel pellets having different linear power densities. Constitution: In a fuel rod comprising a first fuel pellet of a relatively low linear power density and a second fuel pellet of a relatively high linear power density, the second fuel pellet is cut at its both end faces by an amount corresponding to the heat expansion of the pellet due to the difference in the linear power density to the adjacent first fuel pellet. Thus, the second fuel pellet takes a smaller space than the first fuel pellet in the fuel can. This can reduce the stress produced in the portion of the fuel can corresponding to the boundary between the adjacent fuel pellets. (Kawakami, Y.)

  5. Resin-based preparation of HTGR fuels: operation of an engineering-scale uranium loading system

    International Nuclear Information System (INIS)

    Haas, P.A.

    1977-10-01

    The fuel particles for recycle of 233 U to High-Temperature Gas-Cooled Reactors are prepared from uranium-loaded carboxylic acid ion exchange resins which are subsequently carbonized, converted, and refabricated. The development and operation of individual items of equipment and of an integrated system are described for the resin-loading part of the process. This engineering-scale system was full scale with respect to a hot demonstration facility, but was operated with natural uranium. The feed uranium, which consisted of uranyl nitrate solution containing excess nitric acid, was loaded by exchange with resin in the hydrogen form. In order to obtain high loadings, the uranyl nitrate must be acid deficient; therefore, nitric acid was extracted by a liquid organic amine which was regenerated to discharge a NaNO 3 or NH 4 NO 3 solution waste. Water was removed from the uranyl nitrate solution by an evaporator that yielded condensate containing less than 0.5 ppM of uranium. The uranium-loaded resin was washed with condensate and dried to a controlled water content via microwave heating. The loading process was controlled via in-line measurements of the pH and density of the uranyl nitrate. The demonstrated capacity was 1 kg of uranium per hour for either batch loading contractors or a continuous column as the resin loading contractor. Fifty-four batch loading runs were made without a single failure of the process outlined in the chemical flowsheet or any evidence of inability to control the conditions dictated by the flowsheet

  6. Thorium in nuclear fuel

    International Nuclear Information System (INIS)

    Stankevicius, Alejandro

    2012-01-01

    We revise the advantages and possible problems on the use of thorium as a nuclear fuel instead of uranium. The following aspects are considered: 1) In the world there are three times more thorium than uranium 2) In spite that thorium in his natural form it is not a fisil, under neutron irradiation, is possible to transform it to uranium 233, a fisil of a high quality. 3) His ceramic oxides properties are superior to uranium or plutonium oxides. 4) During the irradiation the U 233 due to n,2n reaction produce small quantities of U 232 and his decay daughters' bismuth 212 and thallium 208 witch are strong gamma source. In turn thorium 228 and uranium 232 became, in time anti-proliferate due to there radiation intensity. 5) As it is described in here and experiments done in several countries reactors PHWR can be adapted to the use of thorium as a fuel element 6) As a problem we should mentioned that the different steps in the process must be done under strong radiation shielding and using only automatized equipment s (author)

  7. Determination of reactivity of multiplying systems filled with spherical HTGR-fuel elements using kinetic methods with regard to the pulsed-neutron method

    International Nuclear Information System (INIS)

    Drueke, V.

    1978-06-01

    At three critical or subcritical facilities - two of them filled with spherical HTGR-fuel elements - the reactivity is determined using kinetic methods. Besides the inverskinetic method the applicability of the pulsed-neutron method is investigated. The experimental results using the pulsed-neutron method are compared partly with the inverskinetic method and partly with diffusion-calculations. It is shown, that in the HTGR the space dependence of the reactivity in radial direction is not remarkable in spite of the 'kinetic distortion'; on the contrary in axial direction - the direction of the external neutron source - space dependent reactivity worths are measured. The results of the pulsed-neutron methods of Sjoestrand and Simmons-King are rather good applicable in all configurations. For the method of Sjoestrand it is necessary to select the detector positions, whereas for Simmons-King the calculated life-time determines the results. Therefore it is proposed to compare calculated and measured decay constants of the prompt neutron field in future. (orig.) [de

  8. HTGR Application Economic Model Users' Manual

    International Nuclear Information System (INIS)

    Gandrik, A.M.

    2012-01-01

    The High Temperature Gas-Cooled Reactor (HTGR) Application Economic Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Application Economic Model calculates either the required selling price of power and/or heat for a given internal rate of return (IRR) or the IRR for power and/or heat being sold at the market price. The user can generate these economic results for a range of reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for up to 16 reactor modules; and for module ratings of 200, 350, or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Application Economic Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Application Economic Model. This model was designed for users who are familiar with the HTGR design and Excel and engineering economics. Modification of the HTGR Application Economic Model should only be performed by users familiar with the HTGR and its applications, Excel, and Visual Basic.

  9. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    Science.gov (United States)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  10. British Nuclear Fuels (Warrington)

    International Nuclear Information System (INIS)

    Hoyle, D.; Cryer, B.; Bellotti, D.

    1992-01-01

    This adjournment debate is about British Nuclear Fuels plc and the 750 redundancies due to take place by the mid-1990s at BNFL, Risley. The debate was instigated by the Member of Parliament for Warrington, the constituency in which BNFL, Risley is situated. Other members pointed out that other industries, such as the textile industry are also suffering job losses due to the recession. However the MP for Warrington argued that the recent restructuring of BNFL restricted the financial flexibility of BNFL so that the benefits of contracts won for THORP at Sellafield could not help BNFL, Risley. The debate became more generally about training, apprentices and employment opportunities. The Parliamentary Under-Secretary of State for Energy explained the position as he saw it and said BNFL may be able to offer more help to its apprentices. Long- term employment prospects at BNFL are dependent on the future of the nuclear industry in general. The debate lasted about half an hour and is reported verbatim. (U.K)

  11. Nuclear fuel tax in court

    International Nuclear Information System (INIS)

    Leidinger, Tobias

    2014-01-01

    Besides the 'Nuclear Energy Moratorium' (temporary shutdown of eight nuclear power plants after the Fukushima incident) and the legally decreed 'Nuclear Energy Phase-Out' (by the 13th AtG-amendment), also the legality of the nuclear fuel tax is being challenged in court. After receiving urgent legal proposals from 5 nuclear power plant operators, the Hamburg fiscal court (4V 154/13) temporarily obliged on 14 April 2014 respective main customs offices through 27 decisions to reimburse 2.2 b. Euro nuclear fuel tax to the operating companies. In all respects a remarkable process. It is not in favour of cleverness to impose a political target even accepting immense constitutional and union law risks. Taxation 'at any price' is neither a statement of state sovereignty nor one for a sound fiscal policy. Early and serious warnings of constitutional experts and specialists in the field of tax law with regard to the nuclear fuel tax were not lacking. (orig.)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

    A new method for performing thermal stress analyses in structures with multiple penetrations was applied to these analyses. This method couples the development of an equivalent thermal conductivity for the blocks, a technique that has been used extensively for modeling the thermal characteristics of reactor cores, with the use of the equivalent solid plate method for stress analysis. Using this equivalent thermal conductivity, which models as one material the heat transfer characteristics of the fuel, coolant, and graphite two-dimensional, steady-state thermal analyses of the fuel and control rod fuel blocks were performed to establish all temperature boundaries required for the stress analyses. In applying the equivalent solid plate method, the region of penetrations being modeled was replaced by a pseudo material having the same dimensions but whose materials properties were adjusted to account for the penetration. The peak stresses and strains were determined by applying stress and strain intensification factors to the calculated distributions. The condition studied was where the blocks were located near the center of the furnace. In this position, the axial surface of the block is heated near one end and cooled near the other. The approximate axial surface temperatures ranged from 1521 0 C at both the heated and the cooled ends to a peak of 1800 0 C near the center. Five specific cases were analyzed: plane (two-dimensional thermal, plane stress strain) analyses of each end of a standard fuel block (2 cases), plane analyses of each end of a control rod fuel block (2 cases), and a two-dimensional analysis of a fuel block treated as an axisymmetric cylind

  13. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

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

  14. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    2014-01-01

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

  15. HTGR generic technology program. Semiannual report ending March 31, 1980

    International Nuclear Information System (INIS)

    1980-05-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the first half of FY-80. It covers a period when the design direction of the National HTGR Program is in the process of an overall review. The HTGR Generic Technology Program activities have continued so as to provide the basic technology required for all HTGR applications. The activities include the need to develop an MEU fuel and the need to qualify materials and components for the higher temperatures of the gas turbine and process heat plants

  16. US HTGR Deployment Challenges and Strategies HTR 2014 Conference Proceedings

    International Nuclear Information System (INIS)

    Shahrokhi, Farshid; Lommers, Lewis; Mayer, John III; Southworth, Finis

    2014-01-01

    The NGNP Industry Alliance (NIA), LLC (www.NGNPAliance.org), is a consortium of high temperature gas-cooled reactor (HTGR) designers, utility plant owner/operators, critical plant hardware suppliers, and end-user groups. The NIA is promoting the design and commercialization of a HTGR for industrial process heat applications and electricity generation. In 2012, NIA selected the AREVA Steam Cycle HTGR (SC-HTGR) as its primary reactor design choice for its first implementation in mid -2020s. The SC-HTGR can produce 625 MWth of process steam at 550°C or 275 MWe of electricity in a co-generation configuration. The standard plant is a four-pack of 625MWth modules providing steam and electricity co-generation. The safety characteristics of the HTGR technology allows close colocation of the nuclear plant and the industrial end-user. The plant design also allows the process steam used for the industrial applications to be completely segregated and separate from primary Helium coolant and the secondary nuclear steam supply systems. The process steam at temperatures up to 550°C is provided for a variety of direct or indirect applications. End-user requirements are met for a wide range of steam flow, pressure and temperature conditions. Very high reliability (>99.99%) is maintained by the use of multi-reactor modules and conventional gas-fired back-up. Intermittent steam loads can also be efficiently met through co-generation of electricity for internal use or external distribution and sale. The NIA technology development and deployment challenges are met with strategies that provide investment and partnerships opportunities for plant design and equipment supply, and by cooperative government research, sovereign or private investment, and philanthropic opportunities. Our goal is to create intellectual property (IP) and investor value as the design matures and a license is obtained. The strategy also includes involvement of the initial customer in sharing the value created in

  17. Reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

    Kidd, S.

    2008-01-01

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

  18. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1979-01-01

    A method and apparatus for the storage of fuel in a stainless steel egg crate structure within a storage pool are described. Fuel is initially stored in a checkerboard pattern or in each opening if the fuel is of low enrichment. Additional fuel (or fuel of higher enrichment) is later stored by adding stainless steel angled plates within each opening, thereby forming flux traps between the openings. Still higher enrichment fuel is later stored by adding poison plates either with or without the stainless steel angles. 8 claims

  19. Transportation of spent nuclear fuels

    International Nuclear Information System (INIS)

    Meguro, Toshiichi

    1976-01-01

    The spent nuclear fuel taken out of reactors is cooled in the cooling pool in each power station for a definite time, then transported to a reprocessing plant. At present, there is no reprocessing plant in Japan, therefore the spent nuclear fuel is shipped abroad. In this paper, the experiences and the present situation in Japan are described on the transport of the spent nuclear fuel from light water reactors, centering around the works in Tsuruga Power Station, Japan Atomic Power Co. The spent nuclear fuel in Tsuruga Power Station was first transported in Apr. 1973, and since then, about 36 tons were shipped to Britain by 5 times of transport. The reprocessing plant in Japan is expected to start operation in Apr. 1977, accordingly the spent nuclear fuel used for the trial will be transported in Japan in the latter half of this year. Among the permission and approval required for the transport of spent nuclear fuel, the acquisition of the certificate for transport casks and the approval of land and sea transports are main tasks. The relevant laws are the law concerning the regulations of nuclear raw material, nuclear fuel and reactors and the law concerning the safety of ships. The casks used in Tsuruga Power Station and EXL III type, and the charging of spent nuclear fuel, the decontamination of the casks, the leak test, land transport with a self-running vehicle, loading on board an exclusive carrier and sea transport are briefly explained. The casks and the ship for domestic transport are being prepared. (Kato, I.)

  20. R and D status and requirements for PIE in the fields of the HTGR fuel and the innovative basic research on High-Temperature Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Sawa, Kazuhiro; Tobita, Tsutomu; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Ishihara, Masahiro; Hayashi, Kimio; Hoshiya, Taiji; Sekino, Hajime; Ooeda, Etsurou

    1999-09-01

    The High Temperature Engineering Test Reactor (HTTR), which is the first high temperature gas-cooled reactor (HTGR) in Japan, achieved its first criticality in November 1998 at the Oarai Research Establishment of the Japan Atomic Energy Research Institute (JAERI). In the field of HTGR fuel development, JAERI will proceed research and development (R and D) works by the following steps: (STEP-1) confirmation of irradiation performance of the first-loading fuel of the HTTR, (STEP-2) study on irradiation performance of high burnup SiC-coated fuel particle and (STEP-3) development of ZrC-coated fuel particle. Requirements for post-irradiation examination (PIE) are different for each R and D step. In STEP-1, firstly, hot cells will be prepared in the HTTR reactor building to handle spent fuels. In parallel, general equipments such as those for deconsolidation of fuel compacts and for handling coated fuel particles will be installed in the Hot Laboratory at Oarai. In STEP-2, precise PIE techniques, for example, Raman spectroscopy for measurement of stress on irradiated SiC layer, will be investigated. In STEP-3, new PIE techniques should be developed to investigate irradiation behavior of ZrC-coated particle. In the field of the innovative basic research on high-temperature engineering, some preliminary tests have been made on the research areas of (1) new materials development, (2) fusion technology, (3) radiation chemistry and (4) high-temperature in-core instrumentation. Requirements for PIE are under investigation, in particular in the field of the new materials development. Besides more general apparatuses including transmission electron microscopy (TEM), some special apparatuses such as an electron spin resonance (ESR) spectrometer, a specific resistance/Hall coefficient measuring system and a differential scanning calorimeter (DSC) are planned to install in the Hot Laboratory at Oarai. Acquisition of advanced knowledge on the irradiation behavior is expected in

  1. Quality assurance of nuclear fuel

    International Nuclear Information System (INIS)

    1994-01-01

    The guide presents the quality assurance requirements to be completed with in the procurement, design, manufacture, transport, handling and operation of the nuclear fuel. The guide also applies to the procurement of the control rods and the shield elements to be placed in the reactor. The guide is mainly aimed for the licensee responsible for the procurement and operation of fuel, for the fuel designer and manufacturer and for other organizations whose activities affect fuel quality, the safety of fuel transport, storage and operation. (2 refs.)

  2. Nuclear Fuel Cycle & Vulnerabilities

    Energy Technology Data Exchange (ETDEWEB)

    Boyer, Brian D. [Los Alamos National Laboratory

    2012-06-18

    The objective of safeguards is the timely detection of diversion of significant quantities of nuclear material from peaceful nuclear activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. The safeguards system should be designed to provide credible assurances that there has been no diversion of declared nuclear material and no undeclared nuclear material and activities.

  3. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Tomihiro.

    1970-01-01

    The present invention relates to fuel assemblies employing wire wrap spacers for retaining uniform spatial distribution between fuel elements. Clad fuel elements are helically wound in the oxial direction with a wave-formed wire strand. The strand is therefore provided with spring action which permits the fuel elements to expand freely in the axial and radial directions so as to retain proper spacing and reduce stresses due to thermal deformation. (Ownes, K.J.)

  4. Nuclear fuel string assembly

    International Nuclear Information System (INIS)

    Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

    1976-01-01

    A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

  5. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.; Macivergan, R.; Mckenzie, G.W.

    1980-01-01

    An apparatus incorporating a microprocessor control is provided for automatically loading nuclear fuel pellets into fuel rods commonly used in nuclear reactor cores. The apparatus comprises a split ''v'' trough for assembling segments of fuel pellets in rows and a shuttle to receive the fuel pellets from the split ''v'' trough when the two sides of the split ''v'' trough are opened. The pellets are weighed while in the shuttle, and the shuttle then moves the pellets into alignment with a fuel rod. A guide bushing is provided to assist the transfer of the pellets into the fuel rod. A rod carousel which holds a plurality of fuel rods presents the proper rod to the guide bushing at the appropriate stage in the loading sequence. The bushing advances to engage the fuel rod, and the shuttle advances to engage the guide bushing. The pellets are then loaded into the fuel rod by a motor operated push rod. The guide bushing includes a photocell utilized in conjunction with the push rod to measure the length of the row of fuel pellets inserted in the fuel rod

  6. Spent Nuclear Fuel project, project management plan

    International Nuclear Information System (INIS)

    Fuquay, B.J.

    1995-01-01

    The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

  7. IAEA activities on nuclear fuel cycle 1997

    International Nuclear Information System (INIS)

    Oi, N.

    1997-01-01

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

  8. IAEA activities on nuclear fuel cycle 1997

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

  9. Alternatives for nuclear fuel disposal

    International Nuclear Information System (INIS)

    Ramirez S, J. R.; Badillo A, V.; Palacios H, J.; Celis del Angel, L.

    2010-10-01

    The spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments in the construction of repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution? or, What is the best technology for a specific solution? Many countries have deferred the decision on selecting an option, while other works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However, currently is under process an extended power up rate to 20% of their original power and also there are plans to extend operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. So this work describes some different alternatives that have been studied in Mexico to define which will be the best alternative to follow. (Author)

  10. Nuclear Fuel in Cofrentes NPP

    International Nuclear Information System (INIS)

    2002-01-01

    Fuel is an essential in the nuclear power generating business because of its direct implications on safety, generating costs and the operating conditions and limitations of the facility. Fuel management in Cofrentes NPP has been targeted at optimized operation, enhanced reliability and the search for an in-depth knowledge of the design and licensing processes that will provide Iberdrola,as the responsible operator, with access to independent control of safety aspects related to fuel and free access to manufacturing markets. (Author)

  11. Conditioning of nuclear reactor fuel

    International Nuclear Information System (INIS)

    1975-01-01

    A method of conditioning the fuel of a nuclear reactor core to minimize failure of the fuel cladding comprising increasing the fuel rod power to a desired maximum power level at a rate below a critical rate which would cause cladding damage is given. Such conditioning allows subsequent freedom of power changes below and up to said maximum power level with minimized danger of cladding damage. (Auth.)

  12. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

    Dazen, J.R.; Denero, J.V.

    1976-01-01

    A nuclear fuel pellet loading machine is described including an inclined rack mounted on a base and having parallel spaced grooves on its upper surface arranged to support fuel rods. A fuel pellet tray is adapted to be placed on a table spaced from the rack, the tray having columns of fuel pellets which are in alignment with the open ends of fuel rods located in the rack grooves. A transition plate is mounted between the fuel rod rack and the fuel pellet tray to receive and guide the pellets into the open ends of the fuel rods. The pellets are pushed into the fuel rods by a number of mechanical fingers mounted on a motor operated block which is moved along the pellet tray length by a drive screw driven by the motor. To facilitate movement of the pellets in the fuel rods the rack is mounted on a number of spaced vibrators which vibrate the fuel rods during fuel pellet insertion. A pellet sensing device movable into an end of each fuel rod indicates to an operator when each rod has been charged with the correct number of pellets

  13. Nuclear power and its fuel cycle

    International Nuclear Information System (INIS)

    Wymer, R.G.

    1986-01-01

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

  14. Nuclear fuels and development of nuclear fuel elements

    International Nuclear Information System (INIS)

    Sundaram, C.V.; Mannan, S.L.

    1989-01-01

    Safe, reliable and economic operation of nuclear fission reactors, the source of nuclear power at present, requires judicious choice, careful preparation and specialised fabrication procedures for fuels and fuel element structural materials. These aspects of nuclear fuels (uranium, plutonium and their oxides and carbides), fuel element technology and structural materials (aluminium, zircaloy, stainless steel etc.) are discussed with particular reference to research and power reactors in India, e.g. the DHRUVA research reactor at BARC, Trombay, the pressurised heavy water reactors (PHWR) at Rajasthan and Kalpakkam, and the Fast Breeder Test Reactor (FBTR) at Kalpakkam. Other reactors like the gas-cooled reactors operating in UK are also mentioned. Because of the limited uranium resources, India has opted for a three-stage nuclear power programme aimed at the ultimate utilization of her abundant thorium resources. The first phase consists of natural uranium dioxide-fuelled, heavy water-moderated and cooled PHWR. The second phase was initiated with the attainment of criticality in the FBTR at Kalpakkam. Fast Breeder Reactors (FBR) utilize the plutonium and uranium by-products of phase 1. Moreover, FBR can convert thorium into fissile 233 U. They produce more fuel than is consumed - hence, the name breeders. The fuel parameters of some of the operating or proposed fast reactors in the world are compared. FBTR is unique in the choice of mixed carbides of plutonium and uranium as fuel. Factors affecting the fuel element performance and life in various reactors e.g. hydriding of zircaloys, fuel pellet-cladding interaction etc. in PHWR and void swelling; irradiation creep and helium embrittlement of fuel element structural materials in FBR are discussed along with measures to overcome some of these problems. (author). 15 refs., 9 tabs., 23 figs

  15. Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Fritsching, G.; Wolf, G. [Internationale Atomreaktorbau G.m.b.H. (INTERATOM), Bergisch Gladbach (Germany, F.R.)

    1981-01-15

    The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen.

  16. Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

    International Nuclear Information System (INIS)

    Fritsching, G.; Wolf, G.

    1981-01-01

    The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen

  17. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Betten, P.R.

    1976-01-01

    Under the invention the fuel assembly is particularly suitable for liquid metal cooled fast neutron breeder reactors. Hence, according to the invention a fuel assembly cladding includes inward corrugations with respect to the remainder of the cladding according to a recurring pattern determined by the pitch of the metal wire helically wound round the fuel rods of the assembly. The parts of the cladding pressed inwards correspond to the areas in which the wire encircling the peripheral fuel rods is generally located apart from the cladding, thereby reducing the play between the cladding and the peripheral fuel rods situated in these areas. The reduction in the play in turn improves the coolant flow in the internal secondary channels of the fuel assembly to the detriment of the flow in the peripheral secondary channels and thereby establishes a better coolant fluid temperature profile [fr

  18. Nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Natori, Hisahide; Kurihara, Kunitoshi.

    1982-01-01

    Purpose: To increase the fuel safety by decreasing the gap conductance between fuels and cladding tubes, as well as improve the reactor core controllability by rendering the void coefficient negative. Constitution: Fuel assemblies in a pressure tube comprise a tie-rod, fuel rods in a central region, and fuel rods with burnable poison in the outer circumference region. Here, B 4 C is used as the burnable poison by 1.17 % by weight ratio. The degrees of enrichment for the fissile plutonium as PuO 2 -UO 2 fuel used in the assemblies are 2.7 %, 2.7 % and 1.5 % respectively in the innermost layer, the intermediate layer and the outermost layer. This increases the burn-up degree to improve the plant utilizability, whereby the void coefficient is rendered negative to improve the reactor core controllability. (Horiuchi, T.)

  19. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sakurai, Shungo; Ogiya, Shunsuke.

    1990-01-01

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

  20. International Nuclear Fuel Cycle Evaluation

    International Nuclear Information System (INIS)

    Carnesale, A.

    1980-01-01

    As nuclear power expands globally, so too expands the capability for producing nuclear weapons. The International Nuclear Fuel Cycle Evaluation (INFCE) was organized in 1977 for the purpose of exploring two areas: (1) ways in which nuclear energy can be made available to help meet world energy needs, and (2) means by which the attendant risk of weapons proliferation can be held to a minimum. INFCE is designed for technical and analytical study rather than negotiation. Its organizational structure and issues under consideration are discussed. Some even broader issues that emerge from consideration of the relationships between the peaceful and military use of nuclear energy are also discussed. These are different notions of the meaning of nuclear proliferation, nuclear export policy, the need of a nuclear policy to be both a domestic as well as a foreign one, and political-military measures that can help reduce incentives of countries to acquire nuclear weapons of their own

  1. Nuclear fuel financing

    International Nuclear Information System (INIS)

    Lurf, G.

    1975-01-01

    Fuel financing is only at its beginning. A logical way of developing financing model is a step by step method starting with the financing of pre-payments. The second step will be financing of natural uranium and enrichment services to the point where the finished fuel elements are delivered to the reactor operator. The third step should be the financing of fuel elements during the time the elements are inserted in the reactor. (orig.) [de

  2. Nuclear fuel cycle. V. 1

    International Nuclear Information System (INIS)

    1983-01-01

    Nuclear fuel cycle information in the main countries that develop, supply or use nuclear energy is presented. Data about Japan, FRG, United Kingdom, France and Canada are included. The information is presented in a tree-like graphic way. (C.S.A.) [pt

  3. Nuclear Fuel Cycle Introductory Concepts

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-02

    The nuclear fuel cycle is a complex entity, with many stages and possibilities, encompassing natural resources, energy, science, commerce, and security, involving a host of nations around the world. This overview describes the process for generating nuclear power using fissionable nuclei.

  4. Nuclear Fuel Cycle Introductory Concepts

    International Nuclear Information System (INIS)

    Karpius, Peter Joseph

    2017-01-01

    The nuclear fuel cycle is a complex entity, with many stages and possibilities, encompassing natural resources, energy, science, commerce, and security, involving a host of nations around the world. This overview describes the process for generating nuclear power using fissionable nuclei.

  5. Nuclear fuel cycle. V. 2

    International Nuclear Information System (INIS)

    1984-01-01

    Nuclear fuel cycle information in some countries that develop, supply or use nuclear energy is presented. Data about Argentina, Australia, Belgium, Netherlands, Italy, Denmarmark, Norway, Sweden, Switzerland, Finland, Spain and India are included. The information is presented in a tree-like graphic way. (C.S.A.) [pt

  6. 3D Nondestructive Visualization and Evaluation of TRISO Particles Distribution in HTGR Fuel Pebbles Using Cone-Beam Computed Tomography

    Directory of Open Access Journals (Sweden)

    Gongyi Yu

    2017-01-01

    Full Text Available A nonuniform distribution of tristructural isotropic (TRISO particles within a high-temperature gas-cooled reactor (HTGR pebble may lead to excessive thermal gradients and nonuniform thermal expansion during operation. If the particles are closely clustered, local hotspots may form, leading to excessive stresses on particle layers and an increased probability of particle failure. Although X-ray digital radiography (DR is currently used to evaluate the TRISO distributions in pebbles, X-ray DR projection images are two-dimensional in nature, which would potentially miss some details for 3D evaluation. This paper proposes a method of 3D visualization and evaluation of the TRISO distribution in HTGR pebbles using cone-beam computed tomography (CBCT: first, a pebble is scanned on our high-resolution CBCT, and 2D cross-sectional images are reconstructed; secondly, all cross-sectional images are restructured to form the 3D model of the pebble; then, volume rendering is applied to segment and display the TRISO particles in 3D for visualization and distribution evaluation. For method validation, several pebbles were scanned and the 3D distributions of the TRISO particles within the pebbles were produced. Experiment results show that the proposed method provides more 3D than DR, which will facilitate pebble fabrication research and production quality control.

  7. HTGR Measurements and Instrumentation Systems

    International Nuclear Information System (INIS)

    Ball, Sydney J.; Holcomb, David Eugene; Cetiner, Mustafa Sacit

    2012-01-01

    This report provides an integrated overview of measurements and instrumentation for near-term future high-temperature gas-cooled reactors (HTGRs). Instrumentation technology has undergone revolutionary improvements since the last HTGR was constructed in the United States. This report briefly describes the measurement and communications needs of HTGRs for normal operations, maintenance and inspection, fuel fabrication, and accident response. The report includes a description of modern communications technologies and also provides a potential instrumentation communications architecture designed for deployment at an HTGR. A principal focus for the report is describing new and emerging measurement technologies with high potential to improve operations, maintenance, and accident response for the next generation of HTGRs, known as modular HTGRs, which are designed with passive safety features. Special focus is devoted toward describing the failure modes of the measurement technologies and assessing the technology maturity.

  8. Spent Nuclear Fuel Project Safety Management Plan

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1996-02-01

    The Spent Nuclear Fuel Project Safety Management Plan describes the new nuclear facility regulatory requirements basis for the Spemt Nuclear Fuel (SNF) Project and establishes the plan to achieve compliance with this basis at the new SNF Project facilities

  9. Device for reprocessing nuclear fuels

    International Nuclear Information System (INIS)

    Hatano, Mamoru.

    1981-01-01

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

  10. Studies of the use of high-temperature nuclear heat from an HTGR for hydrogen production

    Science.gov (United States)

    Peterman, D. D.; Fontaine, R. W.; Quade, R. N.; Halvers, L. J.; Jahromi, A. M.

    1975-01-01

    The results of a study which surveyed various methods of hydrogen production using nuclear and fossil energy are presented. A description of these methods is provided, and efficiencies are calculated for each case. The process designs of systems that utilize the heat from a general atomic high temperature gas cooled reactor with a steam methane reformer and feed the reformer with substitute natural gas manufactured from coal, using reforming temperatures, are presented. The capital costs for these systems and the resultant hydrogen production price for these cases are discussed along with a research and development program.

  11. Studies of the use of high-temperature nuclear heat from an HTGR for hydrogen production

    International Nuclear Information System (INIS)

    Peterman, D.D.; Fontaine, R.W.; Quade, R.N.; Halvers, L.J.; Jahromi, A.M.

    1975-01-01

    The results of a study which surveyed various methods of hydrogen production using nuclear and fossil energy are presented. A description of these methods is provided, and efficiencies are calculated for each case. The process designs of systems that utilize the heat from a general atomic high temperature gas cooled reactor with a steam methane reformer and feed the reformer with substitute natural gas manufactured from coal, using reforming temperatures, are presented. The capital costs for these systems and the resultant hydrogen production price for these cases are discussed along with a research and development program

  12. Nuclear fuel element end fitting

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    A typical embodiment of the invention has an array of sockets that are welded to the intersections of the plates that form the upper and lower end fittings of a nuclear reactor fuel element. The sockets, which are generally cylindrical in shape, are oriented in directions that enable the longitudinal axes of the sockets to align with the longitudinal axes of the fuel rods that are received in the respective sockets. Detents impressed in the surfaces of the sockets engage mating grooves that are formed in the ends of the fuel rods to provide for the structural integrity of the fuel element

  13. Nuclear fuel recycling system

    International Nuclear Information System (INIS)

    Lee, H.R.; Koch, A.K.; Krawczyk, A.

    1981-01-01

    A process is provided for recycling sintered uranium dioxide fuel pellets rejected during fuel manufacture and the swarf from pellet grinding. The scrap material is prepared mechanically by crushing and milling as a high solids content slurry, using scrap sintered UO 2 pellets as the grinding medium under an inert atmosophere

  14. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Obara, Hiroshi.

    1981-01-01

    Purpose: To suppress iodine release thereby prevent stress corrosion cracks in fuel cans by dispersing ferrous oxide at the outer periphery of sintered uranium dioxide pellets filled and sealed within zirconium alloy fuel cans of fuel elements. Constitution: Sintered uranium dioxide pellets to be filled and sealed within a zirconium alloy fuel can are prepared either by mixing ferric oxide powder in uranium dioxide powder, sintering and then reducing at low temperature or by mixing iron powder in uranium dioxide powder, sintering and then oxidizing at low temperature. In this way, ferrous oxide is dispersed on the outer periphery of the sintered uranium dioxide pellets to convert corrosive fission products iodine into iron iodide, whereby the iodine release is suppressed and the stress corrosion cracks can be prevented in the fuel can. (Moriyama, K.)

  15. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Borrman, B.; Nylund, O.

    1984-01-01

    A fuel assembly with a fuel channel which surrounds a plurality of fuel rods and which is divided, by means of a stiffening device of cruciform cross-section and four wings, into four sub-channels each of which comprises a bundle of fuel rods. Each fuel channel side has a plurality of stamped, inwardly-directed projections, arranged vertically one after the other, aid projections being welded to one and the same stiffening wing. Each one of the wall portions located between the projections defines, together with two adjacently positioned projections and a portion of the stiffening wing, a communiation opening between two bundles located on on one side each of the stiffening wing. (Author)

  16. Nuclear fuel element

    International Nuclear Information System (INIS)

    Yamanaka, Tsuneyasu.

    1976-01-01

    Purpose: To provide a mechanism for the prevention of fuel pellet dislocation in fuel can throughout fuel fablication, fuel transportation and reactor operation. Constitution: A plenum spacer as a mechanism for the prevention of fuel pellet dislocation inserted into a cladding tube comprises split bodies bundled by a frame and an expansion body being capable of inserting into the central cavity of the split bodies. The expansion body is, for example, in a conical shape and the split bodies are formed so that they define in the center portion, when disposed along the inner wall of the cladding tube, a gap capable of inserting the conical body. The plenum spacer is assembled by initially inserting the split bodies in a closed state into the cladding tube after the loading of the pellets, pressing their peripheral portions and then inserting the expansion body into the space to urge the split bodies to the inner surface of the cladding tube. (Kawakami, Y.)

  17. Nuclear fuels accounting interface: River Bend experience

    International Nuclear Information System (INIS)

    Barry, J.E.

    1986-01-01

    This presentation describes nuclear fuel accounting activities from the perspective of nuclear fuels management and its interfaces. Generally, Nuclear Fuels-River Bend Nuclear Group (RBNG) is involved on a day-by-day basis with nuclear fuel materials accounting in carrying out is procurement, contract administration, processing, and inventory management duties, including those associated with its special nuclear materials (SNM)-isotopics accountability oversight responsibilities as the Central Accountability Office for the River Bend Station. As much as possible, these duties are carried out in an integrated, interdependent manner. From these primary functions devolve Nuclear Fuels interfacing activities with fuel cost and tax accounting. Noting that nuclear fuel tax accounting support is of both an esoteric and intermittent nature, Nuclear Fuels-RBNG support of developments and applications associated with nuclear fuel cost accounting is stressed in this presentation

  18. Rack for nuclear fuel elements

    International Nuclear Information System (INIS)

    Rubinstein, H.J.; Gordon, C.B.; Robison, A.; Clark, P.M.

    1977-01-01

    Disclosed is a rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed spent fuel elements. Each fuel element is supported at the lower end thereof by a respective support that rests on the floor of the spent fuel pool for a nuclear power plant. An open rack frame is employed as an upright support for the enclosures containing the spent fuel elements. Legs at the lower corners of the frame rest on the floor of the pool to support the frame. In one exemplary embodiment, the support for the fuel element is in the form of a base on which a fuel element rests and the base is supported by legs. In another exemplary embodiment, each fuel element is supported on the pool floor by a self-adjusting support in the form of a base on which a fuel element rests and the base rests on a ball or swivel joint for self-alignment. The lower four corners of the frame are supported by legs adjustable in height for leveling the frame. Each adjustable frame leg is in the form of a base resting on the pool floor and the base supports a threaded post. The threaded post adjustably engages a threaded column on which rests the lower end of the frame. 16 claims, 14 figures

  19. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Wada, Toyoji.

    1979-01-01

    Purpose: To remove failures caused from combination of fuel-cladding interactions, hydrogen absorptions, stress corrosions or the likes by setting the quantity ratio of uranium or uranium and plutonium relative to oxygen to a specific range in fuel pellets and forming a specific size of a through hole at the center of the pellets. Constitution: In a fuel rods of a structure wherein fuel pellets prepared by compacting and sintering uranium dioxide, or oxide mixture consisting of oxides of plutonium and uranium are sealed with a zirconium metal can, the ratio of uranium or uranium and plutonium to oxygen is specified as 1 : 2.01 - 1 : 2.05 in the can and a passing hole of a size in the range of 15 - 30% of the outer diameter of the fuel pellet is formed at the center of the pellet. This increases the oxygen partial pressure in the fuel rod, oxidizes and forms a protection layer on the inner surface of the can to control the hydrogen absorption and stress corrosion. Locallized stress due to fuel cladding interaction (PCMI) can also be moderated. (Horiuchi, T.)

  20. Material input of nuclear fuel

    International Nuclear Information System (INIS)

    Rissanen, S.; Tarjanne, R.

    2001-01-01

    The Material Input (MI) of nuclear fuel, expressed in terms of the total amount of natural material needed for manufacturing a product, is examined. The suitability of the MI method for assessing the environmental impacts of fuels is also discussed. Material input is expressed as a Material Input Coefficient (MIC), equalling to the total mass of natural material divided by the mass of the completed product. The material input coefficient is, however, only an intermediate result, which should not be used as such for the comparison of different fuels, because the energy contents of nuclear fuel is about 100 000-fold compared to the energy contents of fossil fuels. As a final result, the material input is expressed in proportion to the amount of generated electricity, which is called MIPS (Material Input Per Service unit). Material input is a simplified and commensurable indicator for the use of natural material, but because it does not take into account the harmfulness of materials or the way how the residual material is processed, it does not alone express the amount of environmental impacts. The examination of the mere amount does not differentiate between for example coal, natural gas or waste rock containing usually just sand. Natural gas is, however, substantially more harmful for the ecosystem than sand. Therefore, other methods should also be used to consider the environmental load of a product. The material input coefficient of nuclear fuel is calculated using data from different types of mines. The calculations are made among other things by using the data of an open pit mine (Key Lake, Canada), an underground mine (McArthur River, Canada) and a by-product mine (Olympic Dam, Australia). Furthermore, the coefficient is calculated for nuclear fuel corresponding to the nuclear fuel supply of Teollisuuden Voima (TVO) company in 2001. Because there is some uncertainty in the initial data, the inaccuracy of the final results can be even 20-50 per cent. The value

  1. Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

    Science.gov (United States)

    Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has

  2. Transport of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    1980-01-01

    In response to public interest in the transport by rail through London of containers of irradiated fuel elements on their way from nuclear power stations to Windscale, the Central Electricity Generating Board and British Rail held three information meetings in London in January 1980. One meeting was for representatives of London Borough Councils and Members of Parliament with a known interest in the subject, and the others were for press, radio and television journalists. This booklet contains the main points made by the principal speakers from the CEGB and BR. (The points covered include: brief description of the fuel cycle; effect of the fission process in producing plutonium and fission products in the fuel element; fuel transport; the fuel flasks; protection against accidents; experience of transporting fuel). (U.K.)

  3. Peach Bottom HTGR decommissioning and component removal

    International Nuclear Information System (INIS)

    Kohler, E.J.; Steward, K.P.; Iacono, J.V.

    1977-07-01

    The prime objective of the Peach Bottom End-of-Life Program was to validate specific HTGR design codes and predictions by comparison of actual and predicted physics, thermal, fission product, and materials behavior in Peach Bottom. Three consecutive phases of the program provide input to the HTGR design methods verifications: (1) Nondestructive fuel and circuit gamma scanning; (2) removal of steam generator and primary circuit components; and (3) Laboratory examinations of removed components. Component removal site work commenced with establishment of restricted access areas and installation of controlled atmosphere tents to retain relative humidity at <30%. A mock-up room was established to test and develop the tooling and to train operators under simulated working conditions. Primary circuit ducting samples were removed by trepanning, and steam generator access was achieved by a combination of arc gouging and grinding. Tubing samples were removed using internal cutters and external grinding. Throughout the component removal phase, strict health physics, safety, and quality assurance programs were implemented. A total of 148 samples of primary circuit ducting and steam generator tubing were removed with no significant health physics or safety incidents. Additionally, component removal served to provide access fordetermination of cesium plateout distribution by gamma scanning inside the ducts and for macroexamination of the steam generator from both the water and helium sides. Evaluations are continuing and indicate excellent performance of the steam generator and other materials, together with close correlation of observed and predicted fission product plateout distributions. It is concluded that such a program of end-of-life research, when appropriately coordinated with decommissioning activities, can significantly advance nuclear plant and fuel technology development

  4. Nuclear fuel management in JMTR

    International Nuclear Information System (INIS)

    Naka, Michihiro; Miyazawa, Masataka; Sato, Hiroshi; Nakayama, Fusao; Ito, Haruhiko

    1999-01-01

    The Japan Materials Testing Reactor (JMTR) is the largest scale materials (author)ted the fission gas release compared with the steady state opkW/l in Japan. JMTR as a multi-purpose reactor has been contributing to research and development on nuclear field with a wide variety of irradiation for performing engineering tests and safety research on fuel and component for light water reactor as well as fast breeder reactor, high temperature gas-cooled reactor etc., for research and development on blanket material for fusion reactor, for fundamental research, and for radio-isotope (RI) production. The driver nuclear fuel used in JMTR is aluminum based MTR type fuel. According to the Reduced Enrichment for Research and Test Reactors (RERTR) Program, the JMTR fuel elements had been converted from 93% high enriched uranium (HEU) fuel to 45% medium enriched uranium (MEU) fuel in 1986, and then to 20% low enriched uranium (LEU) fuel in 1994. The cumulative operation cycles until March 1999 reached to 127 cycles since the first criticality in 1968. JMTR has used 1,628 HEU, 688 MEU and 308 LEU fuel elements for these operation cycles. After these spent fuel elements were cooled in the JMTR water canal more than one year after discharged from the JMTR core, they had been transported to reprocessing plants in Europe, and then to plants in USA in order to extract the uranium remaining in the spent fuel. The JMTR spent fuel transportation for reprocessing had been continued until the end of 1988. However, USA had ceased spent fuel reprocessing in 1989, while USDOE committed to prepare an environmental review of the impacts of accepting spent fuels from foreign research reactors. After that, USDOE decided to implement a new acceptance policy in 1996, the spent fuel transportation from JMTR to Savannah River Site was commenced in 1997. It was the first transportation not only in Japan but in Asia also. Until resuming the transportation, the spent fuel elements stored in JMTR

  5. Fuel assembly for nuclear reactor

    International Nuclear Information System (INIS)

    Yamanaka, Akihiro; Haikawa, Katsumasa; Haraguchi, Yuko; Nakamura, Mitsuya; Aoyama, Motoo; Koyama, Jun-ichi.

    1996-01-01

    In a BWR type fuel assembly comprising first fuel rods filled with nuclear fission products and second fuel rods filled with burnable poisons and nuclear fission products, the concentration of the burnable poisons mixed to a portion of the second fuel rods is controlled so that it is reduced at the upper portion and increased at the lower portion in the axial direction. In addition, a product of the difference of an average concentration of burnable poisons between the upper portion and the lower portion and the number of fuel rods is determined to higher than a first set value determined corresponding to the limit value of a maximum linear power density. The sum of the difference of the average concentration of the burnable poisons between the upper portion and the lower portion of the second fuel rod and the number of the second fuel rods is determined to lower than a second set value determined corresponding to a required value of a surplus reactivity. If the number of the fuel rods mixed with the burnable poisons is increased, the infinite multiplication factor at an initial stage of the burning is lowered and, if the concentration of the mixed burnable poisons is increased, the time of exhaustion of the burnable poisons is delayed. As a result, the maximum value of the infinite multiplication factor is suppressed thereby enabling to control surplus reactivity. (N.H.)

  6. Inspection of nuclear fuel transport in Spain

    International Nuclear Information System (INIS)

    Lobo Mendez, J.

    1977-01-01

    The experience acquired in inspecting nuclear fuel shipments carried out in Spain will serve as a basis for establishing the regulations wich must be adhered to for future transports, as the transport of nuclear fuels in Spain will increase considerably within the next years as a result of the Spanish nuclear program. The experience acquired in nuclear fuel transport inspection is described. (author) [es

  7. Proliferation Resistant Nuclear Reactor Fuel

    International Nuclear Information System (INIS)

    Gray, L.W.; Moody, K.J.; Bradley, K.S.; Lorenzana, H.E.

    2011-01-01

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and

  8. 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

  9. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Takeda, Tadashi; Sato, Kenji; Goto, Masakazu.

    1984-01-01

    Purpose: To facilitate identification of a fuel assembly upon fuel exchange in BWR type reactors. Constitution: Fluorescent material is coated or metal plating is applied to the impressed portion of a upper tie plate handle of a fuel assembly, and the fluorescent material or the metal plating surface is covered with a protective membrane made of transparent material. This enables to distinguish the impressed surface from a distant place and chemical reaction between the impressed surface and the reactor water can be prevented. Furthermore, since the protective membrane is formed such that it protrudes toward the upper side relative to the impressed surface, there is no risk of depositions of claddings thereover. (Moriyama, K.)

  10. Nuclear fuel cladding material

    International Nuclear Information System (INIS)

    Nakahigashi, Shigeo.

    1982-01-01

    Purpose: To largely improve the durability and the safety of fuel cladding material. Constitution: Diffusion preventive layers, e.g., aluminum or the like are covered on both sides of a zirconium alloy base layer of thin material, and corrosion resistant layers, e.g., copper or the like are covered thereon. This thin plate material is intimately wound in a circularly tubular shape in a plurality of layers to form a fuel cladding tube. With such construction, corrosion of the tube due to fuel and impurity can be prevented by the corrosion resistant layers, and the diffusion of the corrosion resistant material to the zirconium alloy can be prevented by the diffusion preventive layers. Since a plurality of layers are cladded, even if the corrosion resistant layers are damaged or cracked due to stress corrosion, only one layer is damaged or cracked, but the other layers are not affected. (Sekiya, K.)

  11. Nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Busch, H.; Mindnich, F.R.

    1973-01-01

    The fuel rod consists of a can with at least one end cap and a plenum spring between this cap and the fuel. To prevent the hazard that a eutectic mixture is formed during welding of the end cap, a thermal insulation is added between the end cap and plenum spring. It consists of a comical extension of the end cap with a terminal disc against which the spring is supported. The end cap, the extension, and the disc may be formed by one or several pieces. If the disc is separated from the other parts it may be manufactured from chrome steel or VA steel. (DG) [de

  12. Modular nuclear fuel assembly rack

    International Nuclear Information System (INIS)

    Davis, C.J.

    1982-01-01

    A modular nuclear fuel assembly rack constructed of an array of identical cells, each cell constructed of a plurality of identical flanged plates. The unique assembly of the plates into a rigid rack provides a cellular compartment for nuclear fuel assemblies and a cavity between the cells for accepting neutron absorbing materials thus allowing a closely spaced array. The modular rack size can be easily adapted to conform with available storage space. U-shaped flanges at the edges of the plates are nested together at the intersection of four cells in the array. A bar is placed at the intersection to lock the cells together

  13. Spent nuclear fuel shipping basket

    International Nuclear Information System (INIS)

    Wells, A.H.

    1990-01-01

    This patent describes a basket for a cask for transporting nuclear fuel elements. It comprises: sleeve members, each of the sleeve members having interior cross-section dimensions for receiving a nuclear fuel assembly such that the assembly is restrained from lateral movement within the sleeve member, apertured disk members, means for axially aligning the apertures in the disk members, and means for maintaining the disk members in fixed spaced relationship to form a disk assembly, comprising an array of disks, the aligned apertures of the disks being adapted to receive the sleeve members and maintain them in fixed spaced relationship

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

    International Nuclear Information System (INIS)

    Duan Wuhua; Zhu Liyang; Zhu Yongjun; Xu Jingming

    2011-01-01

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

  15. European research and development on HTGR process heat applications

    International Nuclear Information System (INIS)

    Verfondern, Karl; Lensa, Werner von

    2003-01-01

    The High-Temperature Gas-Cooled Reactor represents a suitable and safe concept of a future nuclear power plant with the potential to produce process heat to be utilized in many industrial processes such as reforming of natural gas, coal gasification and liquefaction, heavy oil recovery to serve for the production of the storable commodities hydrogen or energy alcohols as future transportation fuels. The paper will include a description of the broad range of applications for HTGR process heat and describe the results of the German long-term projects ''Prototype Nuclear Process Heat Reactor Project'' (PNP), in which the technical feasibility of an HTGR in combination with a chemical facility for coal gasification processes has been proven, and ''Nuclear Long-Distance Energy Transportation'' (NFE), which was the demonstration and verification of the closed-cycle, long-distance energy transmission system EVA/ADAM. Furthermore, new European research initiatives are shortly described. A particular concern is the safety of a combined nuclear/chemical facility requiring a concept against potential fire and explosion hazards. (author)

  16. GCRA perspective on the HTGR-GT plant configuration

    International Nuclear Information System (INIS)

    1979-06-01

    Design specifications for the HTGR type reactor and gas turbine combination are presented concerning the turbomachinery; generator and isophase bus duct; PCRV and internals; heat exchangers; operability; maintenance; safety and licensing; core design; and fuel design

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-06-01

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

  18. Spent nuclear fuel in Bulgaria

    International Nuclear Information System (INIS)

    Peev, P.; Kalimanov, N.

    1999-01-01

    The development of the nuclear energy sector in Bulgaria is characterized by two major stages. The first stage consisted of providing a scientific basis for the programme for development of the nuclear energy sector in the country and was completed with the construction of an experimental water-water reactor. At present, spent nuclear fuel from this reactor is placed in a water filled storage facility and will be transported back to Russia. The second stage consisted of the construction of the 6 NPP units at the Kozloduy site. The spent nuclear fuel from the six units is stored in at reactor pools and in an additional on-site storage facility which is nearly full. In order to engage the government of the country with the on-site storage problems, the new management of the National Electric Company elaborated a policy on nuclear fuel cycle and radioactive waste management. The underlying policy is de facto the selection of the 'deferred decision' option for its spent fuel management. (author)

  19. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-02

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

  20. Fuel bundle for nuclear reactor

    International Nuclear Information System (INIS)

    Long, J.W.; Flora, B.S.; Ford, K.L.

    1977-01-01

    The invention concerns a new, simple and inexpensive system for assembling and dismantling a nuclear reactor fuel bundle. Several fuel rods are fitted in parallel rows between two retaining plates which secure the fuel rods in position and which are maintained in an assembled position by means of several stays fixed to the two end plates. The invention particularly refers to an improved apparatus for fixing the stays to the upper plate by using locking fittings secured to rotating sleeves which are applied against this plate [fr

  1. Regulating nuclear fuel waste

    International Nuclear Information System (INIS)

    1995-01-01

    When Parliament passed the Atomic Energy Control Act in 1946, it erected the framework for nuclear safety in Canada. Under the Act, the government created the Atomic Energy Control Board and gave it the authority to make and enforce regulations governing every aspect of nuclear power production and use in this country. The Act gives the Control Board the flexibility to amend its regulations to adapt to changes in technology, health and safety standards, co-operative agreements with provincial agencies and policy regarding trade in nuclear materials. This flexibility has allowed the Control Board to successfully regulate the nuclear industry for more than 40 years. Its mission statement 'to ensure that the use of nuclear energy in Canada does not pose undue risk to health, safety, security and the environment' concisely states the Control Board's primary objective. The Atomic Energy Control Board regulates all aspects of nuclear energy in Canada to ensure there is no undue risk to health, safety, security or the environment. It does this through a multi-stage licensing process

  2. HTGR-INTEGRATED COAL TO LIQUIDS PRODUCTION ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Anastasia M Gandrik; Rick A Wood

    2010-10-01

    As part of the DOE’s Idaho National Laboratory (INL) nuclear energy development mission, the INL is leading a program to develop and design a high temperature gas-cooled reactor (HTGR), which has been selected as the base design for the Next Generation Nuclear Plant. Because an HTGR operates at a higher temperature, it can provide higher temperature process heat, more closely matched to chemical process temperatures, than a conventional light water reactor. Integrating HTGRs into conventional industrial processes would increase U.S. energy security and potentially reduce greenhouse gas emissions (GHG), particularly CO2. This paper focuses on the integration of HTGRs into a coal to liquids (CTL) process, for the production of synthetic diesel fuel, naphtha, and liquefied petroleum gas (LPG). The plant models for the CTL processes were developed using Aspen Plus. The models were constructed with plant production capacity set at 50,000 barrels per day of liquid products. Analysis of the conventional CTL case indicated a potential need for hydrogen supplementation from high temperature steam electrolysis (HTSE), with heat and power supplied by the HTGR. By supplementing the process with an external hydrogen source, the need to “shift” the syngas using conventional water-gas shift reactors was eliminated. HTGR electrical power generation efficiency was set at 40%, a reactor size of 600 MWth was specified, and it was assumed that heat in the form of hot helium could be delivered at a maximum temperature of 700°C to the processes. Results from the Aspen Plus model were used to perform a preliminary economic analysis and a life cycle emissions assessment. The following conclusions were drawn when evaluating the nuclear assisted CTL process against the conventional process: • 11 HTGRs (600 MWth each) are required to support production of a 50,000 barrel per day CTL facility. When compared to conventional CTL production, nuclear integration decreases coal

  3. HTGR-Integrated Coal To Liquids Production Analysis

    International Nuclear Information System (INIS)

    Gandrik, Anastasia M.; Wood, Rick A.

    2010-01-01

    As part of the DOE's Idaho National Laboratory (INL) nuclear energy development mission, the INL is leading a program to develop and design a high temperature gas-cooled reactor (HTGR), which has been selected as the base design for the Next Generation Nuclear Plant. Because an HTGR operates at a higher temperature, it can provide higher temperature process heat, more closely matched to chemical process temperatures, than a conventional light water reactor. Integrating HTGRs into conventional industrial processes would increase U.S. energy security and potentially reduce greenhouse gas emissions (GHG), particularly CO2. This paper focuses on the integration of HTGRs into a coal to liquids (CTL) process, for the production of synthetic diesel fuel, naphtha, and liquefied petroleum gas (LPG). The plant models for the CTL processes were developed using Aspen Plus. The models were constructed with plant production capacity set at 50,000 barrels per day of liquid products. Analysis of the conventional CTL case indicated a potential need for hydrogen supplementation from high temperature steam electrolysis (HTSE), with heat and power supplied by the HTGR. By supplementing the process with an external hydrogen source, the need to 'shift' the syngas using conventional water-gas shift reactors was eliminated. HTGR electrical power generation efficiency was set at 40%, a reactor size of 600 MWth was specified, and it was assumed that heat in the form of hot helium could be delivered at a maximum temperature of 700 C to the processes. Results from the Aspen Plus model were used to perform a preliminary economic analysis and a life cycle emissions assessment. The following conclusions were drawn when evaluating the nuclear assisted CTL process against the conventional process: (1) 11 HTGRs (600 MWth each) are required to support production of a 50,000 barrel per day CTL facility. When compared to conventional CTL production, nuclear integration decreases coal consumption by 66

  4. World nuclear fuel cycle

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    A coloured pull-out wall chart is presented showing the fuel cycle interests of the world. Place names are marked and symbols are used to indicate regions associated with uranium or thorium deposits, mining, milling, enrichment, reprocessing and fabrication. (UK)

  5. Contracting for nuclear fuels

    International Nuclear Information System (INIS)

    Schuessler, C.M.

    1981-10-01

    This paper deals with uranium sales contracts, i.e. with contractual arrangements in the first steps of the fuel cycle, which cover uranium production and conversion. The various types of contract are described and, where appropriate, their underlying business philosophy and their main terms and conditions. Finally, the specific common features of such contracts are reviewed. (NEA) [fr

  6. Nuclear fuel cycle studies

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    For the metal-matrix encapsulation of radioactive waste, brittle-fracture, leach-rate, and migration studies are being conducted. For fuel reprocessing, annular and centrifugal contactors are being tested and modeled. For the LWBR proof-of-breeding project, the full-scale shear and the prototype dissolver were procured and tested. 5 figures

  7. Axially alignable nuclear fuel pellets

    International Nuclear Information System (INIS)

    Johansson, E.B.; Klahn, D.H.; Marlowe, M.O.

    1978-01-01

    An axially alignable nuclear fuel pellet of the type stacked in end-to-end relationship within a tubular cladding is described. Fuel cladding failures can occur at pellet interface locations due to mechanical interaction between misaligned fuel pellets and the cladding. Mechanical interaction between the cladding and the fuel pellets loads the cladding and causes increased cladding stresses. Nuclear fuel pellets are provided with an end structure that increases plastic deformation of the pellets at the interface between pellets so that lower alignment forces are required to straighten axially misaligned pellets. Plastic deformation of the pellet ends results in less interactions beween the cladding and the fuel pellets and significantly lowers cladding stresses. The geometry of pellets constructed according to the invention also reduces alignment forces required to straighten fuel pellets that are tilted within the cladding. Plastic deformation of the pellets at the pellet interfaces is increased by providing pellets with at least one end face having a centrally-disposed raised area of convex shape so that the mean temperature and shear stress of the contact area is higher than that of prior art pellets

  8. 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.

  9. Means for supporting nuclear fuel

    International Nuclear Information System (INIS)

    Cocker, P.; Price, M.A.

    1975-01-01

    Reference is made to means for supporting nuclear fuel pins in a reactor coolant channel and the problems that arise in this connection. For reasons of nuclear reactivity and neutron economy 'parasitic' material in a reactor core must be kept to a minimum, whilst for heat transfer reasons the use of fuel pins of large cross-sectional areas should be avoided. Fuel pins tend to be long thin objects having a can of minimum thickness and typically a pin may have a length/diameter ratio of about 500/1 and for fast reactor fuel pins, the outside diameter may be about 0.2 inch. The long slender pins must also be spaced very close together. A fast reactor fuel assembly may involve 200 to 300 fuel pins, each a few tenths of an inch in diameter, supported end on to coolant flowing up a channel of about 22 square inches in total area. The pins have a heavy metal oxide filling and require support. Details are given of a suitable method of support. Such support also allows withdrawal of pins from a fuel channel without the risk of breach of the can, after irradiation. (U.K.)

  10. 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.)

  11. New small HTGR power plant concept with inherently safe features - an engineering and economic challenge

    International Nuclear Information System (INIS)

    McDonald, C.F.; Sonn, D.L.

    1983-01-01

    Studies are in a very early design stage to establish a modular concept High-Temperature Gas-Cooled Reactor (HTGR) plant of about 100-MW(e) size to meet the special needs of small energy users in the industrialized and developing nations. The basic approach is to design a small system in which, even under the extreme conditions of loss of reactor pressure and loss of forced core cooling, the temperature would remain low enough so that the fuel would retain essentially all the fission products and the owner's investment would not be jeopardized. To realize economic goals, the designer faces the challenge of providing a standardized nuclear heat source, relying on a high percentage of factory fabrication to reduce site construction time, and keeping the system simple. While the proposed nuclear plant concept embodies new features, there is a large technology base to draw upon for the design of a small HTGR

  12. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    D'Eye, R.W.M.; Shennan, J.V.; Ford, L.H.

    1977-01-01

    Fuel element with particles from ceramic fissionable material (e.g. uranium carbide), each one being coated with pyrolitically deposited carbon and all of them being connected at their points of contact by means of an individual crossbar. The crossbar consists of silicon carbide produced by reaction of silicon metal powder with the carbon under the influence of heat. Previously the silicon metal powder together with the particles was kneaded in a solvent and a binder (e.g. epoxy resin in methyl ethyl ketone plus setting agent) to from a pulp. The reaction temperature lies at 1750 0 C. The reaction itself may take place in a nitrogen atmosphere. There will be produced a fuel element with a high overall thermal conductivity. (DG) [de

  13. HTGR generic technology program plan (FY 80)

    International Nuclear Information System (INIS)

    1980-01-01

    Purpose of the program is to develop base technology and to perform design and development common to the HTGR Steam Cycle, Gas Turbine, and Process Heat Plants. The generic technology program breaks into the base technology, generic component, pebble-bed study, technology transfer, and fresh fuel programs

  14. Spent nuclear fuel transport problems

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.I.

    1977-01-01

    The paper considers the problems of shipping spent fuel from nuclear power stations to reprocessing plants and also the principal ways of solving these problems with a view to achieving maximum economy and safety in transport. The increase in the number of nuclear power plants in the USSR will entail an intensification of spent-fuel shipments. Higher burnup and the need to reduce cooling time call for heavier and more complex shipping containers. The problem of shipping spent fuel should be tackled comprehensively, bearing in mind the requirements of safety and economy. One solution to these problems is to develop rational and cheap designs of such containers. In addition, the world-wide trend towards more thorough protection of the environment against pollution and of the health of the population requires the devotion of constant attention to improving the reliability and safety of shipments. The paper considers the prospects for nuclear power development in the USSR and in other member countries of the CMEA (1976-1980), the composition and design of some Soviet packaging assemblies, the appropriate cooling time for spent fuel from thermal reactor power stations, procedures for reducing fuel-shipping costs, some methodological problems of container calculation and design, and finally problems of testing and checking containers on test rigs. (author)

  15. Disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    1979-12-01

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste

  16. Fire resistant nuclear fuel cask

    International Nuclear Information System (INIS)

    Heckman, R.C.; Moss, M.

    1979-01-01

    The disclosure is directed to a fire resistant nuclear fuel cask employing reversibly thermally expansible bands between adjacent cooling fins such that normal outward flow of heat is not interfered with, but abnormal inward flow of heat is impeded or blocked

  17. Storage arrangements for nuclear fuel

    International Nuclear Information System (INIS)

    Ealing, C.J.

    1985-01-01

    A storage arrangement for nuclear fuel has a plurality of storage tubes connected by individual pipes to manifolds which are connected, in turn, to an exhaust system for maintaining the tubes at sub-atmospheric pressure, and means for producing a flow of a cooling fluid, such as air, over the exterior surfaces of the tubes. (author)

  18. World nuclear fuel cycle requirements 1991

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-10

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

  19. World nuclear fuel cycle requirements 1991

    International Nuclear Information System (INIS)

    1991-01-01

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

  20. Study Of Thorium As A Nuclear Fuel.

    Directory of Open Access Journals (Sweden)

    Prakash Humane

    2017-10-01

    Full Text Available Conventional fuel sources for power generation are to be replacing by nuclear power sources like nuclear fuel Uranium. But Uranium-235 is the only fissile fuel which is in 0.72 found in nature as an isotope of Uranium-238. U-238 is abundant in nature which is not fissile while U-239 by alpha decay naturally converted to Uranium- 235. For accompanying this nuclear fuel there is another nuclear fuel Thorium is present in nature is abundant can be used as nuclear fuel and is as much as safe and portable like U-235.

  1. High-temperature gas-cooled reactor (HTGR): long term program plan

    International Nuclear Information System (INIS)

    1980-01-01

    The FY 1980 effort was to investigate four technology options identified by program participants as potentially viable candidates for near-term demonstration: the Gas Turbine system (HTGR-GT), reflecting its perceived compatibility with the dry-cooling market, two systems addressing the process heat market, the Reforming (HTGR-R) and Steam Cycle (HTGR-SC) systems, and a more developmental reactor system, The Nuclear Heat Source Demonstration Reactor (NHSDR), which was to serve as a basis for both the HTGR-GT and HTGR-R systems as well as the further potential for developing advanced applications such as steam-coal gasification and water splitting

  2. Innovative microstructures in nuclear fuels

    International Nuclear Information System (INIS)

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

    2009-01-01

    For cleaner and safe nuclear power, new processes are required to design better nuclear fuels and make more efficient reactors to generate nuclear power. Therefore, one must understand how the microstructure changes during reactor operation. Accordingly, the materials scientists and engineers can then design and fabricate fuels with higher reliability and performance. Microstructure and its evolution are big unknowns in nuclear fuel. The basic requirements for the high performance of a fuel are: a) Soft pellets - To reduce Pellet clad mechanical interaction (PCMI) b) Large grain size - To reduce fission gas release (FGR). The strength of the pellet at room temperature is related to grain size by the Hall-Petch relation. Accordingly, the lower grain sized pellets will have high strength. But at high temperature (above equicohesive temperature) the grain boundaries becomes weaker than grain matrix. Since the small grain sized pellets have more grain boundary areas, these pellet become softer than pellet that have large grain sizes. Also as grain size decreases, creep rate of the fuel increases. Therefore, pellets with small grain size have higher creep rate and better plasticity. Therefore, these pellets will be useful to reduce the PCMI. On the other hand, pellet with large grain size is beneficial to reduce the fission gas release. In developing thermal reactor fuels for high burn-up, this factor should be taken into consideration. The question being asked is whether the microstructure can be tailored for irradiation hardening, fracture resistance, fission-gas release. This paper deals with the role played by microstructure for better irradiation performance. (author)

  3. Apparatus for locating defective nuclear fuel elements

    International Nuclear Information System (INIS)

    Lawrie, W.E.

    1979-01-01

    An ultrasonic search unit for locating defective fuel elements within a fuel assembly used in a water cooled nuclear reactor is presented. The unit is capable of freely traversing the restricted spaces between the fuel elements

  4. Fuel containing vessel for transporting nuclear fuel

    International Nuclear Information System (INIS)

    Yoshizawa, Hiroyasu; Shimizu, Fukuzo; Tanaka, Nobuyuki.

    1996-01-01

    A shock absorbing mechanism is disposed on an inner bottom of a vessel main body. The shock absorbing mechanism comprises a shock absorbing member disposed on the upper surface of a bottom wall, an annular metal plate disposed on the upper surface of the shock absorbing member and an annular spacer disposed on the upper surface of the metal plate. The shock absorbing member is made of a material such as of wood, lead, metal honeycomb or a metal mesh, which plastically deforms when applied with load higher than a predetermined level, and is formed in a square block-like form covering the upper surface of the bottom wall. The spacer is made of a thin soft material such as tetrafluoroethylene, and is formed in such a shape as capable of preventing direct contact of the lower end of the cylindrical member in a lower tie plate of nuclear fuels with the metal portion. This can ensure integrity of nuclear fuels even when they fall from a high place upon an assumed dropping accident. (I.N.)

  5. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Marmonier, Pierre; Mesnage, Bernard; Nervi, J.C.

    1975-01-01

    This invention refers to fuel assemblies for a liquid metal cooled fast neutron reactor. Each assembly is composed of a hollow vertical casing, of regular polygonal section, containing a bundle of clad pins filled with a fissile or fertile substance. The casing is open at its upper end and has a cylindrical foot at its lower end for positioning the assembly in a housing provided in the horizontal diagrid, on which the core assembly rests. A set of flat bars located on the external surface of the casing enables it to be correctly orientated in its housing among the other core assemblies [fr

  6. Safety-related Innovative Nuclear Reactor Technology Elements R and D (SINTER) Network and Global HTGR R and D Network (GHTRN). Strategic benefits of international networking

    International Nuclear Information System (INIS)

    Von Lensa, W.

    1998-01-01

    Action on 'Safety-related Innovative Nuclear Reactor Technology Elements - R and D - (SINTER) Network' both aim at the identification of priority items for sustainable innovations of nuclear technologies and work-shared European collaboration structures. Such an approach can also be realised for future R and D on HTGR-related R and D under the umbrella of the IAEA as already proposed by the 'International Working Group on Gas-Cooled Reactors (IWGGCR)' in 1996 and illustrated in this paper for the construction of a 'Global HTGR R and D Network (GHTRN)'. 3 refs

  7. Spent fuel management and closed nuclear fuel cycle

    International Nuclear Information System (INIS)

    Kudryavtsev, E.G.

    2012-01-01

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

  8. Uranium - the nuclear fuel

    International Nuclear Information System (INIS)

    Smith, E.E.N.

    1976-01-01

    A brief history is presented of Canadian uranium exploration, production, and sales. Statistics show that Canada is a good customer for its own uranium due to a rapidly expanding nuclear power program. Due to an average 10 year lag between commencement of exploration and production, and with current producers sold out through 1985, it is imperative that exploration efforts be increased. (E.C.B.)

  9. Storage arrangements for nuclear fuel

    International Nuclear Information System (INIS)

    Deacon, D.

    1982-01-01

    A storage arrangement for spent nuclear fuel either irradiated or pre-irradiated or for vitrified waste after spent fuel reprocessing, comprises a plenum chamber which has a base pierced by a plurality of openings each of which has sealed to it an open topped tube extending downwards and closed at its lower end. The plenum chamber, with the tubes, forms an air-filled enclosure associated with an exhaust system for exhausting air from the system through filters to maintain the interior of the enclosure at sub-atmospheric pressure. The tubes are arranged to accommodate the stored fuel and the arrangement includes a means for producing a flow of cooling air over the exterior of the tubes so that the latter effectively form a plurality of heat exchangers in close proximity to the fuel. The air may be caused to flow over the tube surfaces by a natural thermosyphon process. (author)

  10. Nuclear fuel and energy policy

    International Nuclear Information System (INIS)

    Ahmed, S.B.

    1979-01-01

    This book examines the uranium resource situation in relation to the future needs of the nuclear economy. Currently the United States is the world's leading producer and consumer of nuclear fuels. In the future US nuclear choices will be highly interdependent with the rest of the world as other countries begin to develop their own nuclear programs. Therefore the world's uranium resource availability has also been examined in relation to the expected growth in the world nuclear industry. Based on resource evaluation, the study develops an economic framework for analyzing and describing the behavior of the US uranium mining and milling industry. An econometric model designed to reflect the underlying structure of the physical processes of the uranium mining and milling industry has been developed. The purpose of this model is to forecast uranium prices and outputs for the period 1977 to 2000. Because uncertainty has sometimes surrounded the economic future of the uranium markets, the results of the econometric modeling should be interpreted with great care and restrictive assumptions. Another aspect of this study is to provide much needed information on the operations of government-owned enrichment plants and the practices used by the government in the determination of fuel enrichment costs. This study discusses possible future developments in enrichment supply and technologies and their implications for future enrichment costs. A review of the operations involving the uranium concentrate conversion to uranium hexafluoride and fuel fabrication is also provided. An economic analysis of these costs provides a comprehensive view of the front-end costs of the nuclear fuel cycle

  11. Ceramics as nuclear reactor fuels

    International Nuclear Information System (INIS)

    Reeve, K.D.

    1975-01-01

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

  12. Fuel assemblies for nuclear reactor

    International Nuclear Information System (INIS)

    Nishi, Akihito.

    1987-01-01

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

  13. Development of high burnup nuclear fuel technology

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone

    1987-04-01

    The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country

  14. Sufficiency of the Nuclear Fuel

    International Nuclear Information System (INIS)

    Pevec, D.; Knapp, V.; Matijevic, M.

    2008-01-01

    Estimation of the nuclear fuel sufficiency is required for rational decision making on long-term energy strategy. In the past an argument often invoked against nuclear energy was that uranium resources are inadequate. At present, when climate change associated with CO 2 emission is a major concern, one novel strong argument for nuclear energy is that it can produce large amounts of energy without the CO 2 emission. Increased interest in nuclear energy is evident, and a new look into uranium resources is relevant. We examined three different scenarios of nuclear capacity growth. The low growth of 0.4 percent per year in nuclear capacity is assumed for the first scenario. The moderate growth of 1.5 percent per year in nuclear capacity preserving the present share in total energy production is assumed for the second scenario. We estimated draining out time periods for conventional resources of uranium using once through fuel cycle for the both scenarios. For the first and the second scenario we obtained the draining out time periods for conventional uranium resources of 154 years and 96 years, respectively. These results are, as expected, in agreement with usual evaluations. However, if nuclear energy is to make a major impact on CO 2 emission it should contribute much more in the total energy production than at present level of 6 percent. We therefore defined the third scenario which would increase nuclear share in the total energy production from 6 percent in year 2020 to 30 percent by year 2060 while the total world energy production would grow by 1.5 percent per year. We also looked into the uranium requirement for this scenario, determining the time window for introduction of uranium or thorium reprocessing and for better use of uranium than what is the case in the once through fuel cycle. The once through cycle would be in this scenario sustainable up to about year 2060 providing most of the expected but undiscovered conventional uranium resources were turned

  15. Method of manufacturing nuclear fuel pellet

    International Nuclear Information System (INIS)

    Oguma, Masaomi; Masuda, Hiroshi; Hirai, Mutsumi; Tanabe, Isami; Yuda, Ryoichi.

    1989-01-01

    In a method of manufacturing nuclear fuel pellets by compression molding an oxide powder of nuclear fuel material followed by sintering, a metal nuclear material is mixed with an oxide powder of the nuclear fuel material. As the metal nuclear fuel material, whisker or wire-like fine wire or granules of metal uranium can be used effectively. As a result, a fuel pellet in which the metal nuclear fuel is disposed in a network-like manner can be obtained. The pellet shows a great effect of preventing thermal stress destruction of pellets upon increase of fuel rod power as compared with conventional pellets. Further, the metal nuclear fuel material acts as an oxygen getter to suppress the increase of O/M ratio of the pellets. Further, it is possible to reduce the swelling of pellet at high burn-up degree. (T.M.)

  16. Nuclear fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Armijo, J S; Coffing, L F

    1979-04-05

    The fuel element with circular cross-section for BWR and PWR consists of a core surrounded by a compound jacket container where there is a gap between the core and jacket during operation in the reactor. The core consists of U, Pu, Th compounds and mixtures of these. The compound jacket consists of zircaloy 2 or 4. In order to for example prevent the corrosion of the compound jacket, its inner surface has a metal barrier with smaller neutron absorbers than the jacket material in the form of a zirconium sponge. The zirconium of this metal barrier has impurities of various elements in the order of magnitude of 1000 to 5000 ppm. The oxygen content is in the range of 200 to 1200 ppm and the thickness of the metal barrier is 1-30% of the thickness of the jacket.

  17. Cooling nuclear reactor fuel

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1975-01-01

    Reference is made to water or water/steam cooled reactors of the fuel cluster type. In such reactors it is usual to mount the clusters in parallel spaced relationship so that coolant can pass freely between them, the coolant being passed axially from one end of the cluster in an upward direction through the cluster and being effective for cooling under normal circumstances. It has been suggested, however, that in addition to the main coolant flow an auxiliary coolant flow be provided so as to pass laterally into the cluster or be sprayed over the top of the cluster. This auxiliary supply may be continuously in use, or may be held in reserve for use in emergencies. Arrangements for providing this auxiliary cooling are described in detail. (U.K.)

  18. HTGR safety research program

    International Nuclear Information System (INIS)

    Barsell, A.W.; Olsen, B.E.; Silady, F.A.

    1981-01-01

    An HTGR safety research program is being performed supporting and guided in priorities by the AIPA Probabilistic Risk Study. Analytical and experimental studies have been conducted in four general areas where modeling or data assumptions contribute to large uncertainties in the consequence assessments and thus, in the risk assessment for key core heat-up accident scenarios. Experimental data have been obtained on time-dependent release of fission products from the fuel particles, and plateout characteristics of condensible fission products in the primary circuit. Potential failure modes of primarily top head PCRV components as well as concrete degradation processes have been analyzed using a series of newly developed models and interlinked computer programs. Containment phenomena, including fission product deposition and potential flammability of liberated combustible gases have been studied analytically. Lastly, the behaviour of boron control material in the core and reactor subcriticality during core heatup have been examined analytically. Research in these areas has formed the basis for consequence updates in GA-A15000. Systematic derivation of future safety research priorities is also discussed. (author)

  19. Fuel element for nuclear reactors

    International Nuclear Information System (INIS)

    Cadwell, D.J.

    1982-01-01

    The invention concerns a fuel element for nuclear reactors with fuel rods and control rod guide tubes, where the control rod guide tubes are provided with flat projections projecting inwards, in the form of local deformations of the guide tube wall, in order to reduce the radial play between the control rod concerned and the guide tube, and to improve control rod movement. This should ensure that wear on the guide tubes is largely prevented which would be caused by lateral vibration of the control rods in the guide tubes, induced by the flow of coolant. (orig.) [de

  20. Spent nuclear fuel sampling strategy

    International Nuclear Information System (INIS)

    Bergmann, D.W.

    1995-01-01

    This report proposes a strategy for sampling the spent nuclear fuel (SNF) stored in the 105-K Basins (105-K East and 105-K West). This strategy will support decisions concerning the path forward SNF disposition efforts in the following areas: (1) SNF isolation activities such as repackaging/overpacking to a newly constructed staging facility; (2) conditioning processes for fuel stabilization; and (3) interim storage options. This strategy was developed without following the Data Quality Objective (DQO) methodology. It is, however, intended to augment the SNF project DQOS. The SNF sampling is derived by evaluating the current storage condition of the SNF and the factors that effected SNF corrosion/degradation

  1. Grids for nuclear fuel elements

    International Nuclear Information System (INIS)

    Nicholson, G.

    1980-01-01

    This invention relates to grids for nuclear fuel assemblies with the object of providing an improved grid, tending to have greater strength and tending to offer better location of the fuel pins. It comprises sets of generally parallel strips arranged to intersect to define a structure of cellular form, at least some of the intersections including a strip which is keyed to another strip at more than one point. One type of strip may be dimpled along its length and another type of strip may have slots for keying with the dimples. (Auth.)

  2. Nuclear fuel element

    International Nuclear Information System (INIS)

    Watarumi, Kazutoshi.

    1992-01-01

    Hollow fuel pellets are piled at multi-stages in a cladding tube to form a pellet stack. A bundle of metal fine wires made of zirconium or an alloy thereof is inserted passing through the hollow portion of each of the hollow pellets over a length of the pellet stack. The metal fine wires are bundled by securing ring at a joining portions of the pellets. Then, the portion between both of adjacent rings is expanded radially and has a spring function biasing in the radial direction. With such a constitution, even if the pellet is expanded radially due to pallet gas swelling, the hollow portion is not closed, and the gas flow channel is ensured. In addition, even if the pellet is cracked due to thermal shocks, the pellet piece is prevented from dropping to the hollow portion. In this case, the thermal conduction between the pellets and the cladding tube is kept satisfactorily by the spring function of the metal wire bundle. (I.N.)

  3. Coal and nuclear electricity fuels

    International Nuclear Information System (INIS)

    Rahnama, F.

    1982-06-01

    Comparative economic analysis is used to contrast the economic advantages of nuclear and coal-fired electric generating stations for Canadian regions. A simplified cash flow method is used with present value techniques to yield a single levelized total unit energy cost over the lifetime of a generating station. Sensitivity analysis illustrates the effects of significant changes in some of the cost data. The analysis indicates that in Quebec, Ontario, Manitoba and British Columbia nuclear energy is less costly than coal for electric power generation. In the base case scenario the nuclear advantage is 24 percent in Quebec, 29 percent in Ontario, 34 percent in Manitoba, and 16 percent in British Columbia. Total unit energy cost is sensitive to variations in both capital and fuel costs for both nuclear and coal-fuelled power stations, but are not very sensitive to operating and maintenance costs

  4. Nuclear fuel shipping inspection device

    International Nuclear Information System (INIS)

    Takahashi, Toshio; Hada, Koji.

    1988-01-01

    Purpose: To provide an nuclear fuel shipping inspection device having a high detection sensitivity and capable of obtaining highly reliable inspection results. Constitution: The present invention concerns a device for distinguishing a fuel assembly having failed fuel rods in LMFBR type reactors. Coolants in a fuel assembly to be inspected are collected by a sampling pipeway and transferred to a filter device. In the filter device, granular radioactive corrosion products (CP) in the coolants are captured, to reduce the background. The coolants, after being passed through the filter device, are transferred to an FP catching device and gamma-rays of iodine and cesium nuclides are measured in FP radiation measuring device. Subsequently, the coolants transferred to a degasing device to separate rare gas FP in the coolants from the liquid phase. In a case if rare gas fission products are detected by the radiation detector, it means that there is a failed fuel rod in the fuel assembly to be inspected. Since the CP and the soluble FP are separated and extracted for the radioactivity measurement, the reliability can be improved. (Kamimura, M.)

  5. Nuclear Fuels: Present and Future

    Directory of Open Access Journals (Sweden)

    Donald R. Olander

    2009-02-01

    Full Text Available The important new developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of these fuels and the reactors they power are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel-rod designs, the hydride fuel with liquid metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the Very High Temperature Reactor and the Sodium Fast Reactor, and the accompanying reprocessing technologies, aqueous-based UREX and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the material's behavior under irradiation and in the reprocessing schemes are emphasized.

  6. Nuclear fuel pellet transfer escalator

    International Nuclear Information System (INIS)

    Huggins, T.B. Sr.; Roberts, E.; Edmunds, M.O.

    1991-01-01

    This patent describes a nuclear fuel pellet escalator for loading nuclear fuel pellets into a sintering boat. It comprises a generally horizontally-disposed pellet transfer conveyor for moving pellets in single file fashion from a receiving end to a discharge end thereof, the conveyor being mounted about an axis at its receiving end for pivotal movement to generally vertically move its discharge end toward and away from a sintering boat when placed below the discharge end of the conveyor, the conveyor including an elongated arm swingable vertically about the axis and having an elongated channel recessed below an upper side of the arm and extending between the receiving and discharge ends of the conveyor; a pellet dispensing chute mounted to the arm of the conveyor at the discharge end thereof and extending therebelow such that the chute is carried at the discharge end of the conveyor for generally vertical movement therewith toward and away from the sintering boat

  7. Nuclear reactor fuel element assemblies

    International Nuclear Information System (INIS)

    Raven, L.F.

    1975-01-01

    A spacer grid for a nuclear fuel element comprises a plurality of cojointed cylindrical ferrules adapted to receive a nuclear fuel pin. Each ferrule has a pair of circumferentially spaced rigid stop members extending inside the ferrule and a spring locating member attached to the ferrule and also extending from the ferrule wall inwardly thereof at such a circumferential spacing relative to the rigid stop members that the line of action of the spring locating member passes in opposition to and between the rigid stop members which lie in the same diametric plane. At least some of the cylindrical ferrules have one rim shaped to promote turbulence in fluid flowing through the grid. (Official Gazette)

  8. HTGR market assessment: interim report

    International Nuclear Information System (INIS)

    1979-09-01

    The purpose of this Assessment is to establish the utility perspective on the market potential of the HTGR. The majority of issues and conclusions in this report are applicable to both the HTGR-Gas Turbine (GT) and the HTGR-Steam Cycle (SC). This phase of the HTGR Market Assessment used the HTGR-GT as the reference design as it is the present focus of the US HTGR Program. A brief system description of the HTGR-GT is included in Appendix A. This initial report provides the proposed structure for conducting the HTGR Market Assessment plus preliminary analyses to establish the magnitude and nature of key factors that affect the HTGR market. The HTGR market factors and their relationship to the present HTGR Program are discussed. This report discusses two of these factors in depth: economics and water availability. The water availability situation in the US and its impact on the potential HTGR market are described. The approach for applying the HTGR within a framework of utility systems analyses is presented

  9. Interfaces in ceramic nuclear fuels

    International Nuclear Information System (INIS)

    Reeve, K.D.

    Internal interfaces in all-ceramic dispersion fuels (such as these for HTGRs) are discussed for two classes: BeO-based dispersions, and coated particles for graphite-based fuels. The following points are made: (1) The strength of a two-phase dispersion is controlled by the weaker dispersed phase bonded to the matrix. (2) Differential expansion between two phases can be controlled by an intermediate buffer zone of low density. (3) A thin ceramic coating should be in compression. (4) Chemical reaction between coating and substrate and mass transfer in service should be minimized. The problems of the nuclear fuel designer are to develop coatings for fission product retention, and to produce radiation-resistant interfaces. 44 references, 18 figures

  10. Electrochemical reprocessing of nuclear fuels

    International Nuclear Information System (INIS)

    Brambilla, G.; Sartorelli, A.

    1980-01-01

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

  11. Storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Machado, O.J.; Moore, J.T.; Cooney, B.F.

    1989-01-01

    This patent describes a rack for storing nuclear fuel assemblies. The rack including a base, an array of side-by-side fuel-storage locations, each location being a hollow body of rectangular transverse cross section formed of metallic sheet means which is readily bent, each body having a volume therein dimensioned to receive a fuel assembly. The bodies being mounted on the base with each body secured to bodies adjacent each body along welded joints, each joint joining directly the respective contiguous corners of each body and of bodies adjacent to each body and being formed by a series of separate welds spaced longitudinally between the tops and bottoms of the secured bodies along each joint. The spacings of the separate welds being such that the response of the rack when it is subjected to the anticipated seismic acceleration of the rack, characteristic of the geographical regions where the rack is installed, is minimized

  12. Fuel assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    In a nuclear fuel assembly, hollow guide posts protrude into a fuel assembly and fitting grill from a biased spring pad with a plunger that moves with the spring pad plugging one end of each of the guide posts. A plate on the end fitting grill that has a hole for fluid discharge partially plugs the other end of the guide post. Pressurized water coolant that fills the guide post volume acts as a shock absorber and should the reactor core receive a major seismic or other shock, the fuel assembly is compelled to move towards a pad depending from a transversely disposed support grid. The pad bears against the spring pad and the plunger progressively blocks the orifices provided by slots in the guide posts thus gradually absorbing the applied shock. After the orifice has been completely blocked, controlled fluid discharge continues through a hole coil spring cooperating in the attenuation of the shock. (author)

  13. Inserts for nuclear fuel elements

    International Nuclear Information System (INIS)

    Cragg, P.J.

    1982-01-01

    An insert for a nuclear fuel pin which comprises a strip. The strip carries notches, which enable a coding arrangement to be carried on the strip. The notches may be of differing sizes and the coding on the strip includes identification and identification checking data. Each notch on the strip may give rise to a signal pulse which is counted by a detector to avoid errors. (author)

  14. Nuclear fuel element and container

    International Nuclear Information System (INIS)

    Grubb, W.T.; King, L.H.

    1981-01-01

    The invention is based on the discovery that a substantial reduction in metal embrittlement or stress corrosion cracking from fuel pellet-cladding interaction can be achieved by the use of a copper layer or liner in proximity to the nuclear fuel, and an intermediate zirconium oxide barrier layer between the copper layer and the zirconium cladding substrate. The intermediate zirconia layer is a good copper diffusion barrier; also, if the zirconium cladding surface is modified prior to oxidation, copper can be deposited by electroless plating. A nuclear fuel element is described which comprises a central core of fuel material and an elongated container using the system outlined above. The method for making the container is again described. It comprises roughening or etching the surface of the zirconium or zirconium alloy container, oxidizing the resulting container, activating the oxidized surface to allow for the metallic coating of such surfaces by electroless deposition and further coating the activated-oxidized surface of the zirconium or zirconium alloy container with copper, iron or nickel or an alloy thereof. (U.K.)

  15. Modeling the Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  16. Spent nuclear fuel storage vessel

    International Nuclear Information System (INIS)

    Watanabe, Yoshio; Kashiwagi, Eisuke; Sekikawa, Tsutomu.

    1997-01-01

    Containing tubes for containing spent nuclear fuels are arranged vertically in a chamber. Heat releasing fins are disposed horizontal to the outer circumference of the containing tubes for rectifying cooling air and promoting cooling of the containing tubes. Louvers and evaporation sides of heat pipes are disposed at a predetermined distance in the chamber. Cooling air flows from an air introduction port to the inside of the chamber and takes heat from the containing tubes incorporated with heat generating spent nuclear fuels, rising its temperature and flows off to an air exhaustion exit. The direction for the rectification plate of the louver is downward from a horizontal position while facing to the air exhaustion port. Since the evaporation sides of the heat pipes are disposed in the inside of the chamber and the condensation side of the heat pipes is disposed to the outside of the chamber, the thermal energy can be recovered from the containing tubes incorporated with spent nuclear fuels and utilized. (I.N.)

  17. Analytical methods for fissionable materials in the nuclear fuel cycle. Covering June 1974--June 1975

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1975-10-01

    Research progress is reported on method development for the dissolution of difficult-to-dissolve materials, the automated analysis of plutonium and uranium, the preparation of plutonium materials for the Safeguard Analytical Laboratory Evaluation (SALE) Program, and the analysis of HTGR fuel and SALE uranium materials. The previously developed Teflon-container, metal-shell apparatus was applied to the dissolution of various nuclear materials. Gas--solid reactions, mainly using chlorine at elevated temperatures, are promising for separating uranium from refractory compounds. An automated spectrophotometer designed for determining plutonium and uranium was tested successfully. Procedures were developed for this instrument to analyze uranium--plutonium mixtures and the effects of diverse ions upon the analysis of plutonium and uranium were further established. A versatile apparatus was assembled to develop electrotitrimetric methods that will serve as the basis for precise automated determinations of plutonium. Plutonium materials prepared for the Safeguard Analytical Laboratory Evaluation (SALE) Program were plutonium oxide, uranium--plutonium mixed oxide, and plutonium metal. Improvements were made in the methods used for determining uranium in HTGR fuel materials and SALE uranium materials. Plutonium metal samples were prepared, characterized, and distributed, and half-life measurements were in progress as part of an inter-ERDA-laboratory program to measure accurately the half-lives of long-lived plutonium isotopes

  18. Regulatory Framework of Safety for HTGR

    International Nuclear Information System (INIS)

    Huh, Chang Wook; Suh, Nam Duk

    2011-01-01

    Recent accident in Fukushima Daiichi plant in Japan makes big impacts on the future of nuclear business. Many countries are changing their nuclear projects and increased safety of nuclear plants is asked for from the public. Without providing safety the society accepts, it might be almost impossible to build new plants further. In this sense high temperature gas-cooled reactor (HTGR) which is under development needs to be licensed reflecting this new expectation regarding safety. It means we should have higher level of safety goal and a systematic regulatory framework to assure the safety. In our previous paper, we evaluated the current safety goal and design practice in view of this new safety expectation after Fukushima accident. It was argued that a top-down approach starting from safety goal is necessary to develop safety requirements or to assure safety. Thus we need to propose an ultimate safety goal public accepts and then establish a systematic regulatory framework. In this paper we are going to provide a conceptual regulatory framework to guarantee the safety of HTGR. Section 2 discusses the recent trend of IAEA safety requirements and then summarize the HTGR design approach. Incorporating these discussions, we propose a conceptual framework of regulation for safety of HTGR

  19. Modification in fuel processing of Mitsubishi Nuclear Fuel's Tokai Works

    International Nuclear Information System (INIS)

    1976-01-01

    Results of the study by the Committee for Examination of Fuel Safety, reported to the AEC of Japan, are presented, concerning safety of the modifications of Tokai Works, Mitsubishi Nuclear Fuel Co., Ltd. Safety has been confirmed thereof. The modifications covered are the following: storage facility of nuclear fuel in increase, analytical facility in transfer, fuel assemblage equipment in addition, incineration facility of combustible solid wastes in installation, experimental facility of uranium recovery in installation, and warehouse in installation. (Mori, K.)

  20. Strategies of management of the nuclear fuel

    International Nuclear Information System (INIS)

    Leon, J.R.; Perez, A.; Filella, J.M.

    1996-01-01

    The management of nuclear fuel is depending on several factors: - Regulatory commission. The enterprises owner of the NPPs.The enterprise owner of the energy distribution. These factors are considered for the management of nuclear fuel. The design of fuel elements, the planning of cycles, the design of core reactors and the costs are analyzed. (Author)

  1. 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

  2. On the nuclear fuel and fossil fuel reserves

    International Nuclear Information System (INIS)

    Fettweis, G.

    1978-01-01

    A short discussion of the nuclear fuel and fossil fuel reserves and the connected problem of prices evolution is presented. The need to regard fuel production under an economic aspect is emphasized. Data about known and assessed fuel reserves, world-wide and with special consideration of Austria, are reviewed. It is concluded that in view of the fuel reserves situation an energy policy which allows for a maximum of options seems adequate. (G.G.)

  3. Nuclear fuels - swords and ploughshares

    Energy Technology Data Exchange (ETDEWEB)

    Franklin, N.L.

    1986-05-01

    In 1986 the problems associated with the implementation of nuclear power programmes mainly arise from difficulties of social acceptability. The scientific and technological achievements are no longer a source of wonder and are taken for granted by a public which has become accustomed to such achievements in other fields. This lecture recounts the history of the nuclear fuel cycle starting around 1955 but continuing, to look at future prospects. The problems are discussed. The technical improvements that have occurred over the years mean that, currently it is possible for all the problems to be overcome technically. Although there is always room for improvements in endurance, design etc. commercial and safety requirements can be met. In economic terms, the real costs of the fuel cycle have reached a plateau and should decrease as the result of lower cost for enriched uranium, lower reprocessing costs and better fuel management. However, in social and political terms, the position is not so certain because of public concern about reprocessing plants and the disposal of radioactive wastes. (U.K.).

  4. Reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

    Schmitt, D.

    1985-01-01

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

  5. An introduction to the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Leuze, R.E.

    1986-01-01

    This overview of the nuclear fuel cycle is divided into three parts. First, is a brief discussion of the basic principles of how nuclear reactors work;second, is a look at the major types of nuclear reactors being used and world-wide nuclear capacity;and third, is an overview of the nuclear fuel cycle and the present industrial capability in the US. 34 figs., 10 tabs

  6. Potential of the HTGR hydrogen cogeneration system in Japan

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo; Mouri, Tomoaki; Kunitomi, Kazuhiko

    2007-01-01

    A high temperature gas cooled reactor (HTGR) is one of the next generation nuclear systems. The HTGR hydrogen cogeneration system can produce not only electricity but also hydrogen. Then it has a potential to supply massive low-cost hydrogen without greenhouse gas emission for the future hydrogen society. Japan Atomic Energy Agency (JAEA) has been carried out the design study of the HTGR hydrogen cogeneration system (GTHTR300C). The thermal power of the reactor is 600 MW. The hydrogen production plant utilizes 370 MW and can supply 52,000 m 3 /h (0.4 Bm 3 /y) of hydrogen. Present industrial hydrogen production capacity in Japan is about 18 Bm 3 /y and it will decrease by 15 Bm 3 /y in 2030 due to the aging facilities. On the other hand, the hydrogen demand for fuel cell vehicle (FCV) in 2030 is estimated at 15 Bm 3 /y at a maximum. Since the hydrogen supply may be short after 2030, the additional hydrogen should be produced by clean hydrogen process to reduce greenhouse gas emission. This hydrogen shortage is a potential market for the GTHTR300C. The hydrogen production cost of GTHTR300C is estimated at 20.5 JPY/Nm 3 which has an economic competitiveness against other industrial hydrogen production processes. 38 units of the GTHTR300C can supply a half of this shortage which accounts for the 33% of hydrogen demand for FCV in 2100. According to the increase of hydrogen demand, the GTHTR300C should be constructed after 2030. (author)

  7. Critical review of nuclear fuel cycle

    International Nuclear Information System (INIS)

    Kuster, N.

    1996-01-01

    Transmutation of long-lived radionuclides is considered as an alternative to the in-depth disposal of spent nuclear fuel, in particular, on the final stage of the nuclear fuel cycle. The majority of conclusions is the result of the common work of the Karlsruhe FZK and the Commissariat on nuclear energy of France (CEA)

  8. Determining fissile content of nuclear fuel elements

    International Nuclear Information System (INIS)

    Arya, S.P.; Grossman, L.N.; Schoenig, F.C.

    1980-01-01

    This invention relates to the determination of the fissile fuel content of fuel for nuclear reactors. A nondestructive method is described for determining rapidly, accurately and simultaneously the fissile content, enrichment and location of fuel material which may also contain amounts of burnable poison, by detecting the γ-rays emitted from the fuel material due to natural radioactive decay. (U.K.)

  9. Transport and reprocessing of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Lenail, B.

    1981-01-01

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

  10. Fuel optimization of Qinshan nuclear power plant

    International Nuclear Information System (INIS)

    Liao Zejun; Li Zhuoqun; Kong Deping; Xue Xincai; Wang Shiwei

    2010-01-01

    Based on the design practice of the fuel replacement of Qin Shan nuclear power plant, this document effectively analyzes the shortcomings of current replacement design of Qin Shan. To address these shortcomings, this document successfully implements the 300 MW fuel optimization program from fuel replacement. fuel improvement and experimentation ,and achieves great economic results. (authors)

  11. 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.)

  12. High-temperature process heat applications with an HTGR

    International Nuclear Information System (INIS)

    Quade, R.N.; Vrable, D.L.

    1980-04-01

    An 842-MW(t) HTGR-process heat (HTGR-PH) design and several synfuels and energy transport processes to which it could be coupled are described. As in other HTGR designs, the HTGR-PH has its entire primary coolant system contained in a prestressed concrete reactor vessel (PCRV) which provides the necessary biological shielding and pressure containment. The high-temperature nuclear thermal energy is transported to the externally located process plant by a secondary helium transport loop. With a capability to produce hot helium in the secondary loop at 800 0 C (1472 0 F) with current designs and 900 0 C (1652 0 F) with advanced designs, a large number of process heat applications are potentially available. Studies have been performed for coal liquefaction and gasification using nuclear heat

  13. Nuclear fuel cycle modelling using MESSAGE

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  14. Method of producing nuclear fuels

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Suzuki, Tokuyuki; Oomura, Hiroshi.

    1985-01-01

    Purpose: To fabricate a nuclear fuel assembly with uniform enrichment degree, in the blanket of a hybrid reactor. Constitution: A vessel charged with powderous source materials is conveyed by a conveying gas through a material charge/discharge tube to the inside of the blanket. Then, plasmas are formed in the inner space of the blanket so as to enrich the source materials by the irradiation of neutrons. After the average degree of enrichment reaches a predetermined level, the material vessel is discharged by the conveying gas onto a conveyor. The powder materials are separated from the source-material vessel and then charged into a source-material hopper. The mixed material of a uniform enrichment degree is supplied to a fuel-assembly-fabrication device. FP gases resulted after the enrichment are effectively separated and removed through an FP gas pipe. (Horiuchi, T.)

  15. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

    Kee, R.W.; Denero, J.V.

    1975-01-01

    An apparatus for loading nuclear fuel pellets on trays for transfer in a system is described. A conveyor supplies pellets from a source to a loading station. When the pellets reach a predetermined position at the loading station, a manual or automatically operated arm pushes the pellets into slots on a tray and this process is repeated until pellet sensing switches detect that the tray is full. Thereupon, the tray is lowered onto a belt or other type conveyor and transferred to other apparatus in the system, such as a furnace for sintering, and in some cases, reduction of UO 2 . 2 to UO 2 . The pellets are retained on the tray and subsequently loaded directly into fuel rods to be used in the reactor core. (auth)

  16. Nuclear fuel pellet production method and nuclear fuel pellet

    International Nuclear Information System (INIS)

    Yuda, Ryoichi; Ito, Ken-ichi; Masuda, Hiroshi.

    1993-01-01

    In a method of manufacturing nuclear fuel pellets by compression-molding UO 2 powders followed by sintering, a sintering agent having a composition of about 40 to 80 wt% of SiO 2 and the balance of Al 2 O 3 , a sintering agent at a ratio of 10 to 500 ppm based on the total amount of UO 2 and UO 2 powders are mixed, compression molded and then sintered at a sintering temperature of about 1500 of 1800degC. The UO 2 particles have an average grain size of about 20 to 60μm, most of the crystal grain boundary thereof is coated with a glassy or crystalline alumina silicate phase, and the porosity is about 1 to 4 vol%. With such a constitution, the sintering agent forms a single liquid phase eutectic mixture during sintering, to promote a surface reaction between nuclear fuel powders by a liquid phase sintering mechanism, increase their density and promote the crystal growth. Accordingly, it is possible to lower the softening temperature, improve the creep velocity of the pellets and improve the resistance against pellet-clad interaction. (T.M.)

  17. HTGR Application Economic Model Users' Manual

    Energy Technology Data Exchange (ETDEWEB)

    A.M. Gandrik

    2012-01-01

    The High Temperature Gas-Cooled Reactor (HTGR) Application Economic Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Application Economic Model calculates either the required selling price of power and/or heat for a given internal rate of return (IRR) or the IRR for power and/or heat being sold at the market price. The user can generate these economic results for a range of reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for up to 16 reactor modules; and for module ratings of 200, 350, or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Application Economic Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Application Economic Model. This model was designed for users who are familiar with the HTGR design and Excel and engineering economics. Modification of the HTGR Application Economic Model should only be performed by users familiar with the HTGR and its applications, Excel, and Visual Basic.

  18. International nuclear fuel cycle evaluation

    International Nuclear Information System (INIS)

    Witt, P.

    1980-01-01

    In the end of February 1980, the two-years work on the International Nuclear Fuel Cycle Evaluation (INFCE) was finished in Vienna with a plenary meeting. INFCE is likely to have been a unique event in the history of international meetings: It was ni diplomatic negotiation meeting, but a techno-analytical investigation in which the participants tenaciously shuggled for many of the formulations. Starting point had been a meeting initiated by President Carter in Washington in Oct. 1979 after the World Economy Summit Meeting in London. The results of the investigation are presented here in a brief and popular form. (orig./UA) [de

  19. Nuclear fuel grid outer strap

    International Nuclear Information System (INIS)

    Duncan, R.; Craver, J.E.

    1989-01-01

    This patent describes a nuclear reactor fuel assembly grid. It comprises a first outer grip strap segment end. The first end having a first tab arranged in substantially the same plane as the plane defined by the first end; a second outer grip strap end. The second end having a second slot arranged in substantially the same plane as the plane defined by the second end, with the tab being substantially disposed in the slot, defining a socket therebetween; and a fort tine interposed substantially perpendicularly in the socket

  20. Radioecology of nuclear fuel cycles

    International Nuclear Information System (INIS)

    Cadwell, L.L.

    1982-01-01

    This study provides information to help assess the environmental impacts and certain potential human hazards associated with nuclear fuel cycles. A data base is being developed to define and quantify biological transport routes, which will permit credible predictions and assessment of routine and potential large-scale releases of radionuclides and other toxic materials. These data, used in assessment models, will increase the accuracy of estimating radiation doses to man and other life forms. Results will provide information to determine if waste management procedures on the Hanford site have caused ecological perturbations, and, if so, to determine the source, nature and magnitude of such disturbances

  1. Container for nuclear fuel powders

    International Nuclear Information System (INIS)

    Etheredge, B.F.; Larson, R.I.

    1982-01-01

    A critically safe container is disclosed for the storage and rapid discharge of enriched nuclear fuel material in powder form is disclosed. The container has a hollow, slab-shaped container body that has one critically safe dimension. A powder inlet is provided on one side wall of the body adjacent to a corner thereof and a powder discharge port is provided at another corner of the body approximately diagonal the powder inlet. Gas plenum for moving the powder during discharge are located along the side walls of the container adjacent the discharge port

  2. Radioecology of nuclear fuel cycles

    International Nuclear Information System (INIS)

    Schreckhise, R.G.; Cadwell, L.L.; Emery, R.M.

    1981-01-01

    This study provides information to help assess the environmental impacts and certain potential human hazards associated with nuclear fuel cycles. A data base is being developed to define and quantify biological transport routes which will permit credible predictions and assessment of routine and potential large-scale releases of radionuclides and other toxic materials. Information obtained from existing storage and disposal sites will provide a meaningful radioecological perspective with which to improve the effectiveness of waste management practices. This paper focuses on terrestrial and aquatic radioecology of waste management areas and biotic transport parameters

  3. Getter for nuclear fuel elements

    International Nuclear Information System (INIS)

    Ross, W.T.; Williamson, H.E.

    1976-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has disposed therein an improved getter capable of gettering reactive gases including a source of hydrogen. The getter comprises a composite with a substrate having thereon a coating capable of gettering reactive gases. The substrate has a greater coefficient of thermal expansion than does the coating, and over a period of time at reactor operating temperatures any protective film on the coating is fractured at various places and fresh portions of the coating are exposed to getter reactive gases. With further passage of time at reactor operating temperatures a fracture of the protective film on the coating will grow into a crack in the coating exposing further portions of the coating capable of gettering reactive gases. 13 claims, 5 drawing figures

  4. Getter for nuclear fuel elements

    International Nuclear Information System (INIS)

    Ross, W.T.; Williamson, H.E.

    1976-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has disposed therein an improved getter capable of gettering reactive gases including a source of hydrogen. The getter comprises a composite with a substrate having thereon a coating capable of gettering reactive gases. The substrate has a greater coefficient of thermal expansion than does the coating, and over a period of time at reactor operating temperatures any protective film on the coating is fractured at various places and fresh portions of the coating are exposed to getter reactive gases. With further passage of time at reactor operating temperatures a fracture of the protective film on the coating will grow into a crack in the coating exposing further portions of the coating capable of gettering reactive gases

  5. Nuclear fuel element leak detection system

    International Nuclear Information System (INIS)

    John, C.D. Jr.

    1978-01-01

    Disclosed is a leak detection system integral with a wall of a building used to fabricate nuclear fuel elements for detecting radiation leakage from the nuclear fuel elements as the fuel elements exit the building. The leak detecting system comprises a shielded compartment constructed to withstand environmental hazards extending into a similarly constructed building and having sealed doors on both ends along with leak detecting apparatus connected to the compartment. The leak detecting system provides a system for removing a nuclear fuel element from its fabrication building while testing for radiation leaks in the fuel element

  6. The Nuclear Fuel Cycle Information System

    International Nuclear Information System (INIS)

    1987-02-01

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

  7. Dissolving method for nuclear fuel oxide

    International Nuclear Information System (INIS)

    Tomiyasu, Hiroshi; Kataoka, Makoto; Asano, Yuichiro; Hasegawa, Shin-ichi; Takashima, Yoichi; Ikeda, Yasuhisa.

    1996-01-01

    In a method of dissolving oxides of nuclear fuels in an aqueous acid solution, the oxides of the nuclear fuels are dissolved in a state where an oxidizing agent other than the acid is present together in the aqueous acid solution. If chlorate ions (ClO 3 - ) are present together in the aqueous acid solution, the chlorate ions act as a strong oxidizing agent and dissolve nuclear fuels such as UO 2 by oxidation. In addition, a Ce compound which generates Ce(IV) by oxidation is added to the aqueous acid solution, and an ozone (O 3 ) gas is blown thereto to dissolve the oxides of nuclear fuels. Further, the oxides of nuclear fuels are oxidized in a state where ClO 2 is present together in the aqueous acid solution to dissolve the oxides of nuclear fuels. Since oxides of the nuclear fuels are dissolved in a state where the oxidizing agent is present together as described above, the oxides of nuclear fuels can be dissolved even at a room temperature, thereby enabling to use a material such as polytetrafluoroethylene and to dissolve the oxides of nuclear fuels at a reduced cost for dissolution. (T.M.)

  8. SPOUTED BED DESIGN CONSIDERATIONS FOR COATED NUCLEAR FUEL PARTICLES

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Douglas W.

    2017-07-01

    High Temperature Gas Cooled Reactors (HTGRs) are fueled with tristructural isotropic (TRISO) coated nuclear fuel particles embedded in a carbon-graphite fuel body. TRISO coatings consist of four layers of pyrolytic carbon and silicon carbide that are deposited on uranium ceramic fuel kernels (350µm – 500µm diameters) in a concatenated series of batch depositions. Each layer has dedicated functions such that the finished fuel particle has its own integral containment to minimize and control the release of fission products into the fuel body and reactor core. The TRISO coatings are the primary containment structure in the HTGR reactor and must have very high uniformity and integrity. To ensure high quality TRISO coatings, the four layers are deposited by chemical vapor deposition (CVD) using high purity precursors and are applied in a concatenated succession of batch operations before the finished product is unloaded from the coating furnace. These depositions take place at temperatures ranging from 1230°C to 1550°C and use three different gas compositions, while the fuel particle diameters double, their density drops from 11.1 g/cm3 to 3.0 g/cm3, and the bed volume increases more than 8-fold. All this is accomplished without the aid of sight ports or internal instrumentation that could cause chemical contamination within the layers or mechanical damage to thin layers in the early stages of each layer deposition. The converging section of the furnace retort was specifically designed to prevent bed stagnation that would lead to unacceptably high defect fractions and facilitate bed circulation to avoid large variability in coating layer dimensions and properties. The gas injection nozzle was designed to protect precursor gases from becoming overheated prior to injection, to induce bed spouting and preclude bed stagnation in the bottom of the retort. Furthermore, the retort and injection nozzle designs minimize buildup of pyrocarbon and silicon carbide on the

  9. MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION

    Directory of Open Access Journals (Sweden)

    Miroslav Cech

    2016-12-01

    Full Text Available This paper describes materials made of zirconium-based alloys used for nuclear fuel cladding fabrication. It is focused on corrosion problems their theoretical description and modeling in nuclear engineering.

  10. Environmental implications of thorium use in selected nuclear fuel cycles. Final

    International Nuclear Information System (INIS)

    Buckley, D.W.; Simmons, G.L.; Ziskind, R.A.

    1978-01-01

    The objective of this study was to assess the environmental implications of the nuclear fuel cycle associated with the highly enriched uranium concept of the High Temperature Gas Cooled Reactor. Model fuel cycles were constructed for the HTGR and a reference light water reactor (LWR) cycle. Mass flows were developed, control technology cases proposed and costed, effluents determined, and population doses calculated. Emphasis was given to the intercomparison of the fuel cycles to delineate areas which show pronounced departure. The dose commitment received by the population both within and outside a radius of 50 miles of each facility was determined. The 100 year population dose commitments due to a single year's plant operation was selected to facilitate intercomparison among fuel cycle components. No account was taken for long term waste sources associated with the fuel cycle such as mill tailing piles or terminal waste storage (study groundrule). The resource utilization and radionuclide activity of various fuel cycle options for using thorium in a Pressurized Water Reactor were studied. These data were contrasted with similar results obtained for a uranium fuel PWR

  11. OECD - HRP Summer School on Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

  12. International Summer School on Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

  13. OECD - HRP Summer School on Nuclear Fuel

    International Nuclear Information System (INIS)

    2000-01-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures

  14. The safety of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2005-01-01

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

  15. The safety of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2005-10-01

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

  16. Nuclear Fusion Fuel Cycle Research Perspectives

    International Nuclear Information System (INIS)

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

    2015-01-01

    As a part of the International Thermonuclear Experimental Reactor (ITER) Project, we at the Korea Atomic Energy Research Institute (KAERI) and our National Fusion Research Institute (NFRI) colleagues are investigating nuclear fusion fuel cycle hardware including a nuclear fusion fuel Storage and Delivery System (SDS). To have a better knowledge of the nuclear fusion fuel cycle, we present our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). To have better knowledge of the nuclear fusion fuel cycle, we presented our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). Our efforts to enhance the tritium confinement will be continued for the development of cleaner nuclear fusion power plants

  17. HTGR Generic Technology Program. Semiannual report for the period ending September 30, 1980

    International Nuclear Information System (INIS)

    1980-11-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the second half of FY-80. It covers a period when the design direction of the National HTGR Program is in the process of an overall review. The HTGR Generic Technology Program activities have continued so as to provide the basic technology required for all HTGR applications. The activities include the need to develop an LEU fuel and the need to qualify materials and components for the higher temperatures of the gas turbines and process heat plants

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

    International Nuclear Information System (INIS)

    Eckhoff, N.D.

    1981-07-01

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

  19. International issue: the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    In this special issue a serie of short articles of informations are presented on the following topics: the EEC's medium term policy regarding the reprocessing and storage of spent fuel, France's natural uranium supply, the Pechiney Group in the nuclear field, zircaloy cladding for nuclear fuel elements, USSI: a major French nuclear engineering firm, gaseous diffusion: the only commercial enrichment process, the transport of nuclear materials in the fuel cycle, Cogema and spent fuel reprocessing, SGN: a leader in the fuel cycle, quality control of mechanical, thermal and termodynamic design in nuclear engineering, Sulzer's new pump testing station in Mantes, the new look of the Ateliers et Chantiers de Bretagne, tubes and piping in nuclear power plants, piping in pressurized water reactor. All these articles are written in English and in French [fr

  20. Spent fuel reprocessing system availability definition by process simulation

    International Nuclear Information System (INIS)

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

    1978-05-01

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