WorldWideScience

Sample records for fuel irradiation experience

  1. Fuel irradiation experience at Halden

    International Nuclear Information System (INIS)

    Vitanza, Carlo

    1996-01-01

    The OECD Halden Reactor Project is an international organisation devoted to improved safety and reliability of nuclear power station through an user-oriented experimental programme. A significant part of this programme consists of studies addressing fuel performance issues in a range of conditions realised in specialised irradiation. The key element of the irradiation carried out in the Halden reactor is the ability to monitor fuel performance parameters by means of in-pile instrumentation. The paper reviews some of the irradiation rigs and the related instrumentation and provides examples of experimental results on selected fuel performance items. In particular, current irradiation conducted on high/very high burn-up fuels are reviewed in some detail

  2. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    International Nuclear Information System (INIS)

    Chichester, Heather Jean MacLean; Hayes, Steven Lowe; Dempsey, Douglas; Harp, Jason Michael

    2016-01-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  3. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  4. The achivements of Japanese fuel irradiation experiments in HBWR

    International Nuclear Information System (INIS)

    Ichikawa, Michio; Yanagisawa, Kazuaki; Domoto, Kazunari

    1984-02-01

    OECD Halden Reactor Project celebrated the 25th anniversary in 1983. The JAERI has been participating in the Project since 1967 on behalf of Japanese Government. Since the participation, thirty-six Japanese instrumented fuel assemblies have been irradiated in HBWR. The irradiation experiments were either sponsored by JAERI or by domestic organizations under the joint research agreements with JAERI, beeing steered by the Committee for the Joint Research Programme. The cooperative efforts have attained significant contributions to the development of water reactor fuel technology in Japan. This report review the irradiation experiments of Japanese fuel assemblies. (author)

  5. Shielding considerations for advanced fuel irradiation experiments

    International Nuclear Information System (INIS)

    Kang, Young-Hwan; Kim, Hee-Moon; Kim, Bong-Goo; Kim, Hark-Rho; Lee, Dong-Soo

    2008-01-01

    An in-pile test program for the development of a high burn-up fuel is planned for the HANARO reactor. The source term originates from a leakage of fission products from the anticipated failed fuels into the gas flow tubes and around the instrumentation and control system. In order to quantify the fuel composition in the event of a fuel failure, the isotope generation and depletion code ORIGEN 2.0 was used. The computer program Microshield 6.2 was used to calculate the doses from specific locations, where a high radioactivity is expected during an irradiation. The results indicate that the equivalent dose in the investigated working areas is less than the permitted dose rate of 6.25 μSv/hr. However, access to the area of a decay vessel may need to be limited, and the installation of a Pb wall with a 20.5 cm thickness is recommended. From the analysis of a radioactive decay with time, most of the concerned gaseous nuclides with short half-lives after 3 months, were decayed, with one exception which was Kr-85, thus it should be released in accordance with applicable government laws after measuring its activity in individual holding vessels. (author)

  6. Industrial experience of irradiated nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Delange, M.

    1981-01-01

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

  7. Light water reactor mixed-oxide fuel irradiation experiment

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

    1998-01-01

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

  8. Irradiation Experiments on Plutonium Fuels for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Frost, B. R.T.; Wait, E. [Atomic Energy Research Establishment Harwell, Berks. (United Kingdom)

    1967-09-15

    An assessment carried out some years ago indicated that cermet fuels might provide the high burn-up and integrity required for fast reactors. An irradiation programme was started at Harwell on (U, Pu)O{sub 2} -SS cermet plates and rods, mainly In thermal neutron fluxes, to gain experience of dimensional stability at temperatures typical of modern sodium-cooled fast reactor designs (600-650 Degree-Sign C). A subsequent assessment showed that cermets carried a large penalty as far as breeding was concerned and (U, Pu)C was chosen by Harwell for long-term study as an alternative, economic, fast reactor fuel. However, the results from the cermet experiments were of sufficient promise to proceed with parallel irradiation programmes on cermets and carbide. The studies of cermets showed that dimensional instability (swelling and cladding rupture) were caused by the pressures exerted on the steel matrix by the fuel particles, and that the initial density of the fuel particles was important in determining the burn-up at which failure occurred. Further, it was shown that cermets provided a useful vehicle for studying the changes occurring in oxide fuel particles with increasing burn-up. The disappearance of initial porosity and its replacement by fission gas bubbles and segregated solid fission products was studied in some detaiL No significant differences were observed between UO{sub 2} and(U,Pu)O{sub 2} particles. The initial studies of (U, Pu)C were concerned with the effect of varying composition and structure on swelling and fission gas release. A tantalum-lined nickel alloy cladding material was used to contain both pellet and powder specimens In an irradiation experiment in the core of the Dounreay fast reactor. This showed that the presence of a metal phase in the fuel led to a high swelling rate, that fission gas release was low up to {approx} 3% bum-up, and that a low density powder accommodated the swelling without excessive straining of the can. A subsequent

  9. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Baek, J. S [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.

    2014-04-30

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

  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. 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

  12. Achievements of Japanese fuel irradiation experiments in HBWR

    International Nuclear Information System (INIS)

    1992-10-01

    OECD NEA Halden Reactor Project started in 1958, and JAERI has been participated in the Project since 1967 on behalf of Japanese Government. During the participation period, not only JAERI but also many Japanese companies and PNC, which cooperated with JAERI, have carried out many irradiation tests of fuel at HBWR. The Committee of the Halden Joint Research Programme was organized by agencies and companies, which joined the cooperative researches, and the committee has worked to promote the cooperative researches. This report summarizes the achievements of the Halden Joint Research Programme on fuel irradiation tests between Jan. 1988 and Dec. 1990., as the Halden Project renews the agreement every three years. Some researches, which have not yet been completed in the period, are also included in this report. (author)

  13. 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

  14. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    International Nuclear Information System (INIS)

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-01-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at ∼2.4, ∼7 and ∼11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of ∼7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10 15 n/cm 2 /s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between ∼410 deg. C and ∼645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  15. Review of direct electrical heating experiments on irradiated mixed-oxide fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Bandyopadhyay, G.

    1982-01-01

    Results of approximately 50 out-of-reactor experiments that simulated various stages of a loss-of-flow event with irradiated fuel are presented. The tests, which utilized the direct electrical heating technique to simulate nuclear heating, were performed either on fuel segments with their original cladding intact or on fuel segments that were extruded into quartz tubes. The test results demonstrated that the macro- and microscopic fuel behavior was dependent on a number of variables including fuel heating rate, thermal history prior to a transient, the number of heating cycles, type of cladding (quartz vs stainless steel), and fuel burnup

  16. UK irradiation experience relevant to advanced carbide fuel concepts for LMFBR's

    International Nuclear Information System (INIS)

    Bagley, K.Q.; Batey, W.; Paris, R.; Sloss, W.M.; Snape, G.P.

    1977-01-01

    Despite discouraging prognoses of fabrication and reprocessing problems, it is recognized that the quest for a carbide fuel pin design which fully exploits the favourable density and thermal conductivity of (U,Pu) monocarbide must be maintained. Studies in aid of carbide fuel development have, therefore, continued in the UK in parallel with those on oxide, albeit at a substantially lower level of effort, and a sufficient body of irradiation experience has been accumulated to allow discrimination of realistic fuel pin designs

  17. Some UK experience and practice in the packaging and transport of irradiated fuel

    International Nuclear Information System (INIS)

    Edney, C.J.; Rutter, R.L.

    1977-01-01

    The origin and growth of irradiated fuel transport within and to the U.K. is described and the role of the organisations presently carrying out transport operations is explained. An explanation of the relevant U.K. regulations and laws affecting irradiated fuel transport and the role of the controlling body, the Department of the Environment is given. An explanation is given of the technical requirements for the transport of irradiated Magnox fuel and of the type of flask used, and the transport arrangements, both within the U.K. and to the U.K., from overseas is discussed. The technical requirements for the transport of C.A.G.R. fuel are outlined and the flask and transport arrangements are discussed. The transport requirements of oxide fuel from water reactors is outlined and the flask and shipping arrangements under which this fuel is brought to the U.K. from overseas is explained. The shipping arrangements are explained with particular reference to current international and national requirements. The requirements of the transport of M.T.R. fuel are discussed and the flask type explained. The expected future expansion of the transport of irradiated fuel within and to the U.K. is outlined and the proposed operating methods are briefly discussed. A summary is given of the U.K. experience and the lessons to be drawn from that experience

  18. The post irradiation examination of three fuel rods from the IFA 429 experiment irradiated in the Halden Reactor

    International Nuclear Information System (INIS)

    Williams, J.

    1979-11-01

    A series of fuel rod irradiation experiments were performed in the Halden Heavy Boiling Water Reactor in Norway. These were designed to provide a range of fuel property data as a function of burn-up. One of these experiments was the IFA-429. This was designed to study the absorption of helium filling gas by the UO 2 fuel pellets, steady state and transient fission gas release and fuel thermal behaviour to high burn-up. This data was to be obtained as a function of fuel density, fuel grain size, initial fuel/cladding gap, average linear heat rating, burn-up and overpower transients. All the fuel is in the form of pressed and sintered UO 2 pellets enriched to 13 weight percent 235 U. All the rods were clad in Zircaloy 4 tube. The details of the experiment are given. The post irradiation examination included: visual examination, neutron radiography, dimensional measurements, gamma scanning, measurement of gases in fuel rods and internal free volume, burn-up analysis, metallographic examination, measurement of retained gas in UO 2 pellets, measurement of bulk density of UO 2 . The results are given and discussed. (U.K.)

  19. Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within a specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.

  20. Irradiation tests of THTR fuel elements in the DRAGON reactor (irradiation experiment DR-K3)

    International Nuclear Information System (INIS)

    Burck, W.; Duwe, R.; Groos, E.; Mueller, H.

    1977-03-01

    Within the scope of the program 'Development of Spherical Fuel Elements for HTR', similar fuel elements (f.e.) have been irradiated in the DRAGON reactor. The f.e. were fabricated by NUKEM and were to be tested under HTR conditions to scrutinize their employability in the THTR. The fuel was in the form of coated particles moulded into A3 matrix. The kernels of the particles were made of mixed oxide of uranium and thorium with an U 235 enrichment of 90%. One aim of the post irradiation examination was the investigation of irradiation induced changes of mechanical properties (dimensional stability and elastic behaviour) and of the corrosion behaviour which were compared with the properties determined with unirradiated f.e. The measurement of the fission gas release in annealing tests and ceramografic examinations exhibited no damage of the coated particles. The measured concentration distribution of fission metals led to conclusions about their release. All results showed, that neither the coated particles nor the integral fuel spheres experienced any significant changes that could impair their utilization in the THTR. (orig./UA) [de

  1. Irradiation experience with KNK II Fast Breeder Fuel Subassemblies

    International Nuclear Information System (INIS)

    Hess, B.

    1993-02-01

    During the operation of the second core of KNK II fuel pin failures occurred, which were caused by local cladding weakening due to mechanical interaction between fuel pins and pin spacers. The present report gives a summarizing presentation of the consequences of these interactions, of the experimental and theoretical investigations to clarify the reason for the interactions and of measures to reduce their consequences in the extended residence time of the second core of KNK II. This type of interaction is caused by thermo-elastic instabilities of the fuel pin bundle, and its strength depends sensitively on the geometry of the pin bundle and the pin power. Finally, measures are described, which were taken for the fuel subassemblies of the third core of KNK II to avoid the wear causing instabilities [de

  2. Summary report on the fuel performance modeling of the AFC-2A, 2B irradiation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pavel G. Medvedev

    2013-09-01

    The primary objective of this work at the Idaho National Laboratory (INL) is to determine the fuel and cladding temperature history during irradiation of the AFC-2A, 2B transmutation metallic fuel alloy irradiation experiments containing transuranic and rare earth elements. Addition of the rare earth elements intends to simulate potential fission product carry-over from pyro-metallurgical reprocessing. Post irradiation examination of the AFC-2A, 2B rodlets revealed breaches in the rodlets and fuel melting which was attributed to the release of the fission gas into the helium gap between the rodlet cladding and the capsule which houses six individually encapsulated rodlets. This release is not anticipated during nominal operation of the AFC irradiation vehicle that features a double encapsulated design in which sodium bonded metallic fuel is separated from the ATR coolant by the cladding and the capsule walls. The modeling effort is focused on assessing effects of this unanticipated event on the fuel and cladding temperature with an objective to compare calculated results with the temperature limits of the fuel and the cladding.

  3. Irradiation experience with HT9-clad metallic fuel

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Tsai, H.; Billone, M.C.

    1991-01-01

    The safe and reliable performance of metallic fuel is currently under study and demonstration in the Integral Fast Reactor program. In-reactor tests of HT9-clad metallic fuel have now reached maturity and have all shown good performance characteristics to burnups exceeding 17.5 at. % in the lead assembly. Because this low-swelling tempered martensitic alloy is the cladding of choice for high fluence applications, the experimental observations and performance modelling efforts reported in this paper play an important role in demonstrating reliability

  4. BNFL's experience in the sea transport of irradiated research reactor fuel to the USA

    International Nuclear Information System (INIS)

    Hudson, I.A.; Porter, I.

    2000-01-01

    BNFL provides worldwide transport for a wide range of nuclear materials. BNFL Transport manages an unique fleet of vessels, designed, built, and operated to the highest safety standards, including the highest rating within the INF Code recommended by the International Maritime Organisation. The company has some 20 years of experience of transporting irradiated research reactor fuel in support of the United States' programme for returning US obligated fuel from around the world. Between 1977 and 1988 BNFL performed 11 shipments of irradiated research reactor fuel from the Japan Atomic Energy Research Institute to the US. Since 1997, a further 3 shipments have been performed as part of an ongoing programme for Japanese research reactor operators. Where possible, shipments of fuel from European countries such as Sweden and Spain have been combined with those from Japan for delivery to the US. (author)

  5. The investigation of fast reactor fuel pin start up behaviour in the irradiation experiment DUELL II

    International Nuclear Information System (INIS)

    Freund, D.; Geithoff, D.

    1988-04-01

    The irradiation experiments DUELL-II within the SNR-300 operational Transient Experimental Program deal with the investigation of fresh mixed oxide fuel behaviour at start-up. The irradiation has been carried out in the HFR Petten in four so-called DUELL capsules with two fuel pin samples each. The fuel pins with a total length of 453 mm contained a fuel column of 150 mm length, consisting of high dense (U,Pu)O 2-x fuel with an initial porosity of 4%, a Pu-content of 20.9%, and an O/Me ratio of 1.96. The fuel pellet diameter was 6.37 mm, the outer diameter of the SS cladding, material No. 1.4970, was 7.6 mm. The irradiation included four phases, consisting of preconditioning at 85% nominal power (corresponds to 550 W/cm), a following increase to full power, and two following full power periods of 1 and 10 days, respectively. Post irradiation examination showed incomplete fuel restructuring in the first capsules with central void diameters of 800 μm in the hot plane, complete restructuring in the last capsule, leading to central voids of approximately 1 mm diameter. The residual gaps between fuel and clad varied between 25 and 44 μm. The clad inner surface did not show any corrosion attack. The analysis of fuel restructuring has been carried out with the computer code SATURN-S showing good agreement with the PIE results. The analysis led to a series of model improvements, especially for crack volume and relocation modelling. (orig./GL) [de

  6. Experience of European LWR irradiated fuel transport: the first five hundred tonnes

    International Nuclear Information System (INIS)

    Curtis, H.W.

    1978-01-01

    The paper describes the service provided by an international company specializing in the transport of LWR irradiated fuel throughout Europe. Methods of transport used to the reprocessing plants at La Hague and Windscale include road transport of 38 te flasks over the whole route; transport of flasks between 55 and 105 te by rail, with rail-head and the reprocessing plant, where required, performed by road using heavy trailers; roll-on, roll-off sea ferries; and charter ships. Different modes of transport have been developed to cater for the various limitations on access to reactor sites arising from geographical and routing considerations. The experience of transporting more than 500 tonnes of irradiated uranium from twenty-one power reactors is used to illustrate the flexibility which the transport organization requires when the access and handling facilities are different at almost every reactor. Variations in fuel cross sections and lengths of fuel elements used in first generation reactors created the need for first generation flasks with sufficient variants to accommodate all reactor fuels but the trend now is towards standardization of flasks to perhaps two basic types. The safety record of irradiated fuel transport is examined with explanation of the means whereby this has been achieved. The problems of programming the movement of a pool of eighteen flasks for twenty-one reactors in eight countries are discussed together with the steps taken to ensure that the service operates fairly to give priority to those reactors with the greatest problems. The transport of irradiated fuel across several national frontiers is an international task which requires an international company. The transport of European irradiated fuel can be presented as an example of international collaboration which works

  7. Experience of European irradiated fuel transport - the first four hundred tonnes

    International Nuclear Information System (INIS)

    Curtis, H.W.

    1977-01-01

    The paper describes the successful integration of the experience of its three shareholders into an international company providing an irradiated fuel transport service throughout Europe. The experience of transporting more than 400 tonnes of irradiated uranium from fifteen power reactors is used to illustrate the flexibility which the transport organisation requires when the access and handling facilities are different at almost every reactor. Variations in fuel cross sections and lengths of fuel elements used in first generation reactors created the need for first generation flasks with sufficient variants to accommodate all reactor fuels but the trend now is towards standardisation of flasks to perhaps two basic types. Increases in fuel rating have raised the flask shielding and heat dissipation requirements and have influenced the design of later flasks. More stringent criticality acceptance criteria have tended to reduce the flask capacity below the maximum number of elements which could physically be contained. Reprocessing plant acceptance criteria initiated because of the presence of substantial quantities of loose crud released in the flask and the need to transport substantial numbers of failed elements have also reduced the flask capacity. Different modes of transport have been developed to cater for the various limitations on access to reactor sites arising from geographical and routing considerations. The safety record of irradiated fuel transport is examined with explanation of the means whereby this has been achieved. The problems of programming the movement of a pool of flasks for fifteen reactors in eight countries are discussed together with the steps taken to ensure that the service operates fairly to give priority to those reactors with the greatest problems. The transport of European irradiated fuel can be presented as an example of international collaboration which works

  8. Design considerations and operating experience with wet storage of Ontario Hydro's irradiated fuel

    International Nuclear Information System (INIS)

    Frost, C.R.; Naqvi, S.J.; McEachran, R.A.

    1987-01-01

    The characteristics of Ontario Hydro's fuel and at-reactor irradiated fuel storage water pools (or irradiated fuel bays) are described. There are two types of bay known respectively as primary bays and auxiliary bays, used for at-reactor irradiated fuel storage. Irradiated fuel is discharged remotely from Ontario Hydro's reactors to the primary bays for initial storage and cooling. The auxiliary bays are used to receive and store fuel after its initial cooling in the primary bay, and provide additional storage capacity as needed. The major considerations in irradiated fuel bay design, including site-specific requirements, reliability and quality assurance, are discussed. The monitoring of critical fuel bay components, such as bay liners, the development of high storage density fuel containers, and the use of several irradiated fuel bays at each reactor site have all contributed to the safe handling of the large quantities of irradiated fuel over a period of about 25 years. Routine operation of the irradiated fuel bays and some unusual bay operational events are described. For safety considerations, the irradiated fuel in storage must retain its integrity. Also, as fuel storage is an interim process, likely for 50 years or more, the irradiated fuel should be retrievable for downstream fuel management phases such as reprocessing or disposal. A long-term experimental program is being used to monitor the integrity of irradiated fuel in long-term wet storage. The well characterized fuel, some of which has been in wet storage since 1962 is periodically examined for possible deterioration. The evidence from this program indicates that there will be no significant change in irradiated fuel integrity (and retrievability) over a 50 year wet storage period

  9. Irradiation experiments on materials for core internals, pressure vessel and fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Materials degradation due to the aging phenomena is one of the key issues for the life assessment and extension of the light water reactors (LWRs). This presentation introduces JAERI`s activities in the field of LWR material researches which utilize the research and testing reactors for irradiation experiments. The activities are including the material studies for the core internals, pressure vessel and fuel cladding. These materials are exposed to the neutron/gamma radiation and high temperature water environments so that it is worth reviewing their degradation phenomena as the continuum. Three topics are presented; For the core internal materials, the irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels is the present major concern. At JAERI the effects of alloying elements on IASCC have been investigated through the post-irradiation stress corrosion cracking tests in high-temperature water. The radiation embrittlement of pressure vessel steels is still a significant issue for LWR safety, and at JAERI some factors affecting the embrittlement behavior such as a dose rate have been investigated. Waterside corrosion of Zircaloy fuel cladding is one of the limiting factors in fuel rod performance and an in-situ measurement of the corrosion rate in high-temperature water was performed in JMTR. To improve the reliability of experiments and to extent the applicability of experimental techniques, a mutual utilization of the technical achievements in those irradiation experiments is desired. (author)

  10. Comparison of Calculated and Measured Neutron Fluence in Fuel/Cladding Irradiation Experiments in HFIR

    International Nuclear Information System (INIS)

    Ellis, Ronald James

    2011-01-01

    A recently-designed thermal neutron irradiation facility has been used for a first series of irradiations of PWR fuel pellets in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. Since June 2010, irradiations of PWR fuel pellets made of UN or UO 2 , clad in SiC, have been ongoing in the outer small VXF sites in the beryllium reflector region of the HFIR, as seen in Fig. 1. HFIR is a versatile, 85 MW isotope production and test reactor with the capability and facilities for performing a wide variety of irradiation experiments. HFIR is a beryllium-reflected, light-water-cooled and -moderated, flux-trap type reactor that uses highly enriched (in 235 U) uranium (HEU) as the fuel. The reactor core consists of a series of concentric annular regions, each about 2 ft (0.61 m) high. A 5-in. (12.70-cm)-diam hole, referred to as the flux trap, forms the center of the core. The fuel region is composed of two concentric fuel elements made up of many involute-shaped fuel plates: an inner element that contains 171 fuel plates, and an outer element that contains 369 fuel plates. The fuel plates are curved in the shape of an involute, which provides constant coolant channel width between plates. The fuel (U 3 O 8 -Al cermet) is nonuniformly distributed along the arc of the involute to minimize the radial peak-to-average power density ratio. A burnable poison (B 4 C) is included in the inner fuel element primarily to reduce the negative reactivity requirements of the reactor control plates. A typical HEU core loading in HFIR is 9.4 kg of 235 U and 2.8 g of 10 B. The thermal neutron flux in the flux trap region can exceed 2.5 x 10 15 n/cm 2 · s while the fast flux in this region exceeds 1 x 10 15 n/cm 2 · s. The inner and outer fuel elements are in turn surrounded by a concentric ring of beryllium reflector approximately 1 ft (0.30 m) thick. The beryllium reflector consists of three regions: the removable reflector, the semi-permanent reflector, and the

  11. Actinide nitride ceramic transmutation fuels for the Futurix-FTA irradiation experiment

    International Nuclear Information System (INIS)

    Voit, St.; McClellan, K.; Stanek, Ch.; Maloy, St.

    2007-01-01

    Full text of publication follows. The transmutation of plutonium and other minor actinides is an important component of an advanced nuclear fuel cycle. The Advanced Fuel Cycle Initiative (AFCI) is currently considering mono-nitrides as potential transmutation fuel material on account of the mutual solubility of actinide mono-nitrides as well as their desirable thermal characteristics. The feedstock is most commonly produced by a carbothermic reduction/nitridisation process, as it is for this programme. Fuel pellet fabrication is accomplished via a cold press/sinter approach. In order to allow for easier investigation of the synthesis and fabrication processes, surrogate material studies are used to compliment the actinide activities. Fuel compositions of particular interest denoted as low fertile (i.e. containing uranium) and non-fertile (i.e. not containing uranium) are (PuAmNp) 0.5 U 0.5 N and (PuAm) 0.42 Zr 0.58 N, respectively. The AFCI programme is investigating the validity of these fuel forms via Advanced Test Reactor (ATR) and Phenix irradiations. Here, we report on the recent progress of actinide-nitride transmutation fuel development and production for the Futurix-FTA irradiation experiment. Furthermore, we highlight specific cases where the complimentary approach of surrogate studies and actinide development aid in the understanding complex material issues. In order to allow for easier investigation of the fundamental materials properties, surrogate materials have been used. The amount of surrogate in each compound was determined by comparing both molar concentration and lattice parameter mismatch via Vegard Law. Cerium was chosen to simultaneously substitute for Pu, Am and Np, while depleted U was chosen to substitute for enriched U. Another goal of this work was the optimisation of added graphite during carbothermic reduction in order to minimise the duration of the carbon removal step (i.e. heat treatment under H 2 containing gas). One proposed

  12. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  13. Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

    After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

  14. Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

    After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

  15. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  16. Full-sized plates irradiation with high UMo fuel loading. Final results of IRIS 1 experiment

    International Nuclear Information System (INIS)

    Huet, F.; Marelle, V.; Noirot, J.; Sacristan, P.; Lemoine, P.

    2003-01-01

    As a part of the French UMo Group qualification program, IRIS 1 experiment contained full-sized plates with high uranium loading in the meat of 8 g.cm -3 . The fuel particles consisted of 7 and 9 wt% Mo-uranium alloys ground powders. The plate were irradiated at OSIRIS reactor in IRIS device up to 67.5% peak burnup within the range of 136 W.cm - '2 for the heat flux and 72 deg. C for the cladding temperature. After each reactor cycle the plates thickness were measured. The results show no swelling behaviour differences versus burnup between UMo7 and UMo9 plates. The maximum plate swelling for peak burnup location remains lower than 6%. The wide set of PIE has shown that, within the studied irradiation conditions, the interaction product have a global formulation of '(U-Mo)Al -7 ' and that there is no aluminium dissolution in UMo particles. IRIS1 experiment, as the first step of the UMo fuel qualification for research reactor, has established the good behaviour of UMo7 and UMo9 high uranium loading full-sized plate within the tested conditions. (author)

  17. Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

    International Nuclear Information System (INIS)

    Elbel, H.; Kummerer, K.; Bojarsky, K.; Lopez Jimenez, J.; Otero de la Gandara, J.L.

    1979-01-01

    Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

  18. 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

  19. Argentine nuclear fuels MOX irradiated in the Petten reactor: Analysis of experience with the BACO code

    Energy Technology Data Exchange (ETDEWEB)

    Marino, A C; Perez, E; Adelfang, P [Argentine Atomic Energy Commission, Buenos Aires (Argentina)

    1997-08-01

    The irradiation of our first prototypes of MOX nuclear fuels fabricated in Argentina began in 1986. These experiences had been made in the HFR-Petten reactor, Holland. The six rods were fabricated in the {alpha} Facility (GAID-CNEA-Argentina). The first rod has been used for destructive pre-irradiation analysis in the KFK (Kernforschungszentrum Karlsruhe), Germany. The second one was a pathfinder for calibrating systems in the HFR. Another two rods included doped pellets based on iodine. One of them included CsI and auxiliary components. The second one included elemental iodine. The concentration of iodine was intended to simulate 15 MWd/ton(M) of burnup. We defined the power histories with the BACO code. We assumed a cycle of 15 days that included interaction treatments of cladding and pellet due to the power cycling. The last ramp is let run until stress corrosion cracking (SCC) is induced. The experience named BU15 was done with the last two rods. The final burnup was 15 MWd/ton(M), and a final ramp test was arranged for one of them. This burnup is the same as the previous two rods. The power level during irradiation was low and without major solicitations, only the normal shutdowns of the HFR. The ramp was similar to that used for the iodine test. We attempt to see the correct correspondence between the BU15 and the doping test. The pathfinder had an excellent behavior in the HFR reactor. The presence of microcracks inside the cladding was observed in the iodine test as we predicted with the BACO code. The post-irradiation tests of the BU15 experience has just ended. The development of the ramp was interrupted due to an increase of activity in the system. We presumed the presence of a failure in the rod. The visual inspection of the rod shows an atypical failure for this kind of fuel, i.e. they found a small circular hole. We use the BACO code for the behavior analysis of the fuel rods. 23 refs, 29 figs, 5 tabs.

  20. Argentine nuclear fuels MOX irradiated in the Petten reactor: Analysis of experience with the BACO code

    International Nuclear Information System (INIS)

    Marino, A.C.; Perez, E.; Adelfang, P.

    1997-01-01

    The irradiation of our first prototypes of MOX nuclear fuels fabricated in Argentina began in 1986. These experiences had been made in the HFR-Petten reactor, Holland. The six rods were fabricated in the α Facility (GAID-CNEA-Argentina). The first rod has been used for destructive pre-irradiation analysis in the KFK (Kernforschungszentrum Karlsruhe), Germany. The second one was a pathfinder for calibrating systems in the HFR. Another two rods included doped pellets based on iodine. One of them included CsI and auxiliary components. The second one included elemental iodine. The concentration of iodine was intended to simulate 15 MWd/ton(M) of burnup. We defined the power histories with the BACO code. We assumed a cycle of 15 days that included interaction treatments of cladding and pellet due to the power cycling. The last ramp is let run until stress corrosion cracking (SCC) is induced. The experience named BU15 was done with the last two rods. The final burnup was 15 MWd/ton(M), and a final ramp test was arranged for one of them. This burnup is the same as the previous two rods. The power level during irradiation was low and without major solicitations, only the normal shutdowns of the HFR. The ramp was similar to that used for the iodine test. We attempt to see the correct correspondence between the BU15 and the doping test. The pathfinder had an excellent behavior in the HFR reactor. The presence of microcracks inside the cladding was observed in the iodine test as we predicted with the BACO code. The post-irradiation tests of the BU15 experience has just ended. The development of the ramp was interrupted due to an increase of activity in the system. We presumed the presence of a failure in the rod. The visual inspection of the rod shows an atypical failure for this kind of fuel, i.e. they found a small circular hole. We use the BACO code for the behavior analysis of the fuel rods. 23 refs, 29 figs, 5 tabs

  1. Irradiation experiments of recycled PuO2-UO2 fuels by SAXTON reactor, (1)

    International Nuclear Information System (INIS)

    Yumoto, Ryozo; Akutsu, Hideo

    1975-01-01

    Seventy two mixed oxide fuel rods made by PNC were irradiated in Saxton Core 3. This paper generally describes the fuel specifications, the power history of the fuel and the post-irradiation examination of the PNC fuel. The specifications of the 4.0 w/o and 5.0 w/o enriched PuO 2 fuel rods with zircaloy-4 cladding are presented in a table and a figure. The positions of PNC fuel rods in the Saxton reactor are shown in a figure. Sixty eight 5.0 w/o PuO 2 -UO 2 fuel rods were assembled in a 9 x 9 rod array together with zircaloy-4 bars, a flux thimble, and a Sb-Be source. The power history of the Saxton Core 3 and the irradiation history of the PNC fuel rods are summarized in tables. The peak power and burnup of each fuel rod and the axial power profile are also presented. The maximum linear power rate and burnup attained were 512W/cm and 8700 MWD/T, respectively. As for the post irradiation examination, the items of nondestructive test, destructive test, and cladding test are presented together with the working flow diagram of the examination. It is concluded that the performance of all fuel rods was safe and satisfactory throughout the power history. (Aoki, K.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-08-01

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

  3. Rules for the licensing of new experiments in BR2: application to the test irradiation of new MTR-fuels

    International Nuclear Information System (INIS)

    Joppen, F.

    2000-01-01

    New types of MTR fuel elements are being developed and require a qualification before routine operation could be authorized. During the test irradiation the new fuel elements .are considered as experimental devices and their irradiation is allowed according to the procedures for experiments. Authorization is based on the advice .of a consultative committee on experiments. This procedure is valid as long as the irradiation is covered by the actual reactor license. An additional license or an amendment is only required if due to the experiment the risk for the workers or the environment is increased in a significant way. A few experimental fuel plates loaded in the primary loop of the reactor will not increase this risk. The source term for potential radioactive releases remains more or less the same. The probability for an accident can be limited by restricting the heat flux and surface temperature. (author)

  4. Post-irradiation examination of a 13000C-HTR fuel experiment Project J 96.M3

    International Nuclear Information System (INIS)

    Bueger, J. de; Roettger, H.

    1977-01-01

    A large variety of loose coated fuel particles have been irradiated in the BR2 at Mol/Belgium at temperatures between 1200 0 C and 1400 0 C and up to a fast neutron fluence of 1.2x1022 cm -2 (E>0.1 MeV) as a Euratom sponsored experiment for the advanced testing of HTR fuel. The specimens have been provided by Belgonucleaire and the Dragon Project. A short description of the experiment as well as the results of post-irradiation examination mainly carried out at Petten (N.H.), The Netherlands, are presented here. The post-irradiation examination has shown that the required performance can be achieved by a number of the tested fuel specimens without serious damage

  5. Drilling Experiments of Dummy Fuel Rods Using a Mock-up Drilling Device and Detail Design of Device for Drilling of Irradiated Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Yong; Lee, H. K.; Chun, Y. B.; Park, S. J.; Kim, B. G

    2007-07-15

    KAERI are developing the safety evaluation method and the analysis technology for high burn-up nuclear fuel rod that is the project, re-irradiation for re-instrumented fuel rod. That project includes insertion of a thermocouple in the center hole of PWR nuclear fuel rod with standard burn-up, 3,500{approx}4,000MWD/tU and then inspection of the nuclear fuel rod's heat performance during re-irradiation. To re-fabricate fuel rod, two devices are needed such as a drilling machine and a welding machine. The drilling machine performs grinding a center hole, 2.5 mm in diameter and 50 mm in depth, for inserting a thermocouple. And the welding machine is used to fasten a end plug on a fuel rod. Because these two equipment handle irradiated fuel rods, they are operated in hot cell blocked radioactive rays. Before inserting any device into hot cell, many tests with that machine have to be conducted. This report shows preliminary experiments for drilling a center hole on dummy of fuel rods and optimized drilling parameters to lessen operation time and damage of diamond dills. And the design method of a drilling machine for irradiated nuclear fuel rods and detail design drawings are attached.

  6. Six years working experience of the Marcoule plant for treatment of irradiated fuel

    International Nuclear Information System (INIS)

    Jouannaud, C.

    1964-01-01

    The irradiated fuel treatment plant at Marcoule began treating rods from the pile G 1 in July 1958. These six years experience of the plant in operation have led to the confirmation or revision of the original ideas concerning the process as well as the technology or methods of exploitation. The process as a whole has suffered little modification, the performances having proved better than originally foreseen; the only alterations made were justified by greater simplicity of operation, better nuclear security (criticality) or for technological reasons. The processes of plutonium reduction from valency IV to valency III by uranium IV, and of concentration of fission product solutions in the presence of formaldehyde, have always given complete satisfaction. The initial concept of direct maintenance of the installations has been justified by experience. Certain maintenance jobs, originally considered impossible after the start of operations, have proved feasible and have been carried out under acceptable conditions; a number of examples are given. From experience it has been possible to define optimal conditions for the design of these installations such as to provide a maximum in robustness and ease of maintenance. The advantages of continuously-operating equipment have been shown. Certain installations have been altered in accordance with these new ideas. Analytical checking in the laboratory has been profoundly modified, and the plans adopted are such that complete safety in work on radioactive solutions is compatible with a very good working speed. Experience has also shown the advantages of having a group on the spot to carry out short-term applied studies. Finally, a strict working discipline and excellent collaboration with the radiation protection service have enabled us to reach the end of these six years, during some of which the exploitation was intensive, without irradiation accident. (authors) [fr

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

    International Nuclear Information System (INIS)

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

    1981-06-01

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

  8. 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

  9. European experience in the transport of irradiated light-water reactor fuel

    International Nuclear Information System (INIS)

    Curtis, H.W.

    1979-01-01

    Various methods of transport of irradiated fuel flasks in Europe are described. While many problems in the transport of heavy flasks have been solved some remain and new ones have appeared. Some of these problems are the accumulation of crud on the surface of fuel elements, the problems of failed fuel, stringent criticality criteria, the ''sweating out'' of contaminated flasks, the access, road or rail, to reactor sites, and the maintenance of the transport vehicles. Some future trends in the direction of heavy flasks in the range of 75 to 100 tonnes are indicated

  10. The behaviour of irradiated fuel under RIA transients: Interpretation of the CABRI experiments

    International Nuclear Information System (INIS)

    Papin, J.; Rigat, H.; Breton, J.P.; Schmitz, F.

    1996-01-01

    Paper presents the results of investigation of highly irradiated PWR fuel behaviour under fast power transients conducted in a sodium loop of CABRI reactor, as well as the results on development and validation of computer code SCANAIR. (author). 8 refs, 9 figs, 2 tabs

  11. Experience with processing irradiated fuel at the Savannah River Plant (1954--1976)

    International Nuclear Information System (INIS)

    Sheldon, E.B.

    1977-09-01

    The processing facilities for recovery of uranium and plutonium from irradiated fuel elements have operated since 1954 without major unplanned interruptions. The operation has comprised capaigns ranging from a few weeks to two years, with no prolonged outages except for a period of about two years when one of the two processing facilities was remodeled to increase its capacity. Over the 23-year period 1954-1976, approximately 30,000 metric tons of irradiated uranium were processed. Since 1958, in addition to recovery of uranium and weapons-grade 239 Pu, the plant has produced 238 Pu, which is used principally as a heat source. Through June 1976, a total of 320 kg of 238 Pu has been shipped offsite. There have been no lost-time injuries due to radiation and no criticality accidents in these or other Savannah River Plant (SRP) facilities. Radiation exposures to individual workers in fuel reprocessing at SRP have averaged 0.3 to 0.7 rem per year. Releases of radioactivity to the atmosphere and to plant streams and environmental levels of radionuclides have been monitored since startup. Fuel irradiated in SRP reactors is stored in a water-filled basin at each reactor for a period of time to permit decay of short-lived radioactivity before shipment to the reprocessing areas. Currently that storage period is a minimum of 200 days. In addition to its fuel processing activities, SRP stores a number of special ERDA-irradiated fuels which require shear-leach dissolution or other major processes not available at SRP. These fuels, containing a total of 2500 kg of 235 U, are stored underwater in the RBOF facility. A number have been in storage since 1968. Storage in RBOF has been without significant incident

  12. Fuel or irradiation subassembly

    International Nuclear Information System (INIS)

    Seim, O.S.; Hutter, E.

    1975-01-01

    A subassembly for use in a nuclear reactor is described which incorporates a loose bundle of fuel or irradiation pins enclosed within an inner tube which in turn is enclosed within an outer coolant tube and includes a locking comb consisting of a head extending through one side of the inner sleeve and a plurality of teeth which extend through the other side of the inner sleeve while engaging annular undercut portions in the bottom portion of the fuel or irradiation pins to prevent movement of the pins

  13. Irradiation experiment conceptual design parameters for MURR LEU U-Mo fuel conversion

    International Nuclear Information System (INIS)

    Stillman, J.; Feldman, E.; Stevens, J.; Wilson, E.

    2013-03-01

    This report contains the results of reactor design and performance calculations for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the nominal steady-state irradiation conditions of a key set of plates containing peak irradiation parameters found in MURR cores fueled with the LEU monolithic U-Mo alloy fuel with 10 wt% Mo.

  14. Final report on development and operation of instrumented irradiation capsules for creep experiments on nuclear fuels at FR2

    International Nuclear Information System (INIS)

    Haefner, H.E.; Philipp, K.; Blumhofer, M.

    1980-02-01

    The capsule test rig No. 154 removed from FR2 in April 1979 was the last irradiation rig in a long series of creep experiments. The target of the irradiation tests, started exactly ten years ago, was to investigate the creep behaviour of various ceramic nuclear fuels under different in-pile irradiation conditions. An irradiation test rig had been developed for this purpose which allowed the continuous measurement of changes in length of fuel specimens. A total of 28 capsule test rigs each containing two packages of creep specimens have been irradiated in FR2 during this decade. They included 23 specimen stacks of UO 2 , 16 specimen stacks of UO 2 -PuO 2 , 4 specimen stacks of UN, 10 specimen stacks of (U,Pu) C, and 13 reference specimens of molybdenum. Besides the description of the test facility, the report provides above all a survey of the operation data applicable to the specimens and of the operating experience gathered as well as of the findings obtained in post-irradiation examinations. (orig.) [de

  15. Irradiation experiments of 3rd, 4th and 5th fuel assemblies by an in-pile gas loop, OGL-1

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Hayashi, Kimio; Minato, Kazuo; Kikuchi, Teruo; Adachi, Mamoru; Iwamoto, Kazumi; Ikawa, Katsuichi; Itami, Hiroharu.

    1986-07-01

    Three irradiation experiments for 3rd, 4th and 5th fuel assemblies which had been composed of VHTR reference coated particle fuels and graphite components were carried out by an in-pile gas loop, OGL-1 during 1979 and 1982. The main purposes of these experiments were to study on bowing of the fuel rod by irradiation for the 3rd fuel assembly, to study on fuel behavior under relatively low burnup irradiation for the 4th fuel assembly, and to study on fuel behavior up to full burnup of VHTR design for the 5th fuel assembly. For understanding in-pile fuel behavior, fractional releases of fission gases from each fuel assembly were estimated by measuring the fission gas concentrations in the primary loop of OGL-1. The post-irradiation examination (PIE) was carried out extensively on the fuel block, the fuel rods and the fuel compacts in Tokai Hot Laboratory. Also, made were the measurements of metallic fission product distributions in the fuel assemblies and the fuel rods. The results in these experiments were given as follows ; bowing of the fuel rod in the 3rd fuel assembly was 0.7 mm, but integrity of the rod was kept under irradiation. Fractional release of the fission gas from the 4th fuel assembly remained in the order of 10 -7 during irradiation, suggesting that the fuel performance was excellent. The fractional release from the 5th fuel assembly, on the other hand, was in the order of 10 -5 which was the same level in the VHTR design. (author)

  16. 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.)

  17. Irradiated fuel bundle counter

    International Nuclear Information System (INIS)

    Campbell, J.W.; Todd, J.L.

    1975-01-01

    The design of a prototype safeguards instrument for determining the number of irradiated fuel assemblies leaving an on-power refueled reactor is described. Design details include radiation detection techniques, data processing and display, unattended operation capabilities and data security methods. Development and operating history of the bundle counter is reported. (U.S.)

  18. Irradiated fuel bundle counter

    International Nuclear Information System (INIS)

    Campbell, J.W.; Todd, J.L.

    1975-01-01

    The design of a prototype safeguards instrument for determining the number of irradiated fuel assemblies leaving an on-power refueled reactor is described. Design details include radiation detection techniques, data processing and display, unattended operation capabilities and data security methods. Development and operating history of the bundle counter is reported

  19. The second Euratom sponsored 9000C HTR fuel irradiation experiment in the HFR Petten Project E 96.02: Pt.2. Post-irradiation examination

    International Nuclear Information System (INIS)

    Roettger, R.; Bueger, J. de; Schoots, T.

    1977-01-01

    A large variety of HTR fuel specimens, loose coated particles, coupons and compacts provided by Belgonucleaire, the Dragon Project and the KFA Juelich have been irradiated in the HFR at Petten at about 900 0 C up to a maximum fast neutron fluence of about 7x10 21 cm -2 (EDN) as a Euratom sponsored experiment. The maximum burn-ups were between 11 and 18.5% FIMA. The results of the post-irradiation examinations, comprising visual inspection, dimensional measurements, microradiography, metallography, and burn-up determinations are presented in this part 2 of the final report. The examinations have shown that the endurance limit of most of the tested fuel varieties is beyond the reached irradiation values

  20. Post-irradiation examination of fifteen UO2/PuO2-fuel pins from the experiment DFR-350

    International Nuclear Information System (INIS)

    Geithoff, D.

    1975-06-01

    Within the framework of the fuel pin development for a sodium-cooled fast reactor a subassembly containing 77 fuel pins has been irradiated up to 5.65% fima in the Dounreay fast reactor. The pins were prototypes in terms of fuel and cladding material. The fuel consisted of mechanically mixed UO 2 (80%) and PuO 2 (20%) pressed into pellets whereas austenitic steels (W.-No. 1,4961 and 1,4988) were used as cladding material. Furthermore a blanket column of UO 2 pellets and a gas plenum were incorporated in the pin. For irradiation the conditions in a fast breeder were simulated by a linear rod power of 450 W/cm and a maximum cladding temperature of 630 0 C. After the successful completion of the irradiation, the subassembly was dismantled and fifteen pins were selected for a nondestructive and destructive examination. The tests included visual control, measurement of external dimensions, γ-spectroscopy, X-ray radiography, fission gas measurement, ceramography, radiochemical burn-up measurement. The results are presented. The most important results of the examinations seem to be the migration of fission product cesium and the fact that no signs of impending pin failure have been found. Thus the pin specification tested in this experiment is capable of achieving higher burnups under the irradiation conditions described above. (orig./AK) [de

  1. Experience with an ultrasonic sealing system for nuclear safeguards in irradiated fuel bay demonstrations

    International Nuclear Information System (INIS)

    White, B.F.; Smith, M.T.

    1985-07-01

    The development of the irradiated fuel safeguards containment assembly for CANDU nuclear generating stations has stimulated the development of the AECL Random Coil Sealing System. The ARC seal combines the identity and integrity elements in an ultrasonically-determined signature. This is verified in situ, in real time with the seal reading system. The maturation of this technology has been facilitated with demonstration trials in the NRU and NPD irradiated fuel bays. The NPD demonstration includes operation of the systems tooling by Ontario Hydro staff. It provides the opportunity for IAEA inspectors from Toronto and Vienna to direct the operational procedures and to perform the data acquisition. The procedures and systems developed in these trials are reviewed. The estimation of the system performance characteristics from the observations is presented. A minimum frequency of reading for individual seals is recommended to be once per annum following initial deployment

  2. An experiment to examine the mechanistic behaviour of irradiated CANDU fuel stored under dry conditions

    International Nuclear Information System (INIS)

    Oldaker, I.E.; Crosthwaite, J.L.; Keltie, R.J.; Truss, K.J.

    1979-01-01

    A program has begun to use the Whiteshell Nuclear Research Establishment dry-storage canisters to store some selected CANDU irradiated fuel bundles in an 'easily retrievable basket.' The object of the experimental program is to study the long-term stability of the Zircaloy-sheathed UO 2 and UC fuel elements when stored in air. Bundles were loaded into a canister in October 1979 following detailed examination and removal of up to three complete elements from most bundles. These elements are currently being subjected to detailed destructive examinations, including metallography and scanning electron micrography, to fully characterize their pre-storage condition. After four years, and every five years thereafter, further elements will be examined similarly to study the effects of the storage environment on the stability of the Zircaloy sheathing, and on its continued ability to contain the fuel safely in an interim storage facility. (author)

  3. Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

    International Nuclear Information System (INIS)

    Harp, Jason M.; Demkowicz, Paul A.

    2014-01-01

    In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10"-"4 to 10"-"5) of as manufactured defects and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from intentionally failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application was considered. Previous experience utilizing similar techniques, the expected activities in AGR-3/4 rings, and analysis of this work indicate using GECT to evaluate AGR-3/4 will be feasible. The GECT technique was also applied to other irradiated nuclear fuel systems currently available in the HFEF hot cell, including oxide fuel pins, metallic fuel pins, and monolithic plate fuel. Results indicate GECT with the HFEF PGS is effective. (author)

  4. Analysis of Core Physics Experiments on Irradiated BWR MOX Fuel in REBUS Program

    International Nuclear Information System (INIS)

    Yamamoto, Toru; Ando, Yoshihira; Hayashi, Yamato

    2008-01-01

    As part of analyses of experimental data of a critical core containing a irradiated BWR MOX test bundle in the REBUS program, depletion calculations was performed for the BWR MOX fuel assemblies from that the MOX test rods were selected by using a general purpose neutronics code system SRAC. The core analyses were carried out using SRAC and a continuous energy Monte Carlo code MVP. The calculated k eff s were compared with those of the core containing a fresh MOX fuel bundle in the program. The SRAC-diffusion calculation underestimates k eff s of the both cores by 1.0 to 1.3 %dk and the k eff s of MVP are 1.001. The difference in k eff between the irradiated BWR MOX test bundle core and the fresh MOX one is 0.4 %dk in the SRAC-diffusion calculation and 0.0 %dk in the MVP calculation. The calculated fission rate distributions are in good agreement with the measurement in the SRAC-diffusion and MVP calculations. The calculated neutron flux distributions are also in good agreement with the measurement. The calculated burnup reactivity in the both calculations well reproduce the measurements. (authors)

  5. Irradiation behaviors of coated fuel particles, (3)

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kashimura, Satoru; Iwamoto, Kazumi; Ikawa, Katsuichi

    1980-07-01

    This report is concerning to the irradiation experiments of the coated fuel particles, which were performed by 72F-6A and 72F-7A capsules in JMTR. The coated particles referred to the preliminary design of VHTR were prepared for the experiments in 1972 and 1973. 72F-6A capsule was irradiated at G-10 hole of JMTR fuel zone for 2 reactor cycles, and 72F-7A capsule had been planned to be irradiated at the same irradiation hole before 72F-6A. However, due to slight leak of the gaseous fission products into the vacuum system controlling irradiation temperature, irradiation of 72F-7A capsule was ceased after 85 hrs since the beginning. In the post irradiation examination, inspection to surface appearance, ceramography, X-ray microradiography and acid leaching for the irradiated particle samples were made, and crushing strength of the two particle samples was measured. (author)

  6. Nondestructive analysis of irradiated fuels

    International Nuclear Information System (INIS)

    Dudey, N.D.; Frick, D.C.

    1977-01-01

    The principal nondestructive examination techniques presently used to assess the physical integrity of reactor fuels and cladding materials include gamma-scanning, profilometry, eddy current, visual inspection, rod-to-rod spacing, and neutron radiography. LWR fuels are generally examined during annual refueling outages, and are conducted underwater in the spent fuel pool. FBR fuels are primarily examined in hot cells after fuel discharge. Although the NDE techniques are identical, LWR fuel examinations emphasize tests to demonstrate adherence to technical specification and reliable fuel performance; whereas, FBR fuel examinations emphasize aspects more related to the relative performance of different types of fuel and cladding materials subjected to variable irradiation conditions

  7. 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

  8. Safeguards approach for irradiated fuel

    International Nuclear Information System (INIS)

    Harms, N.L.; Roberts, F.P.

    1987-03-01

    IAEA verification of irradiated fuel has become more complicated because of the introduction of variations in what was once presumed to be a straightforward flow of fuel from reactors to reprocessing plants, with subsequent dissolution. These variations include fuel element disassembly and reassembly, rod consolidation, double-tiering of fuel assemblies in reactor pools, long term wet and dry storage, and use of fuel element containers. This paper reviews future patterns for the transfer and storage of irradiated LWR fuel and discusses appropriate safeguards approaches for at-reactor storage, reprocessing plant headend, independent wet storage, and independent dry storage facilities

  9. Container for irradiated fuel

    International Nuclear Information System (INIS)

    Guy, R.

    1978-01-01

    The transport container for irradiated or used nuclear fuel is provided with an identical heat shield against fires on the top and bottom sides. Each heat shield consists of two inner nickel plates, whose contact surfaces are polished to a mirror finish and an outer plate of stainless steel. The nickel plate on the box is spot welded to it while the second nickel plate is spot welded to the steel plate. Both together are in turn welded so as to be leaktight to the edges of the box. For extreme heat effects and based on the different (bimetal) coefficients of expansion, the steel plate with the nickel plate attached to it bulges away from the box. The second nickel plate remains at the box, so that a subpressure space is formed with the mirror nickel surfaces. The heat radiation and heat conduction to the box are greatly reduced by this. (DG) [de

  10. 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

  11. Fission product phases in irradiated carbide fuels

    International Nuclear Information System (INIS)

    Ewart, F.T.; Sharpe, B.M.; Taylor, R.G.

    1975-09-01

    Oxide fuels have been widely adopted as 'first charge' fuels for demonstration fast reactors. However, because of the improved breeding characteristics, carbides are being investigated in a number of laboratories as possible advanced fuels. Irradiation experiments on uranium and mixed uranium-plutonium carbides have been widely reported but the instances where segregate phases have been found and subjected to electron probe analysis are relatively few. Several observations of such segregate phases have now been made over a period of time and these are collected together in this document. Some seven fuel pins have been examined. Two of the irradiations were in thermal materials testing reactors (MTR); the remainder were experimental assemblies of carbide gas bonded oxycarbide and sodium bonded oxycarbide in the Dounreay Fast Reactor (DFR). All fuel pins completed their irradiation without failure. (author)

  12. An intercomparison experiment on isotope dilution thermal ionisation mass spectrometry using plutonium-239 spike for the determination of plutonium concentration in dissolver solution of irradiated fuel

    International Nuclear Information System (INIS)

    Aggarwal, S.K.; Shah, P.M.; Saxena, M.K.; Jain, H.C.; Gurba, P.B.; Babbar, R.K.; Udagatti, S.V.; Moorthy, A.D.; Singh, R.K.; Bajpai, D.D.

    1996-01-01

    Determination of plutonium concentration in the dissolver solution of irradiated fuel is one of the key measurements in the nuclear fuel cycle. This report presents the results of an intercomparison experiment performed between Fuel Chemistry Division (FCD) at BARC and PREFRE, Tarapur for determining plutonium concentration in dissolver solution of irradiated fuel using 239 Pu spike in isotope dilution thermal ionisation mass spectrometry (ID-TIMS). The 239 Pu spike method was previously established at FCD as viable alternative to the imported enriched 242 Pu or 244 Pu; the spike used internationally for plutonium concentration determination by IDMS in dissolver solution of irradiated fuel. Precision and accuracy achievable for determining plutonium concentration are compared under the laboratory and the plant conditions using 239 Pu spike in IDMS. For this purpose, two different dissolver solutions with 240 Pu/ 239 Pu atom ratios of about 0.3 and 0.07 corresponding, respectively, to high and low burn-up fuels, were used. The results of the intercomparison experiment demonstrate that there is no difference in the precision values obtained under the laboratory and the plant conditions; with mean precision values of better than 0.2%. Further, the plutonium concentration values determined by the two laboratories agreed within 0.3%. This exercise, therefore, demonstrates that ID-TIMS method using 239 Pu spike can be used for determining plutonium concentration in dissolver solution of irradiated fuel, under the plant conditions. 7 refs., 8 tabs

  13. Transportation of irradiated fuel elements

    International Nuclear Information System (INIS)

    1980-01-01

    A critique is presented of current methods of transporting spent nuclear fuel and the inadequacies of the associated contingency plans, with particular reference to the transportation of irradiated fuel through London. Anti-nuclear and pro-nuclear arguments are presented on a number of factors, including tests on flasks, levels of radiation exposure, routine transport arrangements and contingency arrangements. (U.K.)

  14. Fabrication of Metallic Fuel Slugs for Irradiation Experiments in Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arun; Prasad, G.J.

    2013-01-01

    Advantages of Metallic fuels for future FBR: → High heavy metal atom density; → Higher thermal conductivity at room temperature that increases with temperature; → Metal fuels can be relatively easily fabricated with close dimensional tolerances; → They have excellent compatibility with liquid metal coolants

  15. Transport experience of NH-25 spent fuel shipping cask for post irradiation examination

    International Nuclear Information System (INIS)

    Mori, Ryuji

    1982-01-01

    Since the Japan Atomic Energy Research Institute and Nippon Nuclear Fuel Development Co. hot laboratories are located far off from the port which can handle spent fuel shipping casks, it is necessary to use a trailer-mounted cask which can be transported by public roads, bridges and intersections for the transportation of spent fuel specimens to these hot laboratories. Model NH-25 shipping cask was designed, manufactured and oualification tested to meet Japanese regulations and was officially registered as a BM type cask. The NH-25 cask accomodates two BWR fuel assemblies, one PWR assembly or one ATR fuel assembly using interchangeable inner containers. The cask weight is 29.2 t. The cask has three concentric stainless steel shells. Gamma shielding is lead cast between the inner shell and the intermediate shell. Neutro n shielding consists of ethylene-glycol-aqueous solution layer formed between the intermediate shell and the outer shell. The NH-25 cask now has been in operation for 2.5 yr. It was used for the transportation of spent fuel assemblies from six LWR power plants to the port on shipping cask carrier ''Hinouramaru'' on the sea, as well as from the port to the hot laboratory on a trailer. The capability of safe handling and transporting of spent fuel assemblies has been well demonstrated. (author)

  16. The transportation of PuO2 and MOX fuel and management of irradiated MOX fuel

    International Nuclear Information System (INIS)

    Dyck, H.P.; Rawl, R.; Durpel, L. van den

    2000-01-01

    Information is given on the transportation of PuO 2 and mixed-oxide (MOX) fuel, the regulatory requirements for transportation, the packages used and the security provisions for transports. The experience with and management of irradiated MOX fuel and the reprocessing of MOX fuel are described. Information on the amount of MOX fuel irradiated is provided. (author)

  17. Experience from transportation of irradiated WWER-440 fuel assemblies at Kozloduy NPP site after a short cooling time

    International Nuclear Information System (INIS)

    Stoyanova, I.; Kamenov, A.; Byrzev, L.; Christoskov, I.

    2003-01-01

    The presented results from the computation and analysis of the radiation characteristics of the irradiated fuel assemblies by the date of their transportation according to the selected loading patterns of the VSPOT cask and following the modified technology of transportation, i.e. without replacement of the pool solution by pure condensate, as well as the corresponding experimental results, confirm the applicability of the newly introduced safety criterion for the selection of a loading pattern of the cask with irradiated fuel assemblies after a short cooling time. The comparison between measured and computed surface dose rates shows that during the procedure of transfer of irradiated fuel assemblies from the pools of Units 1 and 2 to the pools of Kozloduy NPP Units 3 and 4 all safety limits, incl. the radiation protection requirements, were met

  18. The BG18, a B(U)F type package used for the transport of irradiated fuel rods - return of experience

    Energy Technology Data Exchange (ETDEWEB)

    Juergen, S.; Herman, S. [Transnubel, Dessel (Belgium)

    2004-07-01

    The purpose of this presentation is to share the return of experience of Transnubel after a period of nearly 3 years operation of the BG18 package in several nuclear power plants and hot cell facilities. This package has been used mainly for the shipment of full scale as well as samples of irradiated fuel rods - UOX or MOX, PWR or BWR.

  19. The BG18, a B(U)F type package used for the transport of irradiated fuel rods - return of experience

    International Nuclear Information System (INIS)

    Juergen, S.; Herman, S.

    2004-01-01

    The purpose of this presentation is to share the return of experience of Transnubel after a period of nearly 3 years operation of the BG18 package in several nuclear power plants and hot cell facilities. This package has been used mainly for the shipment of full scale as well as samples of irradiated fuel rods - UOX or MOX, PWR or BWR

  20. Direct electrical heating of irradiated metal fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

    1985-01-01

    The Integral Fast Reactor (IFR) concept proposed by Argonne National Laboratory utilizes a metal fuel core. Reactor safety analysis requires information on the potential for fuel axial expansion during severe thermal transients. In addition to a comparatively large thermal expansion coefficient, metallic fuel has a unique potential for enhanced pre-failure expansion driven by retained fission gas and ingested bond sodium. In this paper, the authors present preliminary results from three direct electrical heating (DEH) experiments performed on irradiated metal fuel to investigate axial expansion behavior. The test samples were from Experimental Breeder Reactor II (EBR-II) driver fuel ML-11 irradiated to 8 at.% burnup. Preliminary analysis of the results suggest that enhanced expansion driven by trapped fission gas can occur

  1. Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

    2015-09-01

    Low-enrichment (235U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

  2. Irradiation effects on fuels for space reactors

    International Nuclear Information System (INIS)

    Ranken, W.A.; Cronenberg, A.W.

    1984-01-01

    A review of irradiation-induced swelling and gas release experience is presented here for the three principal fuels UO 2 , UC, and UN. The primary advantage of UC and UN over UO 2 is higher thermal conductivity and attendant lower fuel temperature at equivalent pellet diameter and power density, while UO 2 offers the distinct benefit of well-known irradiation performance. Irradiation test results indicate that at equivalent burnup, temperature, and porosity conditions, UC experiences higher swelling than UO 2 or UN. Fission gas swelling becomes important at fuel temperatures above 1320 K for UC, and at somewhat higher temperatures for UO 2 and UN. Evidence exists that at equivalent fuel temperatures and burnups, high density UO 2 and UN experience comparable swelling behavior; however, differences in thermal conductivity influence overall irradiation performance. The low conductivity of UO 2 results in higher thermal gradients which contribute to fuel microcracking and gas release. As a result UO 2 exhibits higher fractional gas release than UN, at least or burnups up to about 3%

  3. Uranium and thorium loadings determined by chemical and nondestructive methods in HTGR fuel rods for the Fort St. Vrain Early Validation Irradiation Experiment

    International Nuclear Information System (INIS)

    Angelini, P.; Rushton, J.E.

    1979-01-01

    The Fort St. Vrain Early Validation Irradiation Experiment is an irradiation test of reference and of improved High-Temperature Gas-Cooled Reactor fuels in the Fort St. Vrain Reactor. The irradiation test includes fuel rods fabricated at ORNL on an engineering scale fuel rod molding machine. Fuel rods were nondestructively assayed for 235 U content by a technique based on the detection of prompt-fission neutrons induced by thermal-neutron interrogation and were later chemically assayed by using the modified Davies Gray potentiometric titration method. The chemical analysis of the thorium content was determined by a volumetric titration method. The chemical assay method for uranium was evaluated and the results from the as-molded fuel rods agree with those from: (1) large samples of Triso-coated fissile particles, (2) physical mixtures of the three particle types, and (3) standard solutions to within 0.05%. Standard fuel rods were fabricated in order to evaluate and calibrate the nondestructive assay device. The agreement of the results from calibration methods was within 0.6%. The precision of the nondestructive assay device was established as approximately 0.6% by repeated measurements of standard rods. The precision was comparable to that estimated by Poisson statistics. A relative difference of 0.77 to 1.5% was found between the nondestructive and chemical determinations on the reactor grade fuel rods

  4. CEA fuel pencil qualification under irradiation: from component conception to fuel assembly irradiation in a power reactor

    International Nuclear Information System (INIS)

    Marin, J.-F.; Pillet, Claude; Francois, Bernard; Morize, Pierre; Petitgrand, Sylvie; Atabek, R.-M.; Houdaille, Brigitte.

    1981-06-01

    Fabrication of fuel pins made of uranium oxide pellets and of a zircaloy 4 cladding is described. Irradiation experiment results are given. Thermomechanical behavior of the fuel pin in a power reactor is examined [fr

  5. Performance of a sphere-pac mixed carbide fuel pin irradiated in the Dounreay Fast Reactor (DFR 527/1 experiment)

    International Nuclear Information System (INIS)

    Bischoff, K.; Smith, L.; Stratton, R.W.

    1980-10-01

    The DFR 527/1 experiment was the first irradiation of EIR sphere-pac uranium-plutonium mixed carbide fuel in a fast flux. The experiment has been successfully irradiated to a burn-up of 7.3% FIMA at ratings between 45 and 62 kW m - 1 and clad temperatures between 300 and 600 0 C. Restructuring and elemental redistribution has been found to be similar to the pattern established for pellet type fuel and follows effects seen in earlier sphere-pac carbide tests. Gas release of 12-14% has been measured. A preliminary comparison of radial temperature distribution calculations using a first version of the fuel behaviour modelling code SPECKLE with the actual metallography has been attempted. (Auth.)

  6. Irradiation performance of metallic fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Porter, D.L.; Batte, G.L.; Hofman, G.L.

    1989-01-01

    Argonne National Laboratory has been working for the past five years to develop and demonstrate the Integral Fast Reactor (IFR) concept. The concept involves a closed system for fast-reactor power generation and on-site fuel reprocessing, both designed specifically around the use of metallic fuel. The Experimental Breeder Reactor-II (EBR-II) has used metallic fuel for all of its 25-year life. In 1985, tests were begun to examine the irradiation performance of advanced-design metallic fuel systems based on U-Zr or U-Pu-Zr fuels. These tests have demonstrated the viable performance of these fuel systems to high burnup. The initial testing program will be described in this paper. 2 figs

  7. Irradiation behaviors of coated fuel particles, (4)

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kashimura, Satoru; Ogawa, Toru; Ikawa, Katsuichi; Iwamoto, Kazumi; Ishimoto, Kiyoshi

    1981-09-01

    Loose coated fuel particles prepared in confirmity to a preliminary design for the multi-purpose VHTR in fiscal 1972 - 1974 were irradiated by 73F - 12A capsule in JMTR. Main purpose for this irradiation experiment was to examine irradiation stability of the candidate TRISO coated fuel particles for the VHTR. Also the coated particles possessing low-density kernel (90%TD), highly anisotropic OLTI-PyC and ZrC coating layer were loaded with the candidate particles in this capsule. The coated particles were irradiated up to 1.5 x 10 21 n/cm 2 of fast neutron fluence (E > 0.18 MeV) and 3.2% FIMA of burnup. In the post irradiation examination it was observed that among three kinds of TRISO particles exposed to irradiation corresponding to the normal operating condition of the VHTR ones possessing poor characteristics of the coating layers did not show a good stability. The particles irradiated under abnormally high temperature condition (> 1800 0 C) revealed 6.7% of max. EOL failure fraction (95% confidence limit). Most of these particles were failed by the ameoba effect. Furthermore, among four kinds of the TRISO particles exposed to irradiation corresponding to the transient condition of the VHTR (--1500 0 C) the two showed a good stability, while the particles possessing highly anisotropic OLTI-PyC or poorly characteristic coating layers were not so good. (author)

  8. The 3rd irradiation test plan of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Park, J. H. and others

    2001-05-01

    The objective of the 3rd irradiation test of DUPIC fuel at the HANARO is to estimate the in-core behaviour of a DUPIC pellet that is irradiated up to more than average burnup of CANDU fuel. The irradiation of DUPIC fuel is planned to start at May 21, 2001, and will be continued at least for 8 months. The burnup of DUPIC fuel through this irradiation test is thought to be more than 7,000 MWd/tHE. The DUPIC irradiation rig instrumented with three SPN detectors will be used to accumulate the experience for the instrumented irradiation and to estimate the burnup of irradiated DUPIC fuel more accurately. Under normal operating condition, the maximum linear power of DUPIC fuel was estimated as 55.06 kW/m, and the centerline temperature of a pellet was calculated as 2510 deg C. In order to assess the integrity of DUPIC fuel under the accident condition postulated at the HANARO, safety analyses on the locked rotor and reactivity insertion accidents were carried out. The maximum centerline temperature of DUPIC fuel was estimated 2590 deg C and 2094 deg C for each accident, respectively. From the results of the safety analysis, the integrity of DUPIC fuel during the HANARO irradiation test will be secured. The irradiated DUPIC fuel will be transported to the IMEF. The post-irradiation examinations are planned to be performed at the PIEF and IMEF.

  9. Post irradiation test report of irradiated DUPIC simulated fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

    2001-12-01

    The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are γ-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

  10. Experience from the transportation of irradiated WWER-440 fuel assemblies at the Kozloduy NPP site after a short cooling time

    International Nuclear Information System (INIS)

    Stoyanova, I.; Kamenov, A.; Byrzev, L.; Christoskov, I.

    2003-01-01

    Results from the analysis and experimental verification of the radiation and shielding characteristics of non-standard loading patterns of the VSPOT transport cask used for transportation of irradiated fuel assemblies after a short cooling time (120 180 days) on the site of the Kozloduy NPP are presented. An additional safety criterion related to the introduced modifications to the standard procedure of using the transport cask is formulated and discussed (Authors)

  11. CANDU reactor experience: fuel performance

    International Nuclear Information System (INIS)

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

    1985-07-01

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

  12. Management of irradiated CANDU fuel

    International Nuclear Information System (INIS)

    Lupien, Mario

    1985-01-01

    The nuclear industry, like any other industrial activity, generates waste and, since these radioactive products are known to be hazardous both to man and his natural environment, they are subject to stringent controls. The irradiated fuel is also highly radioactive and remains so for thousands of years. It is estimated that by the year 2000, nuclear reactors in Canada alone will have produced some 50 Gg of radioactive fuel which is stored at the nuclear plant site itself. The nuclear industry plays a leading role in the research and development effort to find suitable waste-management methods. Its R and D programs cover many scientific fields, including chemistry, and therefore demand a considerable amount of coordination. The knowledge acquired in this multidisciplinary context should form a basis for solving many of today's industrial-waste problems. This paper describes the various stages in the long management process. In the medium term, the irradiated fuel will be stored in surface installations but the long-term solution proposed is to emplace the used fuel or the fuel recycle waste deep underground in a stable geologic formation

  13. Transportation of irradiated fuel elements

    International Nuclear Information System (INIS)

    Preece, A.H.

    1980-01-01

    The report falls under the headings: introduction (explaining the special interest of the London Borough of Brent, as forming part of the route for transportation of irradiated fuel elements); nuclear power (with special reference to transport of spent fuel and radioactive wastes); the flask aspect (design, safety regulations, criticisms, tests, etc.); the accident aspect (working manual for rail staff, train formation, responsibility, postulated accident situations); the emergency arrangements aspect; the monitoring aspect (health and safety reports); legislation; contingency plans; radiation - relevant background information. (U.K.)

  14. Out-of-pile burnout experiments in a full-scale simulation of an irradiation rig in a HIFAR hollow fuel element facility

    International Nuclear Information System (INIS)

    Chapman, A.G.; Hargreaves, N.D.

    1986-06-01

    Flow measurement and burnout experiments were performed in an out-of-pile test rig which simulates the conditions of UO 2 irradiation rig in a hollow fuel element facility of the HIFAR reactor. One per cent of the coolant flow in the fuel element passed through the irradiation rig. A burnout heat flux of 90 W cm -2 was observed at the surface of an electrically-heated, dummy irradiation can. When the coolant flow rate in the irradiation rig was increased by a factor of 2.5, the burnout heat flux rose by 30 per cent to 117 W cm -2 . A simple modification to the supporting frame for the cans improved the burnout heat flux by 3 per cent at 1 per cent of the coolant flow, but enhanced it by 17 per cent at 2.5 per cent of the coolant flow. Of ten burnout events observed, eight were located upstream of the end of the heated length of the can. The burnout results form a self-consistent, credible set of data and provide a rational basis for the establishment of maximum permissible operating heat fluxes in irradiation rigs of the type simulated. Recommendations are made for the practical application of the results

  15. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  16. Gamma scanning of the irradiated HANARO fuels

    International Nuclear Information System (INIS)

    Hong, Kwon Pyo; Lee, K. S.; Park, D. G.; Baik, S. Y.; Song, W. S.; Kim, T. Y.; Seo, C. K.

    1997-02-01

    To conform the burnup state of the fuels, we have transported the irradiated HANARO fuels from the reactor to IMEF (Irradiated Material Examination Facility), and executed gamma scanning for the fuels. By measuring the gamma-rays from the irradiated fuels we could see the features of the relative burnup distributions in the fuel bundles. All of 17 fuel bundles were taken in and out between HANARO and IMEF from March till August in 1996, and we carried out the related regulations. Longitudinal gamma scanning and angular gamma scanning are done for each fuel bundle without dismantlement of the bundles. (author). 5 tabs., 25 figs

  17. The MARINE experiment: Irradiation of sphere-pac fuel and pellets of UO{sub 2−x} for americium breeding blanket concept

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E., E-mail: elio.dagata@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Hania, P.R. [Nuclear Research and Consultancy Group, P.O. Box 25, NL-1755 ZG Petten (Netherlands); Freis, D.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Bejaoui, S. [Commissariat à l’Energie Atomique et aux Energies Alternatives, DEN/DEC, F-13108 St. Paul lez Durance Cedex (France); Charpin, F.F.; Baas, P.J.; Okel, R.A.F.; Til, S. van [Nuclear Research and Consultancy Group, P.O. Box 25, NL-1755 ZG Petten (Netherlands); Lapetite, J.-M. [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Delage, F. [Commissariat à l’Energie Atomique et aux Energies Alternatives, DEN/DEC, F-13108 St. Paul lez Durance Cedex (France)

    2017-01-15

    Highlights: • MARINE is designed to check the behaviour of MABB sphere-pac concept. • MABB sphere-pac are compared with MABB pellet. • Swelling and helium release behaviour will be the main output of the experiment. • An experiment to check sphere-pac MADF fuel behaviour has been already performed. - Abstract: Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like {sup 241}Am is therefore an option for the reduction of radiotoxicity and heat production of waste packages to be stored in a repository. The MARINE irradiation experiment is the latest of a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS, SPHERE) performed in the High Flux Reactor (HFR). The MARINE experiment is developed and carried out in the framework of the collaborative research project PELGRIMM of the EURATOM 7th Framework Programme (FP7). During the past years of experimental works in the field of transmutation and tests of innovative nuclear fuels, the release or trapping of helium as well as swelling have been shown to be the key issues for the design of such kind of fuel both as drivers and even more for Am-bearing blanket targets (due to the higher Am contents). The main objective of the MARINE experiment is to study the in-pile behaviour of uranium oxide fuel containing 13% of americium and to compare the behaviour of sphere-pac versus pellet fuel, in particular the role of microstructure and temperature on fission gas release and He on fuel swelling. The MARINE experiment will be irradiated in 2016 in the HFR in Petten (The Netherlands) and is expected to be completed in spring 2017. This paper discusses the rationale and objective of the MARINE experiment and provides a general description of its design for which some innovative features have been adopted.

  18. 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.

  19. Six years working experience of the Marcoule plant for treatment of irradiated fuel; Experience de 6 annees de fonctionnement de l'usine de retraitement de Marcoule

    Energy Technology Data Exchange (ETDEWEB)

    Jouannaud, C [CEA Marcoule, Centre de Production de Plutonium, 30 (France)

    1964-07-01

    The irradiated fuel treatment plant at Marcoule began treating rods from the pile G 1 in July 1958. These six years experience of the plant in operation have led to the confirmation or revision of the original ideas concerning the process as well as the technology or methods of exploitation. The process as a whole has suffered little modification, the performances having proved better than originally foreseen; the only alterations made were justified by greater simplicity of operation, better nuclear security (criticality) or for technological reasons. The processes of plutonium reduction from valency IV to valency III by uranium IV, and of concentration of fission product solutions in the presence of formaldehyde, have always given complete satisfaction. The initial concept of direct maintenance of the installations has been justified by experience. Certain maintenance jobs, originally considered impossible after the start of operations, have proved feasible and have been carried out under acceptable conditions; a number of examples are given. From experience it has been possible to define optimal conditions for the design of these installations such as to provide a maximum in robustness and ease of maintenance. The advantages of continuously-operating equipment have been shown. Certain installations have been altered in accordance with these new ideas. Analytical checking in the laboratory has been profoundly modified, and the plans adopted are such that complete safety in work on radioactive solutions is compatible with a very good working speed. Experience has also shown the advantages of having a group on the spot to carry out short-term applied studies. Finally, a strict working discipline and excellent collaboration with the radiation protection service have enabled us to reach the end of these six years, during some of which the exploitation was intensive, without irradiation accident. (authors) [French] L'usine de traitement des combustibles irradies de

  20. Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

    Science.gov (United States)

    Chang, G. S.; Lillo, M. A.

    2009-08-01

    The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y

  1. HANARO fuel irradiation test (II): revision

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, D. S.; Kim, H.; Chae, H. T.; Lee, C. S.; Kim, B. G.; Lee, C. B

    2001-04-01

    In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiated at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%. This report is the revision of KAERI/TR-1816/2001 on the irradiation test for HANARO fuel.

  2. Gamma spectrometrical examination of irradiated fuel

    International Nuclear Information System (INIS)

    Kristof, Edvard; Pregl, Gvido

    1988-01-01

    Gamma scanning is the only non-destructive technique for quantitative measuring of fission or activation products in spent fuel. The negligence of local variation of the linear attenuation coefficient of gamma rays in the irradiated fuel remains the main source of systematic error. To eliminate it we combine the (single) emission gamma ray scanning technique with a transmission measurement. Mathematical procedure joined with the experiment is particularly convenient for fuel elements of circular cross-section. In such a manner good results are obtainable even for relatively small number of measuring data. Accomplished routines enable to esteem the finite width of the collimation slit. The experiment has been partially automated. Trial measurements were carried out, and the measured data were successfully processed

  3. ENEL's experience in the management of irradiated fuel. Old and new problems encountered by nuclear station operators

    International Nuclear Information System (INIS)

    Ariemma, A.; Cuzzaniti, M.; Zaffiro, B.; Bertini, A.

    1983-01-01

    The experience acquired in recent years in the management of spent fuel discharged from ENEL's reactors has revealed a substantial change in the attitude of nuclear station operators in connection with the end of the fuel cycle downstream from the reactor (back-end). While in the past, after President Carter's outline of US policy in April 1977, the utilities had to face great difficulty in reprocessing their fuel owing to inadequate capacity, today the same problem is regarded as a matter of cost-benefit analysis from an industrial standpoint and of appropriate planning for the utilization of the recovered fissile materials. Since the present technology allows spent fuel storage (dry or underwater) to be planned for rather long periods and plutonium utilization requires a very stringent schedule, the present trend is to ensure medium-term storage of spent fuel and to seek a greater flexibility in the final reprocessing stages so as to render plutonium availability consistent with the programmes for its utilization. As a consequence, the solution to the problems posed by high-activity waste disposal is being delayed, thus allowing an exhaustive and detailed analysis of all the possible solutions to be made. The paper describes a number of solutions to the problems ENEL has encountered in the fuel cycle back-end. (author)

  4. Detailed measurements of local thickness changes for U-7Mo dispersion fuel plates with Al-3.5Si matrix after irradiation at different powers in the RERTR-9B experiment

    Science.gov (United States)

    Keiser, Dennis D.; Williams, Walter; Robinson, Adam; Wachs, Dan; Moore, Glenn; Crawford, Doug

    2017-10-01

    The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. Swelling is an important irradiation behavior that needs to be well understood. Data from high resolution thickness measurements performed on U-7Mo dispersion fuel plates with Al-Si alloy matrices that were irradiated at high power is sparse. This paper reports the results of detailed thickness measurements performed on two dispersion fuel plates that were irradiated at relatively high power to high fission densities in the Advanced Test Reactor in the same RERTR-9B experiment. Both plates were irradiated to similar fission densities, but one was irradiated at a higher power than the other. The goal of this work is to identify any differences in the swelling behavior when fuel plates are irradiated at different powers to the same fission densities. Based on the results of detailed thickness measurments, more swelling occurs when a U-7Mo dispersion fuel with Al-3.5Si matrix is irradiated to a high fission density at high power compared to one irradiated at a lower power to high fission density.

  5. Gamma irradiation plants using reactor fuel elements

    International Nuclear Information System (INIS)

    Suckow, W.

    1976-11-01

    Recent irradiation plants utilizing fuel elements are described. Criteria for optimizing such plants, evaluation of the plants realized so far, and applications for the facilities are discussed. (author)

  6. Small-scale irradiated fuel electrorefining

    International Nuclear Information System (INIS)

    Benedict, R.W.; Krsul, J.R.; Mariani, R.D.; Park, K.; Teske, G.M.

    1993-01-01

    In support of the metallic fuel cycle development for the Integral Fast Reactor (IFR), a small scale electrorefiner was built and operated in the Hot Fuel Examination Facility (HFEF) at Argonne National Laboratory-West. The initial purpose of this apparatus was to test the single segment dissolution of irradiated metallic fuel via either direct dissolution in cadmium or anodic dissolution. These tests showed that 99.95% of the uranium and 99.99% of the plutonium was dissolved and separated from the fuel cladding material. The fate of various fission products was also measured. After the dissolution experiments, the apparatus was upgraded to stady fission product behavior during uranium electrotransport. Preliminary decontamination factors were estimated for different fission products under different processing conditions. Later modifications have added the following capabilities: Dissolution of multiple fuel segments simultaneously, electrotransport to a solid cathode or liquid cathode and actinide recovery with a chemical reduction crucible. These capabilities have been tested with unirradiated uranium-zirconium fuel and will support the Fuel Cycle Demonstration program

  7. UK experience on fuel and cladding interaction in oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Batey, W [Dounreay Experimental Reactor Establishment, Thurso, Caithness (United Kingdom); Findlay, J R [AERE, Harwell, Didcot, Oxon (United Kingdom)

    1977-04-01

    The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed.

  8. UK experience on fuel and cladding interaction in oxide fuels

    International Nuclear Information System (INIS)

    Batey, W.; Findlay, J.R.

    1977-01-01

    The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed

  9. KEY RESULTS FROM IRRADIATION AND POST-IRRADIATION EXAMINATION OF AGR-1 UCO TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; Morris, Robert N.

    2016-11-01

    The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior within the US program. The fuel exhibited a very low incidence of TRISO coating failure during irradiation and post-irradiation safety testing at temperatures up to 1800°C. Advanced PIE methods have allowed particles with SiC coating failure to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of

  10. HANARO fuel irradiation test(II)

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, D. S.; Kim, H. R.; Chae, H. T.; Lee, B. C.; Lee, C. S.; Kim, B. G.; Lee, C. B.; Hwang, W

    2001-04-01

    In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiatied at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%.

  11. Irradiation performance of HTGR fertile fuel in HFIR target capsules HT-12 through HT-15. Part I. Experiment description and fission product behavior

    International Nuclear Information System (INIS)

    Kania, M.J.; Lindemer, T.B.; Morgan, M.T.; Robbins, J.M.

    1977-02-01

    Sixteen types of Biso-coated designs, on ThO 2 kernels, were irradiated in High Flux Isotope Reactor target capsules HT-12 through HT-15. The report addresses the description of the experiment and extensive postirradiation analyses and experiments to determine fertile-particle burnup, fuel coating failures, and fission product behavior. Several low-temperature isotropic (LTI) pyrocarbon coatings, which ''survived'' according to visual inspection, were shown to have developed permeability during irradiation. These particles were irradiated at temperatures approximately equal to 1250 0 C and to burnups equal to or greater than 8 percent fission per initial heavy-metal atom (FIMA). No evidence of permeability was found in similar particles irradiated at temperatures approximately equal to 1550 0 C and burnups approximately equal to 16 percent FIMA. Failures due to permeability were not detectable by visual inspection but required a more extensive investigation by the 1000 0 C gaseous chlorine leaching technique. Maximum particle surface operating temperatures were found to be approximately 300 0 C in excess of design limits of 900 0 C (low-temperature magazines) and 1250 0 C (high-temperature magazines). The extremes of high temperatures and fast neutron fluences up to 1.6 x 10 22 neutrons/cm 2 produced severe degradation and swelling of the Poco graphite magazines and sample holders

  12. 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

  13. EDF energy generation UK transport of irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    James, R. [EDF Energy, London, (United Kingdom)

    2015-07-01

    This paper give an overview of irradiated fuel transport in the UK. It describes the design of irradiated fuel flask used by EDF Energy; operational experience and good practices learnt from over 50 years of irradiated fuel transport. The AGRs can store approximately 9 months generation of spent fuel, hence the ability to transport irradiated fuel is vital. Movements are by road to the nearest railhead, typically less than 2 miles and then by rail to Sellafield, up to 400 miles, for reprocessing or long term storage. Road and rail vehicles are covered. To date in the UK: over 30,000 Magnox flask journeys and over 15,000 AGR A2 flask journeys have been carried out.

  14. RECH-1 test fuel irradiation status report

    International Nuclear Information System (INIS)

    Marin, J.; Lisboa, J.; Olivares, L.; Chavez, J.

    2005-01-01

    Since May 2003, one RECH-1 fuel element has been submitted to irradiation at the HFR-Petten, Holland. By November 2004 the irradiation has achieved its pursued goal of 55% burn up. This irradiation qualification service will finish in the year 2005 with PIE tests, as established in a contractual agreement between the IAEA, NRG, and CCHEN. This report presents the objectives and the current results of this fuel qualification under irradiation. Besides, a brief description of CHI/4/021, IAEA's Technical Cooperation Project that has supported this irradiation test, is also presented here. (author)

  15. Results of the post-irradiation examination of a highly-rated mixed oxide fuel rod from the Mol 7B experiment

    International Nuclear Information System (INIS)

    Coquerelle, M.; Walker, C.T.; Whitlow, W.H.

    1980-01-01

    The experiment MOL 7B was carried out in a epithermal flux in the Belgian reactor BR2. The pin examined contained fuel of initial composition (Usub(0.7)Pusub(0.3))Osub(1.98). It had been irradiated to a maximum burn-up of 13.2 at% at a maximum linear power of 568Wcm -1 . The fuel was clad with coldworked stainless steel. Electron microprobe analysis indicated that a Cr 2 O 3 type oxide was the main constituent of the grey phases in the gap. A metallic phase on the fuel surface had apparently resulted from the mechanical compaction of fragments of cladding that had been depleted in chromium by oxidation. Thus the main components of the phase were iron and nickel. Chromium loss from the inner cladding surface was significant only in the upper regions of the pin. In pin sections that were metallographically examined sigma phase and carbides of the type M 23 C 6 were present at the grain boundaries of the cladding. Cladding corrosion was not an Arrhenius function of the cladding temperature: the amount of metal lost from the inner cladding surface decreased with rise in cladding temperature above 910 K. A contributor to metal loss was the mechanical detachment of fragments of cladding which reformed as a metallic layer on the surface of the fuel. Chromium depletion and sigma phase formation at grain boundaries lowered the cohesive forces between grains which were then mechanically detached. Chromium loss from grain boundaries is mainly the results of oxidation of the cladding by the mixed oxide fuel. Data are presented to support the view that the local average O/M of the fuel determined the rate of oxidation and consequently the extent of chromium depletion. Fuel-cladding mechanical interactions were weak in the upper regions of the pin where metal loss was small

  16. Status of irradiation testing and PIE of MOX (Pu-containing) fuel

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Zhou, Y.N.; Ryz, M.A.

    1995-01-01

    This paper describes AECL's mixed oxide (MOX) fuel-irradiation and post-irradiation examination (PIE) program. Post-irradiation examination results of two major irradiation experiments involving several (U, Pu)O 2 fuel bundles are highlighted. One experiment involved bundles irradiated to burnups ranging fro 400 to 1200 MWh/kgHe in the Nuclear Power Demonstration (NPD) reactor. The other experiment consisted of several (U, Pu)O 2 bundles irradiated to burnups of up to 500 Mwh/kgHe in the National Research Universal (NRU) reactor. Results of these experiments demonstrate the excellent performance of CANDU MOX fuel. This paper also outlines the status of current MOX fuel irradiation tests, including the irradiation of various (U, Pu)O 2 bundles. The strategic importance of MOX fuel to CANDU fuel-cycle flexibility is discussed. (author)

  17. Irradiation behavior of miniature experimental uranium silicide fuel plates

    International Nuclear Information System (INIS)

    Hofman, G.L.; Neimark, L.A.; Mattas, R.F.

    1983-01-01

    Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 10 20 cm -3 , far short of the approximately 20 x 10 20 cm -3 goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix

  18. Nonintrusive irradiated fuel inventory confirmation technique

    International Nuclear Information System (INIS)

    Dowdy, E.J.; Nicholson, N.; Caldwell, J.T.

    1980-01-01

    Successful tests showing correlation between the intensity of the Cerenkov glow surrounding irradiated fuel assemblies in water-filled spent fuel storage ponds and the exposure and cooling times of assemblies have been concluded. Fieldable instruments used in subsequent tests confirmed that such measurements can be made easily and rapidly, without fuel assembly movement or the introduction of apparatus into the storage ponds

  19. Irradiation Testing of TRISO-Coated Particle Fuel in Korea

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Yeo, Sunghwan; Jeong, Kyung-Chai; Eom, Sung-Ho; Kim, Yeon-Ku; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung; Kim, Yong Wan

    2014-01-01

    In Korea, coated particle fuel is being developed to support development of a VHTR. At the end of March 2014, the first irradiation test in HANARO at KAERI to demonstrate and qualify TRISO-coated particle fuel for use in a VHTR was terminated. This experiment was conducted in an inert gas atmosphere without on-line temperature monitoring and control, or on-line fission product monitoring of the sweep gas. The irradiation device contained two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The duration of irradiation testing at HANARO was about 135 full power days from last August 2013. The maximum average power per particle was about 165 mW/particle. The calculated peak burnup of the TRISO-coated fuel was a little less than 4 atom percent. Post-irradiation examination is being carried out at KAERI’s Irradiated Material Examination Facility beginning in September of 2014. This paper describes characteristics of coated particle fuel, the design of the test rod and irradiation device for this coated particle fuel, and discusses the technical results of irradiation testing at HANARO. (author)

  20. Irradiation testing of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-01-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

  1. The post irradiation examination of a sphere-pac (UPu)C fuel pin irradiated in the BR-2 reactor (MFBS 7 experiment)

    International Nuclear Information System (INIS)

    Smith, L.; Aerne, E.T.; Buergisser, B.; Flueckiger, U.; Hofer, R.; Petrik, F.

    1979-09-01

    A pin fuelled with Swiss made (UPu)C microspheres has been successfully irradiated to a peak burn-up of 6% fima in the Belgian BR2 Reactor. The pin, rated up to 95 kW/m, was intact after irradiation and exhibited a peak strain of just over 0.5%. The results of the post irradiation examination are reported. (Auth.)

  2. Status of fuel irradiation tests in HANARO

    International Nuclear Information System (INIS)

    Kim, Hark Rho; Lee, Choong Sung; Lee, Kye Hong; Jun, Byung Jin; Lee, Ji Bok

    1999-01-01

    Since 1996 after finishing the long-term operational test, HANARO (High-Flux Advanced Neutron Application Reactor) has been extensively used for material irradiation tests, beam application research, radioisotope production and neutron activation analysis. This paper presents the fuel irradiation test activities which are now conducted or have been finished in HANARO. KAERI developed LEU fuel using an atomization method for the research reactors. Using this LEU, we have set up and conducted three irradiation programs: (1) medium power irradiation test using a short-length mini-assembly made of 3.15 gU/cc U 3 Si, (2) high power irradiation tests using full-length test assemblies made of 3.15 gU/cc U 3 Si, and (3) irradiation test using a short-length mini-plate made of 4.8 gU/cc U 3 Si 2 . DUPIC (Direct Use of spent PWR fuels in CANDU Reactors) simulation fuel pellets, of which compositions are very similar to DUPIC pellets to keep the similarity in the thermo-mechanical property, were developed. Three mini-elements including 5 pellets each were installed in a capsule. This capsule has been irradiated for 2 months and unloaded from the HANARO core at the end of September 1999. Another very important test is the HANARO fuel qualification program at high power, which is required to resolve the licensing issue. This test is imposed on the HANARO operation license due to insufficient test data under high power environment. To resolve this licensing issue, we have been carrying out the required irradiation tests and PIE (Post-irradiation Examination) tests. Through this program, it is believed that the resolution of the licensing issue is achieved. In addition to these programs, several fuel test plans are under way. Through these vigorous activities of fuel irradiation test programs, HANARO is sure to significantly contribute to the national nuclear R and D programs. (author)

  3. Advanced disassembling technique of irradiated driver fuel assembly for continuous irradiation of fuel pins

    International Nuclear Information System (INIS)

    Ichikawa, Shoichi; Haga, Hiroyuki; Katsuyama, Kozo; Maeda, Koji; Nishinoiri, Kenji

    2012-01-01

    It was necessary to carry out continuous irradiation tests in order to obtain the irradiation data of high burn-up fuel and high neutron dose material for FaCT (Fast Reactor Cycle Technology Development) project. There, the disassembling technique of an irradiated fuel assembly was advanced in order to realize further continuous irradiation tests. Although the conventional disassembling technique had been cutting a lower end-plug of a fuel pin needed to fix fuel pins to an irradiation vehicle, the advanced disassembling technique did not need cutting a lower end-plug. As a result, it was possible to supply many irradiated fuel pins to various continuous irradiation tests for FaCT project. (author)

  4. Fission gas retention in irradiated metallic fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Gruber, E.E.; Kramer, J.M.

    1987-01-01

    Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5Fs) are presented. The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations to the measurements shows quantitative agreement with both the magnitude and the axial variation of the retained gas content

  5. Neutron Flux Depression in the UO2-PuO2 (15 to 30%) Fuel Rods from IVO-FR2-Vg7-Irradiation Experiment

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Fernandez Marron, J.L.

    1983-01-01

    The thermal-neutron flux depression within a fuel rod has a great influence in the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO 2 -PUO 2 (15 to 30% PUO 2 ) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (Author) 22 refs

  6. Neutron flux depression in the UO2-PuO2 (15 to 30%) fuel rods from IVO-FR2-Vg7-Irradiation experiment

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Fernandez Marron, J.L.

    1983-01-01

    The thermal-neutron flux depression within a fuel rod has a great influence on the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO 2 -PuO 2 (15 to 30% PuO 2 ) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (author)

  7. Irradiation and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

    Bae, Ki Kwang; Yang, M. S.; Song, K. C.

    2000-05-01

    The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis

  8. Irradiation and performance evaluation of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Ki Kwang; Yang, M S; Song, K C [and others

    2000-05-01

    The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis.

  9. Irradiated nuclear fuel transport from Japan to Europe

    International Nuclear Information System (INIS)

    Kavanagh, M.T.; Shimoyama, S.

    1976-01-01

    Irradiated nuclear fuel has been transported from Japan to Europe since 1969, although U.K. experience goes back almost two decades. Both magnox and oxide fuel have been transported, and the technical requirements associated with each type of fuel are outlined. The specialized ships used by British Nuclear Fuels Limited (BNFL) for this transport are described, as well as the ships being developed for future use in the Japan trade. The ship requirements are related to the regulatory position both in the United Kingdom and internationally, and the Japanese regulatory requirements are described. Finally, specific operational experience of a Japanese reactor operator is described

  10. Irradiation testing of high density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-10-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U 2 Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions

  11. New trends in nuclear fuel experimental irradiation. Modern control and acquisition of the irradiation data

    International Nuclear Information System (INIS)

    Preda, M.; Ciocanescu, M.; Ana, E.M.

    2010-01-01

    With the irradiation devices used in the irradiation tests, the following experiments have been performed in TRIGA-SCN reactor: a) In capsule-type irradiation devices: - fission gases composition determination; - dimensional measurements; - fission gases pressure measurement; - power pre-ramp and ramp; - power cycling; - structural materials testing. b) In loop-type irradiation device: - power ramp; - multiple power ramps; - overpower. Aiming to develop irradiation tests for advanced nuclear fuel elements, it is mandatory to increase the maximum neutron flux in the core with about 20%. This will lead to reactor power increase up to 21 MW. This objective can be reached through: - increasing the number of fuel clusters in the reactor core; - using the 6x6 fuel cluster to replace the present 5x5 clusters; - relocation of the control rods. In this context, the new control system and the data acquisition system operates online and allows real-time data evaluation. (author)

  12. The physics of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Robin, M.

    1980-01-01

    The knowledge of the neutron irradiation effect is essential in dealing with all subjects related to the fuel. Neutron irradiation provokes fission reactions within the fuel and produces new nuclides. The formation chains are described and the importance of each isotope in the fuel cycle is explained with regards to its own characteristics. To solve the system of equations giving the evolution of different nuclides concentrations, the corresponding effective cross-sections and flux received are given by standard codes used for reactor calculations. A good test for calculation methods is the experimental study of irradiated fuel. Many techniques have been developed for this purpose. The last chapter compares fuel evolution in different reactors, in connection with some specific characteristics. (author)

  13. Irradiated fuel examination using the Cerenkov technique

    International Nuclear Information System (INIS)

    Nicholson, N.; Dowdy, E.J.

    1981-03-01

    A technique for monitoring irradiated nuclear fuel inventories located in water filled storage ponds has been developed and demonstrated. This technique provides sufficient qualitative information to be useful as a confirmatory technique to International Atomic Energy Agency inspectors. Measurements have been made on the Cerenkov glow light intensity from irradiated fuel that show the intensity of this light to be proportional to the cooling time. Fieldable instruments used in several tests confirm that such measurements can be made easily and rapidly, without fuel assembly movement or the introduction of apparatus into the storage ponds. The Cerenkov technique and instrumentation have been shown to be of potential use to operators of reactor spent fuel facilities and away from reactor storage facilities, and to the International Atomic Energy Agency inspectors who provide surveillance of the irradiated fuel stored in these facilities

  14. Irradiation testing of miniature fuel plates for the RERTR program

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

    1983-08-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

  15. Irradiation behavior of uranium-molybdenum dispersion fuel: Fuel performance data from RERTR-1 and RERTR-2

    International Nuclear Information System (INIS)

    Meyer, M.K.; Clark, C.R.; Hayes, S.L.; Strain, R.V.; Hofman, G.L.; Snelgrove, J.L.; Park, J.M.; Kim, K.H.

    1999-01-01

    This paper presents quantitative data on the irradiation behavior of uranium-molybdenum fuels from the low temperature RERTR-1 and -2 experiments. Fuel swelling measurements of U-Mo fuels at ∼40% and ∼70% burnup are presented. The rate of fuel-matrix interaction layer growth is estimated. Microstructures of fuel in the pre- and postirradiation condition were compared. Based on these data, a qualitative picture of the evolution of the U-Mo fuel microstructure during irradiation has been developed. Estimates of uranium-molybdenum fuel swelling and fuel-matrix interaction under high-power research reactor operating conditions are presented. (author)

  16. Loading procedures for shipment of irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bates, E F; Feltz, D E; Sandel, P S; Schoenbucher, B [Texas A and M University (United States)

    1974-07-01

    The Nuclear Science Center at Texas A and M does not have proper equipment and facilities for transferring irradiated fuel from the reactor pool to the transport vehicle. To accomplish the transfer of 23 MTR type fuel elements procedures were developed using a modified fork lift and flex-lift obtained locally. The transfer was accomplished without incident and with negligible personnel exposure. (author)

  17. Loading procedures for shipment of irradiated fuel

    International Nuclear Information System (INIS)

    Bates, E.F.; Feltz, D.E.; Sandel, P.S.; Schoenbucher, B.

    1974-01-01

    The Nuclear Science Center at Texas A and M does not have proper equipment and facilities for transferring irradiated fuel from the reactor pool to the transport vehicle. To accomplish the transfer of 23 MTR type fuel elements procedures were developed using a modified fork lift and flex-lift obtained locally. The transfer was accomplished without incident and with negligible personnel exposure. (author)

  18. Fuel fabrication and post-irradiation examination

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-06-01

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

  19. Fuel fabrication and post-irradiation examination

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  20. Horizontal modular dry irradiated fuel storage system

    Science.gov (United States)

    Fischer, Larry E.; McInnes, Ian D.; Massey, John V.

    1988-01-01

    A horizontal, modular, dry, irradiated fuel storage system (10) includes a thin-walled canister (12) for containing irradiated fuel assemblies (20), which canister (12) can be positioned in a transfer cask (14) and transported in a horizontal manner from a fuel storage pool (18), to an intermediate-term storage facility. The storage system (10) includes a plurality of dry storage modules (26) which accept the canister (12) from the transfer cask (14) and provide for appropriate shielding about the canister (12). Each module (26) also provides for air cooling of the canister (12) to remove the decay heat of the irradiated fuel assemblies (20). The modules (26) can be interlocked so that each module (26) gains additional shielding from the next adjacent module (26). Hydraulic rams (30) are provided for inserting and removing the canisters (12) from the modules (26).

  1. VVER fuel. Results of post irradiation examination

    International Nuclear Information System (INIS)

    Smirnov, V.P.; Markov, D.V.; Smirnov, A.V.; Polenok, V.S.; Perepelkin, S.O.; Ivashchenko, A.A.

    2005-01-01

    The present paper presents the main results of post-irradiation examination of more than 40 different fuel assemblies (FA) operated in the cores of VVER-1000 and VVER-440-type power reactors in a wide range of fuel burnup. The condition of fuel assembly components from the viewpoint of deformation, corrosion resistance and mechanical properties is described here. A serviceability of the FA design as a whole and interaction between individual FA components under vibration condition and mechanical load received primary emphasis. The reasons of FA damage fuel element failure in a wide range of fuel burnup are also analyzed. A possibility and ways of fuel burnup increase have been proved experimentally for the case of high-level serviceability maintenance of fuel elements to provide for advanced fuel cycles. (author)

  2. Irradiation behavior of metallic fast reactor fuels

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  3. Consolidation equipment for irradiated nuclear fuel channels

    International Nuclear Information System (INIS)

    Taguchi, M.; Komatsu, Y.; Ose, T.

    1989-01-01

    The authors have developed and put into use a new type of mechanical consolidation equipment for irradiated nuclear fuel channels. This includes round-slice cutting of the top 100mm of the fuel channel with a guillotine cutter, and press cutting of the two corners of the remaining length of the fuel channel. Four guillotine blades work in combination with receiving blades arranged inside the fuel channel to cut the top 100mm, including the clips and spacers, of the fuel channel into a round slice. A press assembled in the consolidation equipment then presses the slice to achieve volume reduction. The press cutting operation uses two press cutting blades arranged inside the fuel channel and the receiving blades outside the fuel channel. The remaining length of fuel channel is cut off into L-shaped pieces by press cutting. This consolidation equipment is highly efficient because the round-slice cutting, pressing, and press cutting are all achieved by one unit

  4. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  5. Irradiation testing of coated particle fuel at Hanaro

    International Nuclear Information System (INIS)

    Goo Kim, Bong; Sung Cho, Moo; Kim, Yong Wan

    2014-01-01

    TRISO-coated particle fuel is developing to support development of VHTR in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in VHTR in the HANARO at KAERI. This experiment is currently undergoing under the atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak burn-up of about 4 at% and a peak fast neutron fluence of about 1.7 x 10 21 n/cm 2 , PIE will be carried out at KAERI's Irradiated Material Examination Facility. This paper is described characteristics of coated particle fuel, the design of test rod and irradiation device for coated particle fuel, discusses the technical results for irradiation testing at HANARO. (authors)

  6. Review of irradiation experiments for water reactor safety research

    International Nuclear Information System (INIS)

    Tobioka, Toshiaki

    1977-02-01

    A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

  7. Post-irradiation examination and R and D programs using irradiated fuels at KAERI

    International Nuclear Information System (INIS)

    Chun, Yong Bum; Min, Duck Kee; Kim, Eun Ka and others

    2000-12-01

    This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

  8. Post-irradiation examination and R and D programs using irradiated fuels at KAERI

    International Nuclear Information System (INIS)

    Chun, Yong Bum; So, Dong Sup; Lee, Byung Doo; Lee, Song Ho; Min, Duck Kee

    2001-09-01

    This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

  9. Characterization and heading of irradiated fuels and their chemical analogs

    International Nuclear Information System (INIS)

    Serrano, J. A.

    2000-01-01

    This work presents results of leaching experiments under deionized water and under synthetic granite at room temperature in air using spent fuel (UO 2 and MOX LWR fuels) and the chemical analogues, natural UO 2 and SIMFUEL. The experimental conditions and procedure for irradiated and non-irradiated materials were kept similar as much as possible. Also dissolution behaviour studies of preoxidised LWR UO 2 and MOX spent fuel up to different on the oxidation degree. For both fuel types, UO 2 and MOX, the fission products considered showed a fractional release normalised to uranium higher than 1, due to either the larger inventory at preferential leaching zones, such as, grain boundaries or to the inherent higher solubility of some of these elements. In contrast to fission products, the fractional release of PU from the UO 2 fuel was not affected by the oxidation level. Finally a thermodynamic study of the experimental leaching results obtained in this work was performed. (Author)

  10. Fission gas retention in irradiated metallic fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Gruber, E.; Kramer, J.M.

    1987-01-01

    Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5 wt. % Fs) are presented. (The symbol 'Fs' designates fissium, a 'pseudo-element' which, in reality, is an alloy whose composition is representative of fission products that remain in reprocessed fuel). The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations with the measurements shows quantitative agreement in both the magnitude and the axial variation of the retained gas content. (orig.)

  11. Irradiated fuel performance evaluation technology development

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Bang, J. G.; Kim, D. H.

    2012-01-01

    Alpha version performance code for dual-cooled annular fuel under steady state operation, so called 'DUOS', has been developed applying performance models and proposed methodology. Furthermore, nonlinear finite element module which could be integrated into transient/accident fuel performance code was also developed and evaluated using commercial FE code. The first/second irradiation and PIE test of annular pellet for dual-cooled annular fuel in the world have been completed. In-pile irradiation test DB of annular pellet up to burnup of 10,000 MWd/MTU through the 1st test was established and cracking behavior of annular pellet and swelling rate at low temperature were studied. To do irradiation test of dual-cooled annular fuel under PWR's simulating steady-state conditions, irradiation test rig/rod design/manufacture of mock-up/performance test have been completed through international collaboration program with Halden reactor project. The irradiation test of large grain pellets has been continued from 2002 to 2011 and completed successfully. Burnup of 70,000 MWd/MTU which is the highest burnup among irradiation test pellets in domestic was achieved

  12. Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A. J.; Haggard, DC; Herter, J. W.; Swank, W. D.; Knudson, D. L.; Cherry, R. S. [Idaho National Laboratory, P.O. Box 1625, MS 4112, Idaho Falls, ID, (United States); Scervini, M. [University of Cambridge, Department of Material Science and Metallurgy, 27 Charles Babbage Road, CB3 0FS, Cambridge, (United Kingdom)

    2015-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly

  13. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    Energy Technology Data Exchange (ETDEWEB)

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

  14. WWER fuel: Results of post irradiation examination

    International Nuclear Information System (INIS)

    Markov, D.V.; Smirnov, V.P.; Smirnov, A.V.; Polenok, V.S.; Perepelkin, S.O.; Ivashchenko, A.A.

    2006-01-01

    Experience in the field of fabrication, operation, testing and post-irradiation examinations (PIE) made it possible to settle the following requirements for a new generation of WWER nuclear fuel: - For WWER-1000 FA, the service life is no less than 5 years, 3 alternative fuel cycles (FC): 12 months x 4 FCs, 12 months x 5 FCs and 18 months x 3 FCs; - For WWER-440 FA, fuel cycle is 12 months x 5 FCs and a part of operating assembly is left for the 6. year; - High fuel burnup - up to 70 MWd/kgU; - Dimensional stability of FA and its components; - FA repairability; - Adaptability of fuel cycles; - Maintenance of maneuvering operating conditions at the NPP; - Reliability of control rod operation; - High serviceability level - FE leakage is no worse than 10-5 l/year. In order to provide the fulfillment of the above-given requirements, designers and production engineers have worked out cumulative measures and engineering solutions, which are introduced in development of a new generation fuel. Currently old design FA-M assemblies provided with steel skeleton are being operated in WWER-1000 reactors at Ukrainian and Bulgarian NPPs. As for Russian NPPs, new-type FAs are operated. These are advanced FAs (AFA), FA-A and FA-2 provided with zirconium alloy skeletons. A design of the second generation of WWER-440 operating assemblies was developed with respect to changes in some geometrical parameters, fastening of FEs in the lower grid (splinting was substituted for collet), usage of reinforcing rib under the lower grid, anti-debris filter and hafnium elements of junction unit as well as hafnium content decrease from 0.05 % mass down to 0.01% mass in zirconium materials. They are basic designs of FAs in order to be introduced in a five-year fuel cycle of WWER-440 NPPs in Czech Republic and Slovakia since 2005 and have got prospects for development. The operating experience of dismountable operating assemblies at the Loviisa NPP, vibration-proof operating assemblies at the

  15. BR2 Reactor: Irradiation of fuels

    International Nuclear Information System (INIS)

    Verwimp, A.

    2005-01-01

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

  16. Irradiation behavior of experimental miniature uranium silicide fuel plates

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Neimark, L.A.; Mattas, R.F.

    1983-01-01

    Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk form, on the order of 7 x 10 20 cm -3 , far short of he approximately 20 x 10 20 cm -3 goal established for the RERTR Program. The purpose of the irradiation experiments on silicide fuels in the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix. The first group of experimental 'mini' fuel plates have recently reached the program's goal burnup and are in various stages of examination. Although the results to date indicate some limitations, it appears that within the range of parameters examined thus far the uranium silicide dispersion holds promise for satisfying most of the needs of the RERTR Program. The twelve experimental silicide dispersion fuel plates that were irradiated to approximately their goal exposure show the 30-vol % U 3 Si-Al plates to be in a stage of relatively rapid fission-gas-driven swelling at a fission density of 2 x 10 20 cm -3 . This fuel swelling will likely result in unacceptably large plate-thickness increases. The U 3 Si plates appear to be superior in this respect; however, they, too, are starting to move into the rapid fuel-swelling stage. Analysis of the currently available post irradiation data indicates that a 40-vol % dispersed fuel may offer an acceptable margin to the onset of unstable thickness changes at exposures of 2 x 10 21 fission/cm 3 . The interdiffusion between fuel and matrix

  17. Fission gas retention and axial expansion of irradiated metallic fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

    1986-05-01

    Out-of-reactor experiments utilizing direct electrical heating and infrared heating techniques were performed on irradiated metallic fuel. The results indicate accelerated expansion can occur during thermal transients and that the accelerated expansion is driven by retained fission gases. The results also demonstrate gas retention and, hence, expansion behavior is a function of axial position within the pin

  18. Cerenkov methodology for monitoring irradiated reactor fuel

    International Nuclear Information System (INIS)

    Nicholson, N.; Dowdy, E.J.

    1984-01-01

    Attribute measurement methods for confirming declared irradiated fuel inventories at nuclear installations under safeguards surveillance are of significant interest to inspectors. High-gain measurements of the intensity of the Cerenkov glow from exposed assemblies in water-filled storage ponds are promising for this purpose because the measured intensities depend on cooling times and burnup. We have developed a Cerenkov Measuring Device, a hand-held instrument that examines irradiated fuel assemblies in water-filled storage ponds and measures the intensity of the associated Cerenkov glow. In addition, we have developed a method for making such high-gain measurements in the presence of intense ambient light

  19. Metal fuel manufacturing and irradiation performance

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Walters, L.C.

    1992-01-01

    The advances in metal fuel by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, and improved passive safety. The goals and the safety philosophy of the Integral Fast Reactor Program are stressed

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

    Energy Technology Data Exchange (ETDEWEB)

    Williams, J.A.

    1977-09-01

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

  1. Nitride fuels irradiation performance data base

    International Nuclear Information System (INIS)

    Brozak, D.E.; Thomas, J.K.; Peddicord, K.L.

    1987-01-01

    An irradiation performance data base for nitride fuels has been developed from an extensive literature search and review that emphasized uranium nitride, but also included performance data for mixed nitrides [(U,Pu)N] and carbonitrides [(U,Pu)C,N] to increase the quantity and depth of pin data available. This work represents a very extensive effort to systematically collect and organize irradiation data for nitride-based fuels. The data base has many potential applications. First, it can facilitate parametric studies of nitride-based fuels to be performed using a wide range of pin designs and operating conditions. This should aid in the identification of important parameters and design requirements for multimegawatt and SP-100 fuel systems. Secondly, the data base can be used to evaluate fuel performance models. For detailed studies, it can serve as a guide to selecting a small group of pin specimens for extensive characterization. Finally, the data base will serve as an easily accessible and expandable source of irradiation performance information for nitride fuels

  2. Segmented fuel irradiation program: investigation on advanced materials

    International Nuclear Information System (INIS)

    Uchida, H.; Goto, K.; Sabate, R.; Abeta, S.; Baba, T.; Matias, E. de; Alonso, J.

    1999-01-01

    The Segmented Fuel Irradiation Program, started in 1991, is a collaboration between the Japanese organisations Nuclear Power Engineering Corporation (NUPEC), the Kansai Electric Power Co., Inc. (KEPCO) representing other Japanese utilities, and Mitsubishi Heavy Industries, Ltd. (MHI); and the Spanish Organisations Empresa Nacional de Electricidad, S.A. (ENDESA) representing A.N. Vandellos 2, and Empresa Nacional Uranio, S.A. (ENUSA); with the collaboration of Westinghouse. The objective of the Program is to make substantial contribution to the development of advanced cladding and fuel materials for better performance at high burn-up and under operational power transients. For this Program, segmented fuel rods were selected as the most appropriate vehicle to accomplish the aforementioned objective. Thus, a large number of fuel and cladding combinations are provided while minimising the total amount of new material, at the same time, facilitating an eventual irradiation extension in a test reactor. The Program consists of three major phases: phase I: design, licensing, fabrication and characterisation of the assemblies carrying the segmented rods (1991 - 1994); phase II: base irradiation of the assemblies at Vandellos 2 NPP, and on-site examination at the end of four cycles (1994-1999). Phase III: ramp testing at the Studsvik facilities and hot cell PIE (1996-2001). The main fuel design features whose effects on fuel behaviour are being analysed are: alloy composition (MDA and ZIRLO vs. Zircaloy-4); tubing texture; pellet grain size. The Program is progressing satisfactorily as planned. The base irradiation is completed in the first quarter of 1999, and so far, tests and inspections already carried out are providing useful information on the behaviour of the new materials. Also, the Program is delivering a well characterized fuel material, irradiated in a commercial reactor, which can be further used in other fuel behaviour experiments. The paper presents the main

  3. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  4. Power-to-melt evaluation of fresh mixed-oxide fast reactor fuel. Technical improvements of the post-irradiation-experiment and the evaluation of the results for the power-to-melt test PTM-2 in 'JOYO'

    International Nuclear Information System (INIS)

    Yamamoto, Kazuya; Kushida, Naoya; Koizumi, Atsuhiro

    1999-11-01

    The second Power-To-Melt (PTM) test, PTM-2, was performed in the experimental fast reactor 'JOYO'. All of the twenty-four fuel pins of the irradiation vehicle, B5D-2, for the PTM-2 test, were provided for post-irradiation-experiment (PIE) to evaluate the PTM values. In this study, the PIE technique for PTM test was established and the PTM results were evaluated. The findings are as follows: The maximum fuel-melting ratio on the transverse section was 10.7%, and was within the limit of fuel-melting in this PTM test enough. Unexpected fuel-melting amount to a ratio of 11.8% was found at ∼24 mm below the peak power elevation in a test fuel pin. It is possible that this arose from secondary fuel-melting. Combination of metallographical observation with X-ray microanalysis of plutonium distribution was very effective for the identification of once-molten fuel zone. The PTM evaluation suggested that dependence of the PTM on the fuel pellet density was stronger than that of previous foreign PTM tests, while the dependence on the pellet-cladding gap and the oxygen-to-metal ratio was indistinctly. The dependence on the cladding temperature and the fill gas composition was not shown as well. (author)

  5. Calculation simulation of equivalent irradiation swelling for dispersion nuclear fuel

    International Nuclear Information System (INIS)

    Cai Wei; Zhao Yunmei; Gong Xin; Ding Shurong; Huo Yongzhong

    2015-01-01

    The dispersion nuclear fuel was regarded as a kind of special particle composites. Assuming that the fuel particles are periodically distributed in the dispersion nuclear fuel meat, the finite element model to calculate its equivalent irradiation swelling was developed with the method of computational micro-mechanics. Considering irradiation swelling in the fuel particles and the irradiation hardening effect in the metal matrix, the stress update algorithms were established respectively for the fuel particles and metal matrix. The corresponding user subroutines were programmed, and the finite element simulation of equivalent irradiation swelling for the fuel meat was performed in Abaqus. The effects of the particle size and volume fraction on the equivalent irradiation swelling were investigated, and the fitting formula of equivalent irradiation swelling was obtained. The results indicate that the main factors to influence equivalent irradiation swelling of the fuel meat are the irradiation swelling and volume fraction of fuel particles. (authors)

  6. Dearomatization of jet fuel on irradiated platinum-supported catalyst

    International Nuclear Information System (INIS)

    Mucka, V.; Ostrihonova, A.; Kopernicky, I.; Mikula, O.

    1983-01-01

    The effect of ionizing radiation ( 60 Co #betta#-rays) on Pt-supported catalyst used for the dearomatization of jet fuel with distillation in the range 395 to 534 K has been studied. Pre-irradiation of the catalyst with doses in the range 10 2 to 5 x 10 4 Gy leads to the partial catalyst activation. Irradiation of the catalyst enhances its resistance to catalyst poisons, particularly to sulphur-compounds, and this is probably the reason for its catalytic activity being approx. 60 to 100% greater than that of un-irradiated catalyst. Optimum conditions for dearomatization on the irradiated catalyst were found and, by means of a rotary three-factorial experiment, it was shown that these lie at lower temperatures and lower pressures than those for un-irradiated catalyst. (author)

  7. Safe transport of irradiated fuel by sea

    International Nuclear Information System (INIS)

    Miller, M.L.

    1997-01-01

    The development is described of a transport system dedicated to the sea transport of irradiated nuclear fuel. The background is reviewed of why shipments were required and the establishment of a specialist shipping company, Pacific Nuclear Transport Limited. A description of the ships, flasks and other equipment utilised is provided, together with details of key procedures implemented to ensure safety and customer satisfaction. (Author)

  8. The sea transport of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Miller, M.L.

    1995-01-01

    The paper describes the development of a transport system dedicated to the sea transport of irradiated nuclear fuel. It reviews the background to why shipments were required and the establishment of a specialist shipping company, Pacific Nuclear Transport Limited. A description of the ships, flasks and other equipment utilized is provided, together with details of key procedures implemented to ensure safety and customer satisfaction

  9. Behaviour of irradiated uranium silicide fuel revisited

    International Nuclear Information System (INIS)

    Finlay, M. Ross; Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

    2002-01-01

    Irradiated U 3 Si 2 dispersion fuels demonstrate very low levels of swelling, even at extremely high burn-up. This behaviour is attributed to the stability of fission gas bubbles that develop during irradiation. The bubbles remain uniformly distributed throughout the fuel and show no obvious signs of coalescence. Close examination of high burn-up samples during the U 3 Si 2 qualification program revealed a bimodal distribution of fission gas bubbles. Those observations suggested that an underlying microstructure was responsible for the behaviour. An irradiation induced recrystallisation model was developed that relied on the presence of sufficient grain boundary surface to trap and pin fission gas bubbles and prevent coalescence. However, more recent work has revealed that the U 3 Si 2 becomes amorphous almost instantaneously upon irradiation. Consequently, the recrystallisation model does not adequately explain the nucleation and growth of fission gas bubbles in U 3 Si 2 . Whilst it appears to work well within the range of measured data, it cannot be relied on to extrapolate beyond that range since it is not mechanistically valid. A review of the mini-plates irradiated in the Oak Ridge Research Reactor from the U 3 Si 2 qualification program has been performed. This has yielded a new understanding of U 3 Si 2 behaviour under irradiation. (author)

  10. Canadian CANDU fuel development program and recent fuel operating experience

    International Nuclear Information System (INIS)

    Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

    1999-01-01

    This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the 2-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These 2 programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

  11. Irradiation test plan of the simulated DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Ki Kwang; Yang, M. S.; Kim, B. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    Simulated DUPIC fuel had been irradiated from Aug. 4, 1999 to Oct. 4 1999, in order to produce the data of its in-core behavior, to verify the design of DUPIC non-instrumented capsule developed, and to ensure the irradiation requirements of DUPIC fuel at HANARO. The welding process was certified for manufacturing the mini-element, and simulated DUPIC fuel rods were manufactured with simulated DUPIC pellets through examination and test. The non-instrumented capsule for a irradiation test of DUPIC fuel has been designed and manufactured referring to the design specification of the HANARO fuel. This is to be the design basis of the instrumented capsule under consideration. The verification experiment, whether the capsule loaded in the OR4 hole meet the HANARO requirements under the normal operation condition, as well as the structural analysis was carried out. The items for this experiment were the pressure drop test, vibration test, integrity test, et. al. It was noted that each experimental result meet the HANARO operational requirements. For the safety analysis of the DUPIC non-instrumented capsule loaded in the HANARO core, the nuclear/mechanical compatibility, thermodynamic compatibility, integrity analysis of the irradiation samples according to the reactor condition as well as the safety analysis of the HANARO were performed. Besides, the core reactivity effects were discussed during the irradiation test of the DUPIC capsule. The average power of each fuel rod in the DUPIC capsule was calculated, and maximal linear power reflecting the axial peaking power factor from the MCNP results was evaluated. From these calculation results, the HANARO core safety was evaluated. At the end of this report, similar overseas cases were introduced. 9 refs., 16 figs., 10 tabs. (Author)

  12. Behavior of pre-irradiated fuel under a simulated RIA condition

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide

    1994-07-01

    This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-3, conducted in the Nuclear Safety Research Reactor (NSSR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and analyses, interpretations, and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 19.6MWd/kgU with average linear heat rate of 25.3 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a deposited energy of 174±6 cal/g·fuel and a peak fuel enthalpy of 130±5 cal/g·fuel under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The cladding surface temperature increased to only 150degC, and the test resulted in slight fuel deformation and no fuel failure. An estimated rod-average fission gas release during the transient was about 2.2%. Through the detailed fuel examinations, the information concerning microstructural change in the fuel pellets were also obtained. (author)

  13. The spent fuel safety experiment

    International Nuclear Information System (INIS)

    Harmms, G.A.; Davis, F.J.; Ford, J.T.

    1995-01-01

    The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

  14. Post irradiation examination on test fuel pins for PWR

    International Nuclear Information System (INIS)

    Fogaca Filho, N.; Ambrozio Filho, F.

    1981-01-01

    Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

  15. Irradiation performance of full-length metallic IFR fuels

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.

    1992-07-01

    An assembly irradiation of 169 full-length U-Pu-Zr metallic fuel pins was successfully completed in FFTF to a goal burnup of 10 at.%. All test fuel pins maintained their cladding integrity during the irradiation. Postirradiation examination showed minimal fuel/cladding mechanical interaction and excellent stability of the fuel column. Fission-gas release was normal and consistent with the existing data base from irradiation testing of shorter metallic fuel pins in EBR-II

  16. System of leak inspection of irradiated fuel

    International Nuclear Information System (INIS)

    Delfin L, A.; Castaneda J, G.; Mazon R, R.; Aguilar H, F.

    2007-01-01

    The International Atomic Energy Agency (IAEA) through the project RLA/04/18 Irradiated Fuel Management in Research reactors, recommended among other that the participant countries (Brazil, Argentina, Chile, Peru and Mexico), develop the sipping tool to generate registrations of the state that keep the irradiated fuels in the facilities of each country. The TRIGA Mark lll Reactor (RTMIII) Department, generated a project that it is based on the dimensions of the used fuel by the RTMIII, for design and to build an inspection system of irradiated fuel well known as SIPPING. This technique, provides a high grade of accuracy in the detection of gassy fission products or liquids that escape from the enveloping of fuels that have flaws or flights. The operation process of the SIPPING is carried out generating the migration of fission products through the creation of a pressure differential gas or vacuum to identify fuel assemblies failed by means of the detection of the xenon and/or krypton presence. The SIPPING system, is a device in revolver form with 4 tangential nozzles, which will discharge the fluid between the external surface of the enveloping of the fuel and the interior surface of the encircling one; the device was designed with independent pieces, with threaded joining and with stamps to impede flights of the fluid toward the exterior of the system. The System homogenizes and it distributes the fluid pressure so that the 4 nozzles work to equality of conditions, for what the device was designed in 3 pieces, an internal that is denominated revolver, one external that calls cover, and a joining called mamelon that will unite with the main encircling of the system. The detection of fission products in failed fuels, its require that inside the encircling one where the irradiated fuel element is introduced, be generated a pressure differential of gas or vacuum, and that it allows the samples extraction of water. For what generated a top for the encircling with the

  17. Fuel performance experience

    International Nuclear Information System (INIS)

    Sofer, G.A.

    1986-01-01

    The history of LWR fuel supply has been characterized by a wide range of design developments and fuel cycle cost improvements. Exxon Nuclear Company, Inc. has pursued an aggressive fuel research and development program aimed at improved fuel performance. Exxon Nuclear has introduced many design innovations which have improved fuel cycle economics and operating flexibility while fuel failures remain at very low levels. The removable upper tie plate feature of Exxon Nuclear assemblies has helped accelerate this development, enabling repeated inspections during successive plant outages. Also, this design feature has made it possible to repair damaged fuel assemblies during refueling outages, thereby minimizing the economic impact of fuel failure from all causes

  18. The Experience of Storage and Shipment for Reprocessing of HEU Nuclear Fuel Irradiated in the IRT-M Research Reactor and Pamir-630 Mobile Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sikorin, S. N.; Polazau, S. A.; Luneu, A. N.; Hrigarovich, T. K. [Joint Institute for Power and Nuclear Research–Sosny of the National Academy of Sciences of Belarus, Minsk (Belarus)

    2014-08-15

    At the end of 2010 under the Global Threat Reduction Initiative (GTRI), the Joint Institute for Power and Nuclear Research–“Sosny” (JIPNR–Sosny) of the National Academy of Sciences of the Republic of Belarus repatriated HEU spent nuclear fuel to the Russian Federation. The spent nuclear fuel was from the decommissioned Pamir-630D mobile reactor and IRT-M research reactor. The paper discusses the Pamir-630D spent nuclear fuel; experience and problems of spent nuclear fuel storage; and various aspects of the shipment including legal framework, preparation activities and shipment logistics. The conceptual project of a new research reactor for Belarus is also presented.

  19. An Analysis of the Thermal and Structure Behaviour of the UO2-PuO2-Fuel in the Irradiation Experiment of the UO2-PuO2-Fuel in the Irradiation Experiment FR2 Capsule Test Series 5a

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Helmut, E.

    1981-01-01

    In the Karlsruhe research reactor FR2 nine fuel pins were irradiated within three irradiation capsules in the course of the test series 5a. The pins contained UO 2 -PuO 2 fuel pellets. They reached bump values of about 6, 17 and 47 Mwd/Kg Me with linear rod powers of 400 to 600 W/cm and clad surface temperature between 500 and 700 degree centigree. A detailed analysis of the fuel structuration data (columnar-grain and equiaxed- -grain growth regions) have allowed to determine, with the help of physic-mathematical models, the radii of these regions and the heat transfer through the contact zone between fuel and clad depending on the bump. The results of the analysis showed that the fuel surface temperature rose with increasing burnup. (Author) 16 refs

  20. EDRP public local inquiry, UKAEA/BNFL precognition on: the transport of irradiated fuel by rail

    International Nuclear Information System (INIS)

    Singleton, Leslie

    1986-02-01

    The experience in handling of irradiated fuel flasks by the British Railways Board is outlined. The steps taken to ensure the effective and safe transport of irradiated fuel and nuclear waste by rail are identified. It is concluded that the proposed rail transport link to the EDRP at Dounreay should prove practicable. (UK)

  1. Development of cutting device for irradiated fuel rod

    International Nuclear Information System (INIS)

    Lee, E. P.; Jun, Y. B.; Hong, K. P.; Min, D. K.; Lee, H. K.; Su, H. S.; Kim, K. S.; Kwon, H. M.; Joo, Y. S.; Yoo, K. S.; Joo, J. S.; Kim, E. K.

    2004-01-01

    Post Irradiation Examination(PIE) on irradiated fuel rods is essential for the evaluation of integrity and irradiation performance of fuel rods of commercial reactor fuel. For PIE, fuel rods should be cut very precisely. The cutting positions selected from NDT data are very important for further destructive examination and analysis. A fuel rod cutting device was developed witch can cut fuel rods longitudinal very precisely and can also cut the fuels into the same length rod cuts repeatedly. It is also easy to remove the fuel cutting powder after cutting works and it can extend the life time of cutting device and lower the contamination level of hot cell

  2. Cost of transporting irradiated fuels and maintenance costs of a chemical treatment plant for irradiated fuels

    International Nuclear Information System (INIS)

    Sousselier, Y.

    1964-01-01

    Numerous studies have been made of the cost of a fuel cycle, but many of them are based on a priori studies and are therefore to be treated with reserve. Thus, in the part dealing with the treatment of irradiated fuels, some important factors in the cost have only rarely been given on the basis of practical experience: the cost of transporting the fuels themselves and the plant maintenance costs. Investigations relating to transport costs are generally based on calculations made from somewhat arbitrary data. The studies carried out in France on the transport of irradiated uranium between the EDF reactors at Chinon and the retreatment plant at La Hague of the irradiated uranium from research reactors to foreign retreatment plants, are reported; they show that by a suitable choice of transport containers and details of expedition it has been possible to reduce the costs very considerably. This has been achieved either by combining rail and road transport or by increasing the writ capacities of the transport containers: an example is given of a container for swimming-pool pile elements which can transport a complete pile core at one time, thus substantially reducing the cost. Studies concerning the maintenance costs of retreatment plants are rarer still, although in direct maintenance plants these figures represent an appreciable fraction of the total treatment cost. An attempt has been made, on the basis of operational experience of a plant, to obtain some idea of these costs. Only maintenance proper has been considered, excluding subsidiary operations such as the final decontamination of apparatus, the burial of contaminated material and radioprotection operations Maintenance has been divided into three sections: mechanical maintenance, maintenance of electrical equipment and maintenance of control and adjustment apparatus. In each of these sections the distinction has been made between manpower and the material side. In order to allow comparisons to be made with

  3. Composite fuel behaviour under and after irradiation

    International Nuclear Information System (INIS)

    Dehaudt, P.; Mocellin, A.; Eminet, G.; Caillot, L.; Delette, G.; Bauer, M.; Viallard, I.

    1997-01-01

    Two kinds of composite fuels have been irradiated in the SILOE reactor. They are made of UO 2 particles dispersed in a molybdenum metallic (CERMET) or a MgAl 2 O 4 ceramic (CERCER) matrix. The irradiation conditions have allowed to reach a 50000 MWd/t U burn-up in these composite fuels after a hundred equivalent full power days long irradiation. The irradiation is controlled by a continuous measure of the pellet centre line temperature. It allows to have information about the TANOX rods thermal behaviour and the fuels thermal conductivities in comparing the centre line temperature versus linear power curves among themselves. Our results show that the CERMET centre line temperature is much lower than the CERCER and UO 2 ones: 520 deg. C against 980 deg. C at a 300W/cm linear power. After pin puncturing tests the rods are dismantled to recover each fuel pellet. In the CERCER case, the cladding peeling off has revealed that the fuel came into contact with the cladding and that some of the pellets were linked together. Optical microscopy observations show a changing of the MgAl 2 O 4 matrix state around the UO 2 particles at the pellets periphery. This transformation may have caused a swelling and would be at the origin of the pellet-cladding and the pellet-pellet interactions. No specific damage is seen after irradiation. The CERMET pellets are not cracked and remain as they were before irradiation. The CERCER crack network is slightly different from that observed in UO 2 . Kr retention was evaluated by annealing tests under vacuum at 1580 deg. C or 1700 deg. C for 30 minutes. The CERMET fission gas release is lower than the CERCER one. Inter- and intragranular fission gas bubbles are observed in the UO 2 particles after heat treatments. The CERCER pellet periphery has also cracked and the matrix has transformed again around UO 2 particles to present a granular and porous aspect. (author). 4 refs, 6 figs, 2 tabs

  4. Leaching of irradiated CANDU UO2 fuel

    International Nuclear Information System (INIS)

    Vandergraaf, T.T.; Johnson, L.H.; Lau, D.W.P.

    1980-01-01

    Irradiated fuel, leached at room temperature with distilled water and with slightly chlorinated river water, releases approx. 4% of its cesium inventory over a comparatively sort period of a few days but releases its actinides and rare earths more slowly. The matrix itself dissolves at a rate conservatively calculated to be less than approx. 2 x 10 -6 g UO 2 /cm 2 day and, with time, the leach rates of the various nuclides approach this value

  5. Irradiation test and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Song, K. C.; Moon, J. S.

    2002-05-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  6. Fuel reliability experience in Finland

    International Nuclear Information System (INIS)

    Kekkonen, L.

    2015-01-01

    Four nuclear reactors have operated in Finland now for 35-38 years. The two VVER-440 units at Loviisa Nuclear Power Plant are operated by Fortum and two BWR’s in Olkiluoto are operated by Teollisuuden Voima Oyj (TVO). The fuel reliability experience of the four reactors operating currently in Finland has been very good and the fuel failure rates have been very low. Systematic inspection of spent fuel assemblies, and especially all failed assemblies, is a good practice that is employed in Finland in order to improve fuel reliability and operational safety. Investigation of the root cause of fuel failures is important in developing ways to prevent similar failures in the future. The operational and fuel reliability experience at the Loviisa Nuclear Power Plant has been reported also earlier in the international seminars on WWER Fuel Performance, Modelling and Experimental Support. In this paper the information on fuel reliability experience at Loviisa NPP is updated and also a short summary of the fuel reliability experience at Olkiluoto NPP is given. Keywords: VVER-440, fuel reliability, operational experience, poolside inspections, fuel failure identification. (author)

  7. Results of Microstructural Examinations of Irradiated LEU U-Mo Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, ID 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organization (Australia)

    2009-06-15

    Introduction: The RERTR program is responsible for converting research reactors that use high-enriched uranium fuels to ones that use low-enriched uranium fuels [1]. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. Based on the results of initial fuel plate testing, adjustments have been made to the characteristics of fuel plates to improve the stability of the fuel microstructure. One improvement has been to add Si to the matrix of a dispersion fuel. This material is also being added at the fuel/cladding interface of a monolithic fuel. This paper will discuss the irradiation performance of these fuels, in terms of the stability of their microstructures during irradiation. Results and discussion: The post-irradiation examinations of fuel plates are performed at the Idaho National Laboratory. These examinations consist of visual examinations of fuel plates, gamma scanning, thickness measurements, oxide thickness measurements, and optical metallographic examinations of the fuel plate microstructures. Microstructural analysis is also performed using scanning electron microscopy. Overall, U-7Mo and U-10Mo alloy fuels have displayed the best irradiation performance, particularly, when a Si-containing Al alloy is used as the dispersion fuel matrix. The benefit of using this type of matrix is that the commonly observed fuel/cladding interaction that occurs during irradiation is reduced and the interaction layer that forms exhibit stable behavior during irradiation. Monolithic-type fuels, which consist of a U-Mo foil encased in Al alloy cladding, are also being developed. These types of fuels are also showing promise and will continue to be developed. One challenge with this type of fuel is in trying to maximize the bond strength at the foil/cladding interface. Fuel/cladding interactions can affect the quality of the boding at this interface. Si is being added to improve the characteristics

  8. Nondestructive assay methods for irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Hsue, S.T.; Crane, T.W.; Talbert, W.L. Jr.; Lee, J.C.

    1978-01-01

    This report is a review of the status of nondestructive assay (NDA) methods used to determine burnup and fissile content of irradiated nuclear fuels. The gamma-spectroscopy method measures gamma activities of certain fission products that are proportional to the burnup. Problems associated with this method are migration of the fission products and gamma-ray attenuation through the relatively dense fuel material. The attenuation correction is complicated by generally unknown activity distributions within the assemblies. The neutron methods, which usually involve active interrogation and prompt or delayed signal counting, are designed to assay the fissile content of the spent-fuel elements. Systems to assay highly enriched spent-fuel assemblies have been tested extensively. Feasibility studies have been reported of systems to assay light-water reactor spent-fuel assemblies. The slowing-down spectrometer and neutron resonance absorption methods can distinguish between the uranium and plutonium fissile contents, but they are limited to the assay of individual rods. We have summarized the status of NDA techniques for spent-fuel assay and present some subjects in need of further investigation. Accuracy of the burnup calculations for power reactors is also reviewed

  9. Prototypic fabrication of TRIGA irradiated fuel shipping casks

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  10. 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)

  11. Irradiation of UO2+x fuels in the TANOX device

    International Nuclear Information System (INIS)

    Dehaudt, P.; Caillot, L.; Delette, G.; Eminet, G.; Mocellin, A.

    1998-01-01

    The TANOX analytical irradiation device is presented and the first results concerning stoichiometric and hyper stoichiometric uranium dioxide fuels with two different grain sizes are given. The TANOX device is designed to obtain rapidly significant burnups in fuels at relatively low temperatures. It is placed at the periphery of the SILOE reactor and translated to adjust the irradiation power. The continuous measure of the centre-line temperature allows to control the experiment and to evaluate the thermal behaviour of the rods. A TANOX fuel rod has a length of 100 mm with 20 fuel pellets in a stainless steel cladding and is inserted in a thick aluminium alloy overcladding which is cooled by the primary water circuit reactor. These conditions of small size pellets and improved thermal exchanges have been designed to dissipate the heat power due to fission densities three to five times higher than in a PWR. The first analytical irradiation was devoted to the study of UO 2.00 , UO 2.01 and UO 2.02 fuels with standard and large grain sizes obtained by annealing. A burnup of about 9000 MWd.t -1 U was reached in these fuels. The thermal analysis shows a degraded conductivity for the UO 2.02 fuel rod due to the hyper stoichiometry. The released fractions of 85 Kr during irradiation are negligible as expected (lower than 0,1%). Some of the pellets were heat treated at 1700 deg. C for 5 hours. The gas release was analysed after 30 minutes and at the end of the treatment. The main results are as follows: the fission gas release (FGR) of the standard UO 2 varies from one sample to another; the FGR of the hyper stoichiometric fuels is of the same order of magnitude than that of the stoichiometric UO 2 fuel of normal grain sizes; the grain size increase has no effect on FGR for UO 2.00 but considerably decreases the FGR for UO 2.01 and UO 2.02 fuels. These heat treated samples are also observed to characterize the inter- and intragranular fission gas bubbles. (author)

  12. PIE Report on the KOMO-3 Irradiation Test Fuels

    International Nuclear Information System (INIS)

    Park, Jong Man; Ryu, H. J.; Yang, J. H.

    2009-04-01

    In the KOMO-3, in-reactor irradiation test had been performed for 12 kinds of dispersed U-Mo fuel rods, a multi wire fuel rod and a tube fuel rod. In this report we described the PIE results on the KOMO-3 irradiation test fuels. The interaction layer thickness between fuel particle and matrix could be reduced by using a large size U-Mo fuel particle or introducing Al-Si matrix or adding the third element in the U-Mo particle. Monolithic fuel rod of multi-wire or tube fuel was also effective in reducing the interaction layer thickness

  13. The reprocessing of irradiated fuels by halides and their compounds

    International Nuclear Information System (INIS)

    Bourgeois, M.; Faugeras, P.

    1964-01-01

    A brief description is given of the experiments leading to the choice of the process volatilization of fluorides by gas phase attack. The chemical process is described for certain current types of clad Fuels: the aluminium or the zirconium cladding is first volatilized as chloride by attack with gaseous hydrogen chloride. The uranium is then transformed into volatile hexafluoride by attack with fluorine. These reactions are carried out consecutively in the same reactor in the presence of a fluidized bed of alumina which facilitates heat exchange. The experiments have been carried out in quantities from 100 gms to several kilograms of fuel, first without activity, and then with tracers. A description is given of the laboratory research which was carried out simultaneously on the separation of uranium and plutonium fluorides. Finally, an apparatus is described which is intended to test the process on irradiated fuel at an activity level of several thousands of curies of fission products. (authors) [fr

  14. Irradiation of novel MTR fuel plates in BR2

    International Nuclear Information System (INIS)

    Verboomen, B.; Aoust, Th.; Beeckmans De Westmeerbeeck, A.; De Raedt, Ch.

    2000-01-01

    Since the end of 1999, novel MTR fuel plates with very high-density meat are being irradiated in BR2. The purpose of the irradiation is to investigate the behaviour of these fuel plates under very severe reactor operation conditions. The novel fuel plates are inserted in two standard six-tube BR2 fuel elements in the locations normally occupied by the standard outer fuel plates. The irradiation in BR2 was prepared by carrying out detailed neutron Monte Carlo calculations of the whole BR2 core containing the two experimental fuel elements for various positions in the reactor and for various azimuthal orientations of the fuel elements. Comparing the thus determined fission density levels and azimuthal profiles in the new MTR fuel plates irradiated in the various channels allowed the experimenters to choose the most appropriate BR2 channel and the most appropriate fuel element orientation. (author)

  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. Characterization of an irradiated RERTR-7 fuel plate using transmission electron microscopy

    International Nuclear Information System (INIS)

    Gan, J.; Keiser, D.D. Jr.; Miller, B.D.; Robinson, A.B.; Medvedev, P.

    2010-01-01

    Transmission electron microscopy (TEM) has been used to characterize an irradiated fuel plate with Al-2Si matrix from the Reduced Enrichment Research and Test Reactor RERTR-7 experiment that was irradiated under moderate reactor conditions. The results of this work showed the presence of a bubble superlattice within the U-7Mo grains that accommodated fission gases (e.g., Xe). The presence of this structure helps the U-7Mo exhibit a stable swelling behaviour during irradiation. Furthermore, TEM analysis showed that the Si-rich interaction layers that develop around the fuel particles at the U-7Mo/matrix interface during fuel plate fabrication and irradiation become amorphous during irradiation. An important question that remains to be answered about the irradiation behaviour of U-Mo dispersion fuels is how do more aggressive irradiation conditions affect the behaviour of fission gases within the U-7Mo fuel particles and in the amorphous interaction layers on the microstructural scale that can be characterized using TEM? This paper will discuss the results of TEM analysis that was performed on a sample taken from an irradiated RERTR-7 fuel plate with Al-2Si matrix. This plate was exposed to more aggressive irradiation conditions than the RERTR-6 plate. The microstructural features present within the U-7Mo and the amorphous interaction layers will be discussed. The results of this analysis will be compared to what was observed in the earlier RERTR-6 fuel plate characterization. (author)

  17. Behavior of irradiated ATR/MOX fuel under reactivity initiated accident conditions (Joint research)

    International Nuclear Information System (INIS)

    Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi

    2000-03-01

    Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20 MWd/kgHM were conducted at the NSRR in JAERI to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Deformation of the fuel rods due to PCMI occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet. (author)

  18. CERCA'S experience in UMO fuel manufacturing

    International Nuclear Information System (INIS)

    Jarousse, Ch.; Lavastre, Y.; Grasse, M.

    2003-01-01

    Considered as a suitable solution for non-proliferation and reprocessing purposes, UMo fuel has been chosen and studied by the RERTR program since 1996. Involved in the RERTR fuel developments since 1978, with more than 20 years of U 3 SI 2 fuel production, and closely linked to the French Commissariat a l'Energie Atomique, CERCA was able to define properly, from the beginning, the right R and D actions plan for UMo fuel development. CERCA has already demonstrated during the last 4 years its ability to manufacture plates and fuel elements with high density UMo fuel. UMo full size plates produced for 4 irradiation experiments in 3 European reactors afforded us a unique experience. In addition, as a main part of our R and D effort, we have always studied in depth a key part of the CERCA process outline which is the plate rolling stage. After some preliminary investigation in order to define the phenomenological model describing the behavior of the fuel core when rolling, we have developed a rolling digital simulator. (author)

  19. High-water-base hydraulic fluid-irradiation experiments

    International Nuclear Information System (INIS)

    Bradley, E.C.; Meacham, S.A.

    1981-10-01

    A remote system for shearing spent nuclear fuel assemblies is being designed under the direction of the Consolidated Fuel Reprocessing Program (CFRP). The design incorporates a dual hydraulic fluid actuation system in which only one of the fluids, a high-water-base (HWBF), would be exposed to ionizing radiation and radioactive contamination. A commercially available synthetic, solution-type HWBF was selected as the reference. Single-sample irradiation experiments were conducted with three commercial fluids over a range of irradiation exposures. The physical and chemical properties of the irradiated HWBFs were analyzed and compared with unirradiated samples. In general, the results of the analyses showed increasing degradation of fluid properties with increasing irradiation dose. The results also indicated that a synthetic solution-type HWBF would perform satisfactorily in the remote shear system where irradiation doses up to 10 6 Gy (10 8 rad) are expected

  20. High-water-base hydraulic fluid-irradiation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, E.C.; Meacham, S.A.

    1981-10-01

    A remote system for shearing spent nuclear fuel assemblies is being designed under the direction of the Consolidated Fuel Reprocessing Program (CFRP). The design incorporates a dual hydraulic fluid actuation system in which only one of the fluids, a high-water-base (HWBF), would be exposed to ionizing radiation and radioactive contamination. A commercially available synthetic, solution-type HWBF was selected as the reference. Single-sample irradiation experiments were conducted with three commercial fluids over a range of irradiation exposures. The physical and chemical properties of the irradiated HWBFs were analyzed and compared with unirradiated samples. In general, the results of the analyses showed increasing degradation of fluid properties with increasing irradiation dose. The results also indicated that a synthetic solution-type HWBF would perform satisfactorily in the remote shear system where irradiation doses up to 10/sup 6/ Gy (10/sup 8/ rad) are expected.

  1. Dry storage of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Tolmie, R.D.

    1983-01-01

    In transferring radioactive material between the preparation and clean chambers of a dry storage complex, irradiated nuclear fuel is posted from the preparation chamber to a sealable canister supported in a closable bucket in the clean chamber, or a contaminated sealed canister is posted from a closed bucket in the clean chamber into the preparation chamber by using a facility comprising two coaxial tubes constituting a closable orifice between the two chambers, the tubes providing sealing means for the bucket, and masking means for the bucket and canister closures together with means for withdrawing the closures into the preparation chamber. (author)

  2. Fabrication of Fast Reactor Fuel Pins for Test Irradiations

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-09-15

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

  3. Pyroelectrochemical process for reprocessing irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Brambilla, G.; Sartorelli, A.

    1982-01-01

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

  4. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

  5. Experimental irradiation of UMo fuel: Pie results and modeling of fuel behaviour

    International Nuclear Information System (INIS)

    Languille, A.; Plancq, D.; Huet, F.; Guigon, B.; Lemoine, P.; Sacristan, P.; Hofman, G.; Snelgrove, J.; Rest, J.; Hayes, S.; Meyer, M.; Vacelet, H.; Leborgne, E.; Dassel, G.

    2002-01-01

    Seven full-sized U Mo plates containing ca. 8 g/cm 3 of uranium in the fuel meat have been irradiated since the beginning of the French U Mo development program. The first three of them with 20% 235 U enrichment were irradiated at maximum surfacic power under 150 W/cm 2 in the OSIRIS reactor up to 50% burn-up and are under examination. Their global behaviour is satisfactory: no failure and a low swelling. The other four plates were irradiated in the HFR Petten at maximum surfacic power between 150 and 250 W/cm 2 with two enrichments 20 and 35%. The experiment was stopped after two cycles due to a fuel failure. The post- irradiation examinations were completed in 2001 in Petten. Examinations showed a correct behaviour of 20% enriched plates and an abnormal behaviour of the two other plates (35%-enriched) with a clad failure on the plate 4. The fuel failure appears to result from a combination of factors that led to high corrosion cladding and high fuel meat temperatures. (author)

  6. The post-irradiated examination of CANDU type fuel irradiated in the Institute for Nuclear Research TRIGA reactor

    International Nuclear Information System (INIS)

    Tuturici, I.L.; Parvan, M.; Dobrin, R.; Popov, M.; Radulescu, R.; Toma, V.

    1995-01-01

    This post-irradiation examination work has been done under the Research Contract No. 7756/RB, concluded between the International Atomic Energy Agency and the Institute for Nuclear Research. The paper contains a general description of the INR post-irradiation facility and methods and the relevant post-irradiation examination results obtained from an irradiated experimental CANDU type fuel element designed, manufactured and tested by INR in a power ramp test in the 100 kW Pressurised Water Irradiation Loop of the TRIGA 14 MW(th) Reactor. The irradiation experiment consisted in testing an assembly of six fuel elements, designed to reach a bumup of ∼ 200 MWh/kgU, with typical CANDU linear power and ramp rate. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  8. Device for taking gaseous samples from irradiated fuel elements

    International Nuclear Information System (INIS)

    Lengacker, B.

    1983-01-01

    The described device allows to take gaseous samples from irradiated fuel elements. It is connected with a gas analyzer and a pressure gage, so that in opening the fuel can the internal pressure can be determined

  9. SP-100 Fuel Pin Performance: Results from Irradiation Testing

    Science.gov (United States)

    Makenas, Bruce J.; Paxton, Dean M.; Vaidyanathan, Swaminathan; Marietta, Martin; Hoth, Carl W.

    1994-07-01

    A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pins are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

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

    International Nuclear Information System (INIS)

    Barner, J.O.

    1979-04-01

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

  11. 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)

  12. Modeling of coated fuel particles irradiation behavior

    International Nuclear Information System (INIS)

    Liang Tongxiang; Phelip, M.

    2006-01-01

    In this report, PANAMA code was used to estimate the CP performance under normal and accident condition. Under the normal irradiation test (1000 degree C 625 efpd, 10% FIMA), for intact CP fuel, failure fraction is in the level of 10 -7 . As-fabricated SiC failed particles results in the through coatings failed particles much earlier than the intact particles does, OPyC layer does not fail immediately after irradiation starts. The significant failures start at beyond the burnup of about 7% FIMA. Under the accident condition, the calculated results showed that when the heating temperature is much higher than 1850 degree C, the failure fraction of coated particle can reach the level of 1 percent. The CP fuel fails significantly if it has a buffer layer thinner than 65 urn, SiC layer thinner than 30 μm. High burnup CP need to develop small size kernel, thick buffer layer and thick SiC layer. (authors)

  13. Overview of MIT, ADIP irradiation experiments

    International Nuclear Information System (INIS)

    Kohse, G.; Harling, O.K.; Grant, N.J.

    1985-06-01

    Various rapidly solidified austenitic, ferritic and copper alloys have been produced at MIT for inclusion in ADIP neutron irradiation experiments. A brief summary of the alloys and their preparation and the achieved or projected irradiation parameters is provided

  14. The Analysis Of Spent Fuel Utilization For A Gamma Irradiator

    International Nuclear Information System (INIS)

    MS, Pudjijanto; Setiyanto

    2002-01-01

    The gamma irradiator using RSG-GAS spent fuels was analyzed. The cylindrical geometry of the irradiator was designed by locating the spent fuels the cylindrical periphery. The analysis was focused to evaluate the feasibilities of the irradiator as a fruits and vegetables irradiator. The spent fuels activities were calculated using Origen2 code, while the dose rate at the irradiation positions was determined by linear attenuation model with transport coefficient. The evaluated results showed that the cylindrical geometry of irradiators with diameter around 1-1.5 m gave the effective dose rate for fruits and vegetables preservation. It can be concluded that one can use the RSG-GAS spent fuels effectively as a gamma irradiator for certain applications

  15. Report of Post Irradiation Examination for Dry Process Fuel

    International Nuclear Information System (INIS)

    Par, Jang Jin; Jung, I. H.; Kang, K. H.; Moon, J. S.; Lee, C. R.; Ryu, H. J.; Song, K. C.; Yang, M. S.; Yoo, B. O.; Jung, Y. H.; Choo, Y. S.

    2006-08-01

    The spent PWR fuel typically contains 0.9 wt.% of fissile uranium and 0.6 wt.% of fissile plutonium, which exceeds the natural uranium fissile content of 0.711 wt.%. The neutron economy of a CANDU reactor is sufficient to utilize the DUPIC fuel, even though the neutron-absorbing fission products contained in the spent PWR fuel were remained in the DUPIC fuel. The DUPIC fuel cycle offers advantages to the countries operating both the PWR and CANDU reactors, such as saving the natural uranium, reducing the spent fuel in both PWR and CANDU, and acquiring the extra energy by reuse of the PWR spent fuel. This report contains the results of post-irradiation examination of the DUPIC fuel irradiated four times at HANARO from May 2000 to August 2006 present except the first irradiation test of simulated DUPIC fuel at HANARO on August 1999

  16. Schedule and status of irradiation experiments

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.; Grossbeck, M.L.; Robertson, J.P.

    1998-01-01

    The current status of reactor irradiation experiments is presented in tables summarizing the experimental objectives, conditions, and schedule. Currently, the program has four irradiation experiments in reactor, and five experiments in the design or construction stages. Postirradiation examination and testing is in progress on ten experiments

  17. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    International Nuclear Information System (INIS)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-01-01

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

  18. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Hunn, John D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Ploger, Scott A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Morris, Robert N.; Baldwin, Charles A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Harp, Jason M.; Winston, Philip L. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Gerczak, Tyler J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Rooyen, Isabella J. van [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Montgomery, Fred C.; Silva, Chinthaka M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States)

    2016-09-15

    Highlights: • Post-irradiation examination was performed on AGR-1 coated particle fuel. • Cesium release from the particles was very low in the absence of failed SiC layers. • Silver release was often substantial, and varied considerably with temperature. • Buffer and IPyC layers were found to play a key role in TRISO coating behavior. • Fission products palladium and silver were found in the SiC layer of particles. - Abstract: The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of {sup 110m}Ag from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10{sup −4} to 5 × 10{sup −4} for {sup 154}Eu and 8 × 10{sup −7} to 3 × 10{sup −5} for {sup 90}Sr. The average {sup 134}Cs fractional release from compacts was <3 × 10{sup −6} when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10{sup 5} in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving {sup 134}Cs fractional release in two capsules to approximately 10{sup −5}. Identification and characterization of these particles has provided unprecedented insight into

  19. Studies of irradiated zircaloy fuel sheathing using XPS

    International Nuclear Information System (INIS)

    Chan, P.K.; Irving, K.G.; Hocking, W.H.; Duclos, A.M.; Gerwing, A.F.

    1995-01-01

    The preliminary results reported here support the hypothesis that CANLUB graphite coating reduces the rate at which oxygen can react with fuel sheathing. X-ray photoelectron spectroscopic (XPS) characterization of Zircaloy sheathing obtained from extended-burnup Bruce-type elements (BDL-406-XY (555 MW.h/kgU) and BDL-406-AAH (731 MW.h/kgU)) irradiated in NRU indicates that CANLUB may reduce fuel sheath oxidation, and hence that fission-liberated oxygen may remain in the fuel. Chemical shifts in the Zr 3d spectra suggest that a stoichiometric (ZrO 2 ) oxide film was formed only on Zircaloy in direct contact with fuel. Particulate fuel adhering to the sheath was also determined to be systematically more oxidized on surfaces with CANLUB than on those without it. The unique association of tin on sheathing specimens with the non-CANLUB-coated specimens might also suggest that the tin had segregated from the sheathing. It must be emphasized that further experiments are required to better define the effect of CANLUB on fuel oxidation. (author). 14 refs., 1 tab., 3 figs

  20. Studies of irradiated zircaloy fuel sheathing using XPS

    Energy Technology Data Exchange (ETDEWEB)

    Chan, P K; Irving, K G [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Hocking, W H; Duclos, A M; Gerwing, A F [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-31

    The preliminary results reported here support the hypothesis that CANLUB graphite coating reduces the rate at which oxygen can react with fuel sheathing. X-ray photoelectron spectroscopic (XPS) characterization of Zircaloy sheathing obtained from extended-burnup Bruce-type elements (BDL-406-XY (555 MW.h/kgU) and BDL-406-AAH (731 MW.h/kgU)) irradiated in NRU indicates that CANLUB may reduce fuel sheath oxidation, and hence that fission-liberated oxygen may remain in the fuel. Chemical shifts in the Zr 3d spectra suggest that a stoichiometric (ZrO{sub 2}) oxide film was formed only on Zircaloy in direct contact with fuel. Particulate fuel adhering to the sheath was also determined to be systematically more oxidized on surfaces with CANLUB than on those without it. The unique association of tin on sheathing specimens with the non-CANLUB-coated specimens might also suggest that the tin had segregated from the sheathing. It must be emphasized that further experiments are required to better define the effect of CANLUB on fuel oxidation. (author). 14 refs., 1 tab., 3 figs.

  1. Microscopic analysis of irradiated AGR-1 coated particle fuel compacts

    Energy Technology Data Exchange (ETDEWEB)

    Ploger, Scott A., E-mail: scott.ploger@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3855 (United States); Demkowicz, Paul A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3855 (United States); Hunn, John D.; Kehn, Jay S. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States)

    2014-05-01

    The AGR-1 experiment involved irradiation of 72 TRISO-coated particle fuel compacts to a peak compact-average burnup of 19.5% FIMA with no in-pile failures observed out of 3 × 10{sup 5} total particles. Irradiated AGR-1 fuel compacts have been cross-sectioned and analyzed with optical microscopy to characterize kernel, buffer, and coating behavior. Six compacts have been examined, spanning a range of irradiation conditions (burnup, fast fluence, and irradiation temperature) and including all four TRISO coating variations irradiated in the AGR-1 experiment. The cylindrical specimens were sectioned both transversely and longitudinally, then polished to expose from 36 to 79 individual particles near midplane on each mount. The analysis focused primarily on kernel swelling and porosity, buffer densification and fracturing, buffer–IPyC debonding, and fractures in the IPyC and SiC layers. Characteristic morphologies have been identified, 981 particles have been classified, and spatial distributions of particle types have been mapped. No significant spatial patterns were discovered in these cross sections. However, some trends were found between morphological types and certain behavioral aspects. Buffer fractures were found in 23% of the particles, and these fractures often resulted in unconstrained kernel protrusion into the open cavities. Fractured buffers and buffers that stayed bonded to IPyC layers appear related to larger pore size in kernels. Buffer–IPyC interface integrity evidently factored into initiation of rare IPyC fractures. Fractures through part of the SiC layer were found in only four classified particles, all in conjunction with IPyC–SiC debonding. Compiled results suggest that the deliberate coating fabrication variations influenced the frequencies of IPyC fractures and IPyC–SiC debonds.

  2. Quality surveillance experience of PHWR fuel

    International Nuclear Information System (INIS)

    Kulkarni, P.G.; Bandyopadhyay, A.K.; Shah, B.K.

    1997-01-01

    Quality Surveillance activities are being carried out for PHWR fuel for over 25 years in India. A large number of fuel bundles of 19 element design have been produced and successfully irradiated. The quality surveillance practices follow the guidelines given in various Quality Assurance Codes and Guides. An independent third party surveillance is provided to cover major manufacturing and quality control operations. A system of design basis review periodic quality audit and regulatory safety review is in place. Over the years there have been modifications in the quality assurance procedures to comply with changing requirements. Also many innovative improvements have been introduced in the manufacturing procedures. Similarly quality control activities are also modified. Developments in fuel has remained a continuous activity. The paper summarizes the experience gathered over many years in this exciting process of innovation and improvement. (author)

  3. Evaluation of fuel rods behavior - under irradiation test

    International Nuclear Information System (INIS)

    Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

    1981-04-01

    By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

  4. An equipment for the dimensional characterization of irradiated fuel channels

    International Nuclear Information System (INIS)

    Cederquist, H.

    1985-01-01

    The reuse of irradiated fuel channels in BWRs is highly beneficial. However, one prerequisite for reuse of a fuel channel is the detailed knowledge of its dimensions, which are affected by irradiation and pressure drop during operation. Therefore an equipment for fast and accurate dimensional measurement of irradiated fuel channels has been developed. The measurements are carried out when the fuel assembly is supported in the same manner as in the reactor core. The equipment utilizes stationary ultrasonic transducers that measure the fuel channel at a number of predetermined axial levels. Measurement data are fed into a computer which calculates the requested dimensional characteristics such as transversal flatness, bow, twist, side perpendicularity etc. Data are automatically printed for subsequent evaluation. Measurements can be performed both when the fuel channel is placed on a fuel bundle and on an empty fuel channel

  5. Overview of the FUTURIX-FTA Irradiation Experiment in the Phénix Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Heather J.M. Chichester; Steve L. Hayes; Kenneth J. McClellan; Jean-Luc Paul; Marc Masson; Stewart L. Voit; Fabienne Delage

    2015-09-01

    The Advanced Fuels Campaign utilizes the Advanced Test Reactor (ATR) for most of its irradiation testing. Cadmium-shrouded baskets are used in ATR to modify the neutron spectrum to simulate a fast reactor environment for the fuel. FUTURIX-FTA is an irradiation experiment conducted in the Phenix fast reactor in France. Results from FUTURIX-FTA and irradiation tests in ATR using identical fuel compositions will be compared to identify and evaluate any differences in fuel behavior due to differences in the irradiation source.

  6. An investigation on the irradiation behavior of atomized U-Mo/Al dispersion rod fuels

    International Nuclear Information System (INIS)

    Park, J.M.; Ryu, H.J.; Lee, Y.S.; Lee, D.B.; Oh, S.J.; Yoo, B.O.; Jung, Y.H.; Sohn, D.S.; Kim, C.K.

    2005-01-01

    The second irradiation fuel experiment, KOMO-2, for the qualification test of atomized U-Mo dispersion rod fuels with U-loadings of 4-4.5 gU/cc at KAERI was finished after an irradiation up to 70 at% U 235 peak burn-up and subjected to the IMEF (Irradiation material Examination Facility) for a post-irradiation analysis in order to understand the fuel irradiation performance of the U-Mo dispersion fuel. Current results for PIE of KOMO-2 revealed that the U-Mo/Al dispersion fuel rods exhibited a sound performance without any break-away swelling, but most of the fuel rods irradiated at a high linear power showed an extensive formation of the interaction phase between the U-Mo particle and the Al matrix. In this paper, the analysis of the PIE results, which focused on the diffusion related microstructures obtained from the optical and EPMA (Electron Probe Micro Analysis) observations, will be presented in detail. And a thermal modeling will be carried out to calculate the temperature of the fuel rod during an irradiation. (author)

  7. In-pile irradiation of rock-like oxide fuels

    International Nuclear Information System (INIS)

    Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H.

    2001-01-01

    Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO 2 single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

  8. In-pile irradiation of rock-like oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2001-07-01

    Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO{sub 2} single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

  9. Damage and failure of unirradiated and irradiated fuel rods tested under film boiling conditions

    International Nuclear Information System (INIS)

    Mehner, A.S.; Hobbins, R.R.; Seiffert, S.L.; MacDonald, P.E.; McCardell, R.K.

    1979-01-01

    Power-cooling-mismatch experiments are being conducted as part of the Thermal Fuels Behavior Program in the Power Burst Facility at the Idaho National Engineering Laboratory to evaluate the behavior of unirradiated and previously irradiated light water reactor fuel rods tested under stable film boiling conditions. The observed damage that occurs to the fuel rod cladding and the fuel as a result of film boiling operation is reported. Analyses performed as a part of the study on the effects of operating failed fuel rods in film boiling, and rod failure mechanisms due to cladding embrittlement and cladding melting upon being contacted by molten fuel are summarized

  10. Post-irradiation examination of CANDU fuel bundles fuelled with (Th, Pu)O2

    International Nuclear Information System (INIS)

    Karam, M.; Dimayuga, F.C.; Montin, J.

    2010-01-01

    AECL has extensive experience with thoria-based fuel irradiations as part of an ongoing R&D program on thorium within the Advanced Fuel Cycles Program. The BDL-422 experiment was one component of the thorium program that involved the fabrication and irradiation testing of six Bruce-type bundles fuelled with (Th, Pu)O 2 pellets. The fuel was manufactured in the Recycle Fuel Fabrication Laboratories (RFFL) at Chalk River allowing AECL to gain valuable experience in fabrication and handling of thoria fuel. The fuel pellets contained 86.05 wt.% Th and 1.53 wt.% Pu in (Th, Pu)O 2 . The objectives of the BDL-422 experiment were to demonstrate the ability of 37-element geometry (Th, Pu)O 2 fuel bundles to operate to high burnups up to 1000 MWh/kgHE (42 MWd/kgHE), and to examine the (Th, Pu)O 2 fuel performance. This paper describes the post-irradiation examination (PIE) results of BDL-422 fuel bundles irradiated to burnups up to 856 MWh/kgHE (36 MWd/kgHE), with power ratings ranging from 52 to 67 kW/m. PIE results for the high burnup bundles (>1000 MWh/kgHE) are being analyzed and will be reported at a later date. The (Th, Pu)O 2 fuel performance characteristics were superior to UO 2 fuel irradiated under similar conditions. Minimal grain growth was observed and was accompanied by benign fission gas release and sheath strain. Other fuel performance parameters, such as sheath oxidation and hydrogen distribution, are also discussed. (author)

  11. Irradiation behavior of uranium oxide - Aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products and as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show that, with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 g U/cm 3 ) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼63% 235 U burnup). (author)

  12. Irradiation behavior of uranium oxide-aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, G.L.; Rest, J.; Snelgrove, J.L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼ 63% 235 U burnup)

  13. Fuel pins irradiation: experimental devices and analytical behaviour

    International Nuclear Information System (INIS)

    Lemaignan, C.

    1996-01-01

    In this text we present the general characteristics of adapted irradiation loops in research reactors and the main results that we can expected with these loops in the behaviour field of PWR and LMFBR fuels( fuel densification, fuel cladding interactions, fission products release, reactor accidents)

  14. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    Markgraf, J.F.W.

    1985-01-01

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  15. A general evaluation of the irradiation behaviour of dispersion fuels

    International Nuclear Information System (INIS)

    Hofman, G.L.

    1995-01-01

    The irradiation behaviour of aluminum-based dispersion fuels is evaluated with emphasis on metallurgical processes that control the dispersion behaviour. Phase transformations and microstructural changes resulting from fuel-matrix interactions and the effect of fissioning in fuel are discussed. (author)

  16. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul Demkowicz; Scott Ploger; John Hunn

    2012-05-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Five irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These five compacts also included all four TRISO coating variations irradiated in the AGR experiment. The five compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. Approximately 40 to 80 particles within each cross section were exposed near enough to mid-plane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 830 classified particles allowed other relationships among morphological types to be established.

  17. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    International Nuclear Information System (INIS)

    Demkowicz, Paul; Ploger, Scott; Hunn, John

    2012-01-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Five irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These five compacts also included all four TRISO coating variations irradiated in the AGR experiment. The five compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. Approximately 40 to 80 particles within each cross section were exposed near enough to mid-plane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 830 classified particles allowed other relationships among morphological types to be established.

  18. First qualitative analysis of fuel irradiation results carried out in the MR reactor on WWER-1000 fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chantoin, P [International Atomic Energy Agency, Vienna (Austria); Dubrovin, K; Platonov, P [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Onufriev, V [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

    1994-12-31

    Four experiments carried out in the MR reactor are evaluated. They are aimed to assess the influence of burnup and the size of the pellet central hole on the fuel temperature and thus on the fuel swelling and fission gas release. The experiments have been performed at different linear rate and burnup of the fuel rods which are above the actual licensed values in WWER power stations. In this paper the results on WWER fuel rod behaviour are examined. The main fabrication and irradiation characteristics for each experiment are given. The main results from destructive and non-destructive examinations are summarized. They include: burnup determination by gamma spectroscopy, caesium shifting along fuel column and accumulation at the end of the fuel stack, fission gas release. fuel rod diameter and length change and macro-graphs showing the central hole size and the morphology after irradiation. From observation of fuel structure, Cs spectrometry and fission gas release, a large degradation of fuel thermal conductivity can be identified at high burnup. If the fuel burnup is the right parameter to be considered, burnup limits identified are: 0 70-75 MWd/kg for rods with large central hole; (2) 58-64 MWd/kg for rods with small central hole. As a general conclusion it is stressed the importance of the study due to irradiation beyond the usual linear rates at high burnup. Up to now the fuel life limiting factor was cladding corrosion when using Zircaloy-4. As the cladding corrosion situation improves, the next life limiting factor to be met could be the fuel itself. The decreasing fuel thermal conductivity is probably of prime importance and should be further studied and modelled. 5 tabs., 5 figs., 3 refs.

  19. Irradiation performance of U-Mo monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, M. K.; Gan, J.; Jue, J. F.; Keiser, D. D.; Perez, E.; Robinson, A.; Wachs, D. M.; Woolstenhulme, N. [Idaho National Laboratory, Idaho (Korea, Republic of); Kim, Y.S.; Hofman, G. L. [Argonne National Laboratory, Lemont (United States)

    2014-04-15

    High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  20. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    Directory of Open Access Journals (Sweden)

    M.K. MEYER

    2014-04-01

    Full Text Available High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  1. Irradiation of Argentine MOX fuels: Post-irradiation results and analysis

    International Nuclear Information System (INIS)

    Marino, A.C.; Perez, E.; Adelfang, P.

    1997-01-01

    The irradiation of the first Argentine prototypes of PHWR MOX fuels began in 1986. These experiments were made in the HFR-Petten reactor, Holland. The rods were prepared and controlled in the CNEA's facility. The postirradiation examinations were performed in the Kernforschungszentrum, Karlsruhe, Germany and in the JRC, Petten. The first rod has been used for destructive pre-irradiation analysis. The second one as a pathfinder to adjust systems in the HFR. Two additional rods including iodine doped pellets were intended to simulate 15000 MWd/T(M) burnup. The remaining two rods were irradiated until 15000 MWd/T(M) (BU15 experiment). One of them underwent a final ramp with the aim of verifying fabrication processes and studying the behaviour under power transients. BACO code was used to define the power histories and to analyze the experiments. This paper presents the postirradiation examinations for the BU15 experiments and a comparison with the BACO outputs for the rod that presented a failure during the ramp test of the BU15 experiment. (author). 17 refs, 30 figs, 5 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  4. Evaluation of burnup characteristics and energy deposition during NSRR pulse irradiation tests on irradiated BWR fuels

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio

    2000-11-01

    Pulse irradiation tests of irradiated fuel are performed in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under Reactivity Initiated Accident Conditions (RIA). The severity of the RIA is represented by energy deposition or peak fuel enthalpy during the power excursion. In case of the irradiated fuel tests, the energy deposition varies depending both on the amounts and distribution of residual fissile and neutron absorbing fission products generated during the base irradiation. Thus, proper fuel burnup characterization, especially for low enriched commercial fuels, is important, because plutonium (Pu) takes a large part of fissile and its generation depends on the neutron spectrum during the base irradiation. Fuel burnup calculations were conducted with ORIGEN2, RODBURN and SWAT codes for the BWR fuels tested in the NSRR. The calculation results were compared with the measured isotope concentrations and used for the NSRR neutron calculations to evaluate energy depositions of the test fuel. The comparison of the code calculations and the measurements revealed that the neutron spectrum change due to difference in void fraction altered Pu generation and energy deposition in the NSRR tests considerably. With the properly evaluated neutron spectrum, the combined burnup and NSRR neutron calculation gave reasonably good evaluation of the energy deposition. The calculations provided radial distributions of the fission product accumulation during the base irradiation and power distribution during the NSRR pulse irradiation, which were important for the evaluation of both burnup characteristics and fission gas release behavior. (author)

  5. Apparatus for inspecting a irradiated nuclear fuel rod

    International Nuclear Information System (INIS)

    Saura, Hideaki; Yonemura, Eizo.

    1975-01-01

    Object: To increase safety and inspection efficiency by operating irradiated fuel rods, which are accommodated in a water-filled pool after being taken out from the reactor. Structure: When making inspection of irradiated fuel rods, particularly the cladding tube thereof, a fuel box which stores irradiated fuel rods in a water pool is secured to a securement mechanism with slime removal apparatus and inspection apparatus on either side capable of being vertically moved, and it is then stopped at a water depth of about 2 meters. When the lid of the box is opened, irradiated fuel rods are taken out with gripping means and then secured together with the gripping means to an operation base provided on the outside of the pool. Thereafter, the box is lowered by operating pedals on the operation base to completely pull out the irradiated fuel rods from the box, and the irradiated fuel rods are then horizontally moved and then held in a suspended state. Next a slime removal apparatus in raised by operating pedals and an inspection element assembly are progressively raised for inspection of the state of the cladding tube of each fuel rod after removal of slime therefrom. (Nakamura, S.)

  6. Irradiation test of fuel containing minor actinides in the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Soga, Tomonori; Sekine, Takashi; Wootan, David; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

    2007-01-01

    The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP, accounting for both prompt and delayed heating components, and then adjusted using E/C for 10 B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO 2-x or AmO 2-x in the (U, Pu)O 2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel. (author)

  7. A disposal centre for irradiated nuclear fuel: conceptual design study

    International Nuclear Information System (INIS)

    1980-09-01

    This report describes a conceptual design of a disposal centre for irradiated nuclear fuel. The surface facilities consist of plants for the preparation of steel cylinders containing irradiated nuclear fuel immobilized in lead, shaft headframe buildings, and all necessary support facilities. The undergound disposal vault is located on one level at a depth of 1000 metres. The cylinders containing the irradiated fuel are emplaced on a one-metre thick layer of backfill material and then completely covered with backfill. All surface and subsurface facilities are described, operations and schedules are summarized, and cost estimates and manpower requirements are given. (auth)

  8. Gamma-ray spectroscopy on irradiated fuel rods

    International Nuclear Information System (INIS)

    Terremoto, Luis Antonio Albiac

    2009-01-01

    The recording of gamma-ray spectra along an irradiated fuel rod allows the fission products to be qualitatively and quantitatively examined. Among all nondestructive examinations performed on irradiated fuel rods by gamma-ray spectroscopy, the most comprehensive one is the average burnup measurement, which is quantitative. Moreover, burnup measurements by means of gamma-ray spectroscopy are less time-consuming and waste-generating than burnup measurements by radiochemical, destructive methods. This work presents the theoretical foundations and experimental techniques necessary to measure, using nondestructive gamma-ray spectroscopy, the average burnup of irradiated fuel rods in a laboratory equipped with hot cells. (author)

  9. Dry storage of irradiated nuclear fuels and vitrified wastes

    International Nuclear Information System (INIS)

    Deacon, D.

    1982-01-01

    A review is given of the work of GEC Energy Systems Ltd. over the years in the dry storage of irradiated fuel. The dry-storage module (designated as Cell 4) for irradiated magnox fuel recently constructed at Wylfa nuclear power station is described. Development work on the long-term dry storage of irradiated oxide fuels is reported. Four different methods of storage are compared. These are the pond, vault, cask and caisson stores. It is concluded that there are important advantages with the passive air-cooled ESL dry stove. (U.K.)

  10. MOX fuel development: Experience in Argentina

    International Nuclear Information System (INIS)

    Marchi, D.E.; Adelfang, P.; Menghini, J.E.

    1999-01-01

    Since 1973, when a laboratory conceived for the safe manipulation of a few hundred grams of plutonium was built, the CNEA (Argentinean Atomic Energy Commission) has been involved in the small-scale development of MOX fuel technology. The plutonium laboratory consists in a glove box facility (α Facility) featuring the necessary equipment to prepare MOX fuel rods for experimental irradiations and to carry out studies on preparative processes development and chemical and physical characterization. The irradiation of the first prototypes of (U,Pu)O 2 fuels fabricated in Argentina began in 1986. These experiments were carried out in the HFR (High Flux Reactor)- Petten , Holland. The rods were prepared and controlled in the CNEA's a Facility. The post-irradiation examinations (PIE) were performed in the KFK (Kernforschungszentrum Karlsruhe), Germany and the JRC (Joint Research Center), Petten. In the period 1991-1995, the development of new laboratory methods of co-conversion of uranium and plutonium were carried out: reverse strike co-precipitation of ADU-Pu(OH) 4 and direct denitration using microwaves. The reverse strike process produced pellets with a high sintered density, excellent micro-homogeneity and good solubility in nitric acid. Liquid wastes showed a very low content of actinides and the process is easy to operate in a glove box environment. The microwave direct denitration was optimized with uranium alone and the conditions to obtain high density pellets, with a good microstructure, without using a milling step, have been developed. At present, new experiments are being carried out to improve the reverse strike co-precipitation process and direct microwave denitration. A new glove box is being installed at the plutonium laboratory, this glove box has process equipment designed to recover scrap from previous fabrication campaigns, and to co-convert mixed U-Pu solutions by direct microwave denitration. (author)

  11. Fission product released experiment of coated fuel particles

    Energy Technology Data Exchange (ETDEWEB)

    Shijiang, Xu; Bing, Yang; Chunhe, Tang; Junguo, Zhu; Jintao, Huang; Binzhong, Zhang [Inst. of Nucl. Energy Technology, Tsinghua Univ., Beijing (China); Jinghan, Luo [Inst. of Atomic Energy, Beijing (China)

    1992-01-15

    Four samples of coated fuel particles were irradiated in the Heavy-Water Research Reactor of the Institute of Atomic Energy. Each of them was divided into two groups and irradiated to the burn up of 0.394% fima and 0.788% fima in two static capsules, respectively. After irradiation and cooling, post irradiation annealing experiment was carried out, the release ratios of the fission product {sup 133}Xe and {sup 131}I were measured, they are in the order of 10{sup -6}{approx}10{sup -7}. The fission product release ratio of naked kernel was also measured under the same conditions as for the coated fuel particles, the ratio of the fission product release of the coated fuel particles and of the naked kernel was in the order of 10{sup -5}{approx}10{sup -4}.

  12. Status of Cea-Minatom collaborative experiment Bora-Bora: fuels with high plutonium content

    Energy Technology Data Exchange (ETDEWEB)

    Zaboudko, L.; Kurina, I. [Institute for Physics and Power Engineering, IPPE, Kaluga region (Russian Federation); Mayorshin, A.; Kisly, V. [Research Institute for Atomic Reactors, RIAR, Uljanovsk region (Russian Federation); Menshikova, T.; Rogozkin, B. [All Russia Research Institute of Inorganic Materials, VNIINM, Moscow (Russian Federation); Pillon-Eymard, S.; Languille, A. [CEA Cadarache, Dir. de l' Energie Nucleaire, 13 - Saint Paul lez Durance (France); Thetford, R. [AEA Technology, Harwell (United Kingdom)

    2001-07-01

    The program of the experiment on the BOR-60 reactor with different fuel types is presented. It includes 3 phases: fuel fabrication, fuel irradiation on BOR-60, post-irradiation examination. The fuels studied are: MOX fuel UPu{sub 0.45}O{sub 2} made by two different techniques, nitride fuel (UPu{sub 0.45}N and UPu{sub 0.6}N) and inert-matrix fuel (PuN + ZrN and PuO{sub 2} + MgO). Results on fuel fabrication are presented. Description of the irradiation facility is given. Fuel irradiation conditions are discussed. Results of nitride fuel performance calculations by NITRAF code are shown. (author)

  13. Study on the behavior of irradiated light water reactor fuel during out-of-pile annealing

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Kanazawa, Hiroyuki; Uno, Hisao; Sasajima, Hideo

    1988-11-01

    Using the pre-irradiated light water reactor fuel (burnup: 35 MWd/kgU) and the slightly irradiated NSRR fuel (burnup: 5.6 x 10 -6 MWd/kgU), FP gas release rate up to the temperature of 2273 K was measured through out-of-pile annealing test. Results of this experiment were compared with those of ORNL annealing test (SFD/HI-test series) performed in USA. Obtained conclusions are: (1) Maximum release rate of Kr gas in light water reactor fuel was 6.4 % min -1 at temperature of 2273 K. This was in good agreement with ORNL data. FP gas release rate during annealing test was increased greatly with increasing fuel burnup and annealing temperature. (2) No FP was detected in NSRR slightly irradiated fuel up to the temperature of 1913 K. (author)

  14. Review of thorium fuel reprocessing experience

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  15. Alternative Aviation Fuel Experiment (AAFEX)

    Science.gov (United States)

    Anderson, B. E.; Beyersdorf, A. J.; Hudgins, C. H.; Plant, J. V.; Thornhill, K. L.; Winstead, E. L.; Ziemba, L. D.; Howard, R.; Corporan, E.; Miake-Lye, R. C.; hide

    2011-01-01

    The rising cost of oil coupled with the need to reduce pollution and dependence on foreign suppliers has spurred great interest and activity in developing alternative aviation fuels. Although a variety of fuels have been produced that have similar properties to standard Jet A, detailed studies are required to ascertain the exact impacts of the fuels on engine operation and exhaust composition. In response to this need, NASA acquired and burned a variety of alternative aviation fuel mixtures in the Dryden Flight Research Center DC-8 to assess changes in the aircraft s CFM-56 engine performance and emission parameters relative to operation with standard JP-8. This Alternative Aviation Fuel Experiment, or AAFEX, was conducted at NASA Dryden s Aircraft Operations Facility (DAOF) in Palmdale, California, from January 19 to February 3, 2009 and specifically sought to establish fuel matrix effects on: 1) engine and exhaust gas temperatures and compressor speeds; 2) engine and auxiliary power unit (APU) gas phase and particle emissions and characteristics; and 3) volatile aerosol formation in aging exhaust plumes

  16. Design of a transportation cask for irradiated CANDU fuel

    International Nuclear Information System (INIS)

    Nash, K.E.; Gavin, M.E.

    1983-01-01

    A major step in the development of a large-scale transportation system for irradiated CANDU fuel is being made by Ontario Hydro in the design and construction of a demonstration cask by 1988/89. The system being designed is based on dry transportation with the eventual fully developed system providing for dry fuel loading and unloading. Research carried out to date has demonstrated that it is possible to transport irradiated CANDU fuel in a operationally efficient and simple manner without any damage which would prejudice subsequent automated fuel handling

  17. Irradiation of inert matrix and mixed oxide fuel in the Halden test reactor

    International Nuclear Information System (INIS)

    Hellwig, Ch.; Kasemeyer, U.

    2001-01-01

    In a new type of fuel, called Inert Matrix Fuel (IMF), plutonium is embedded in a U-free matrix. This offers advantages for more efficient plutonium consumption, higher proliferation resistance, and for inert behaviour later in a waste repository. In the fuel type investigated at PSI, plutonium is dissolved in yttrium-stabilized zirconium oxide (YSZ), a highly radiation-resistant cubic phase, with addition of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ-based IMF is ongoing in the OECD Material Test Reactor in Halden (HBWR), together with MOX fuel (Rig IFA-651.1). The experiment is described herein and results are presented of the first 120 days of irradiation with an average assembly burnup of 47 kWd/cm 3 . The results are compared with neutronic calculations performed before the experiment, and are used to model the fuel behaviour with the PSI-modified TRANSURANUS code. The measured fuel temperatures are within the expected range. An unexpectedly strong densification of the IMF during the first irradiation cycle does not alter the fuel temperatures. An explanation for this behaviour is proposed. The irradiation at higher linear heat rates during forthcoming cycles will deliver information about the fission gas release behaviour of the IMF. (author)

  18. Irradiation of inert matrix and mixed oxide fuel in the Halden test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hellwig, Ch.; Kasemeyer, U

    2001-03-01

    In a new type of fuel, called Inert Matrix Fuel (IMF), plutonium is embedded in a U-free matrix. This offers advantages for more efficient plutonium consumption, higher proliferation resistance, and for inert behaviour later in a waste repository. In the fuel type investigated at PSI, plutonium is dissolved in yttrium-stabilized zirconium oxide (YSZ), a highly radiation-resistant cubic phase, with addition of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ-based IMF is ongoing in the OECD Material Test Reactor in Halden (HBWR), together with MOX fuel (Rig IFA-651.1). The experiment is described herein and results are presented of the first 120 days of irradiation with an average assembly burnup of 47 kWd/cm{sup 3}. The results are compared with neutronic calculations performed before the experiment, and are used to model the fuel behaviour with the PSI-modified TRANSURANUS code. The measured fuel temperatures are within the expected range. An unexpectedly strong densification of the IMF during the first irradiation cycle does not alter the fuel temperatures. An explanation for this behaviour is proposed. The irradiation at higher linear heat rates during forthcoming cycles will deliver information about the fission gas release behaviour of the IMF. (author)

  19. Examination of irradiated fuel elements using gamma scanning technique

    International Nuclear Information System (INIS)

    Ichim, O.; Mincu, M.; Man, I.; Stanica, M.

    2016-01-01

    The purpose of this paper is to validate the gamma scanning technique used to calculate the activity of gamma fission products from CANDU/TRIGA irradiated fuel elements. After a short presentation of the equipments used and their characteristics, the paper describes the calibration technique for the devices and how computed tomography reconstruction is done. Following the previously mentioned steps is possible to obtain the axial and radial profiles and the computed tomography reconstruction for calibration sources and for the irradiated fuel elements. The results are used to validate the gamma scanning techniques as a non-destructive examination method. The gamma scanning techniques will be used to: identify the fission products in the irradiated CANDU/TRIGA fuel elements, construct the axial and radial distributions of fission products, get the distribution in cross section through computed tomography reconstruction, and determine the nuclei number and the fission products activity of the irradiated CANDU/TRIGA fuel elements. (authors)

  20. Some results on development, irradiation and post-irradiation examinations of fuels for fast reactor-actinide burner (MOX and inert matrix fuel)

    International Nuclear Information System (INIS)

    Poplavsky, V.; Zabudko, L.; Moseev, L.; Rogozkin, B.; Kurina, I.

    1996-01-01

    Studies performed have shown principal feasibility of the BN-600 and BN-800 cores to achieve high efficiency of Pu burning when MOX fuel with Pu content up to 45% is used. Valuable experience on irradiation behaviour of oxide fuel with high Pu content (100%) was gained as a result of operation of two BR-10 core loadings where the maximum burnup 14 at.% was reached. Post-irradiation examination (PIE) allowed to reveal some specific features of the fuel with high plutonium content. Principal irradiation and PIE results are presented in the paper. Use of new fuel without U-238 provides the maximum burning capability as in this case the conversion ratio is reduced to zero. Technological investigations of inert matrix fuels have been continued now. Zirconium carbide, zirconium nitride, magnesium oxide and other matrix materials are under consideration. Inert matrices selection criteria are discussed in the paper. Results of technological study, of irradiation in the BOR-60 reactor and PIE results of some inert matrix fuels are summarized in this report. (author). 2 refs, 1 fig., 3 tabs

  1. Aluminum cladding oxidation of prefilmed in-pile fueled experiments

    Energy Technology Data Exchange (ETDEWEB)

    Marcum, W.R., E-mail: marcumw@engr.orst.edu [Oregon State University, School of Nuclear Science and Engineering, 116 Radiation Center, Corvallis, OR 97331 (United States); Wachs, D.M.; Robinson, A.B.; Lillo, M.A. [Idaho National Laboratory, Nuclear Fuels & Materials Department, 2525 Fremont Ave., Idaho Falls, ID 83415 (United States)

    2016-04-01

    A series of fueled irradiation experiments were recently completed within the Advanced Test Reactor Full size plate In center flux trap Position (AFIP) and Gas Test Loop (GTL) campaigns. The conduct of the AFIP experiments supports ongoing efforts within the global threat reduction initiative (GTRI) to qualify a new ultra-high loading density low enriched uranium-molybdenum fuel. This study details the characterization of oxide growth on the fueled AFIP experiments and cross-correlates the empirically measured oxide thickness values to existing oxide growth correlations and convective heat transfer correlations that have traditionally been utilized for such an application. This study adds new and valuable empirical data to the scientific community with respect to oxide growth measurements of highly irradiated experiments, of which there is presently very limited data. Additionally, the predicted oxide thickness values are reconstructed to produce an oxide thickness distribution across the length of each fueled experiment (a new application and presentation of information that has not previously been obtainable in open literature); the predicted distributions are compared against experimental data and in general agree well with the exception of select outliers. - Highlights: • New experimental data is presented on oxide layer thickness of irradiated aluminum fuel. • Five oxide growth correlations and four convective heat transfer correlations are used to compute the oxide layer thickness. • The oxide layer thickness distribution is predicted via correlation for each respective experiment. • The measured experiment and predicted distributions correlate well, with few outliers.

  2. Development status of irradiation devices and instrumentation for material and nuclear fuel irradiation tests in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, Jae Min; Choo, Kee Nam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-04-15

    The High flux Advanced Neutron Application ReactOr (HANARO), an open-tank-in-pool type reactor, is one of the multi-purpose research reactors in the world. Since the commencement of HANARO's operations in 1995, a significant number of experimental facilities have been developed and installed at HANARO, and continued efforts to develop more facilities are in progress. Owing to the stable operation of the reactor and its frequent utilization, more experimental facilities are being continuously added to satisfy various fields of study and diverse applications. The irradiation testing equipment for nuclear fuels and materials at HANARO can be classified into capsules and the Fuel Test Loop (FTL). Capsules for irradiation tests of nuclear fuels in HANARO have been developed for use under the dry conditions of the coolant and materials at HANARO and are now successfully utilized to perform irradiation tests. The FTL can be used to conduct irradiation testing of a nuclear fuel under the operating conditions of commercial nuclear power plants. During irradiation tests conducted using these capsules in HANARO, instruments such as the thermocouple, Linear Variable Differential Transformer (LVDT), small heater, Fluence Monitor (F/M) and Self-Powered Neutron Detector (SPND) are used to measure various characteristics of the nuclear fuel and irradiated material. This paper describes not only the status of HANARO and the status and perspective of irradiation devices and instrumentation for carrying out nuclear fuel and material tests in HANARO but also some results from instrumentation during irradiation tests

  3. Thermal conductivity of fresh and irradiated U-Mo fuels

    Science.gov (United States)

    Huber, Tanja K.; Breitkreutz, Harald; Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Elgeti, Stefan; Reiter, Christian; Robinson, Adam. B.; Smith, Frances. N.; Wachs, Daniel. M.; Petry, Winfried

    2018-05-01

    The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, thermal conductivity of fresh dispersion fuel at a temperature of 150 °C decreased from 59 W/m·K to 18 W/m·K at a burn-up of 4.9·1021 f/cc and further to 9 W/m·K at a burn-up of 6.1·1021 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep than for dispersion fuel. For a burn-up of 3.5·1021 f/cc of monolithic fuel, a thermal conductivity of 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. (2015). The difference of decrease for both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increased burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice also affect both dispersion and monolithic fuel.

  4. Metallographic analysis of irradiated RERTR-3 fuel test specimens

    International Nuclear Information System (INIS)

    Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

    2000-01-01

    The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date

  5. 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.)

  6. A Retailer's Experience with Irradiated Foods

    International Nuclear Information System (INIS)

    James P. Corrigan

    2000-01-01

    A food irradiation success story comes from Northbrook, Illinois, where Carrot Top, Inc., has been routinely carrying irradiated food for more than 7 yr. This paper presents the experiences of Carrot Top during those years, details the marketing approaches used, and summarizes the resulting sales figures

  7. TEM investigation of irradiated U-7 weight percent Mo dispersion fuel

    International Nuclear Information System (INIS)

    Van den Berghe, S.

    2009-01-01

    In the FUTURE experiment, fuel plates containing U-7 weight percent Mo atomized powder were irradiated in the BR2 reactor. At a burn-up of approximately 33 percent 235 U (6.5 percent FIMA or 1.41 10 21 fissions/cm 3 meat), the fuel plates showed an important deformation and the irradiation was stopped. The plates were submitted to detailed PIE at the Laboratory for High and Medium level Activity. The results of these examinations were reported in the scientific report of last year and published in open literature. Since then, the microstructural aspects of the FUTURE fuel were studied in more detail using transmission electron microscopy (TEM), in an attempt to understand the nature of the interaction phase and the fission gas behavior in the atomized U(Mo) fuel. The FUTURE experiment is regarded as the definitive proof that the classical atomized U(Mo) dispersion fuel is not stable under irradiation, at least in the conditions required for normal operation of plate-type fuel. The main cause for the instability was identified to be the irradiation behavior of the U(Mo)-Al interaction phase which is formed between the U(Mo) particles and the pure aluminum matrix during irradiation. It is assumed to become amorphous under irradiation and as such cannot retain the fission gas in stable bubbles. As a consequence, gas filled voids are generated between the interaction layer and the matrix, resulting in fuel plate pillowing and failure. The objective of the TEM investigation was the confirmation of this assumption of the amorphisation of the interaction phase. A deeper understanding of the actual nature of this layer and the fission gas behaviour in these fuels in general can allow a more oriented search for a solution to the fuel failures

  8. Visual in-pile fuel disruption experiments

    International Nuclear Information System (INIS)

    Cano, G.L.; Ostensen, R.W.; Young, M.F.

    1978-01-01

    In a loss-of-flow (LOF) accident in an LMFBR, the mode of disruption of fuel may determine the probability of a subsequent energetic excursion. To investigate these phenomena, in-pile disruption of fission-heated irradiated fuel pellets was recorded by high speed cinematography. Instead of fuel frothing or dust-cloud breakup (as used in the SAS code) massive and very rapid fuel swelling, not predicted by analytical models, occurred. These tests support massive fuel swelling as the initial mode of fuel disruption in a LOF accident. (author)

  9. Indian experience in fuel reprocessing

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  10. Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

    1994-10-01

    This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result's from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of ∼75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of ∼1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of ∼11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  12. Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

    International Nuclear Information System (INIS)

    Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

    2011-01-01

    The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

  13. Some general requirements for irradiation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Myers, H P; Skjoeldebrand, R

    1960-05-15

    This paper is limited to the interests of the EAES-symposium, namely the use of reactors for materials research and testing, for convenience we exclude consideration of chemical effects and problems of coolant technology. Its purpose is to try to define the general requirements for irradiation experiments and the reactors housing them; to see what facilities for irradiation experiments are available within the European Atomic Energy Society countries and finally, to point out possible limitations of these facilities.

  14. Halden fuel and material experiments beyond operational and safety limits

    International Nuclear Information System (INIS)

    Volkov, Boris; Wiesenack, Wolfgang; McGrath, M.; Tverberg, T.

    2014-01-01

    One of the main tasks of any research reactor is to investigate the behavior of nuclear fuel and materials prior to their introduction into the market. For commercial NPPs, it is important both to test nuclear fuels at a fuel burn-up exceeding current limits and to investigate reactor materials for higher irradiation dose. For fuel vendors such tests enable verification of fuel reliability or for the safety limits to be found under different operational conditions and accident situations. For the latter, in-pile experiments have to be performed beyond some normal limits. The program of fuel tests performed in the Halden reactor is aimed mainly at determining: The thermal FGR threshold, which may limit fuel operational power with burn-up increase, the “lift-off effect” when rod internal pressure exceeds coolant pressure, the effects of high burn-up on fuel behavior under power ramps, fuel relocation under LOCA simulation at higher burn-up, the effect of dry-out on high burn-up fuel rod integrity. This paper reviews some of the experiments performed in the Halden reactor for understanding some of the limits for standard fuel utilization with the aim of contributing to the development of innovative fuels and cladding materials that could be used beyond these limits. (author)

  15. New JMTR irradiation test plan on fuels and materials

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Nishiyama, Yutaka; Chimi, Yasuhiro; Sasajima, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Suzuki, Masahide; Kawamura, Hiroshi

    2009-01-01

    In order to maintain and enhance safety of light water reactors (LWRs) in long-term and up-graded operations, proper understanding of irradiation behavior of fuels and materials is essentially important. Japanese government and the Japan Atomic Energy Agency (JAEA) have decided to refurbish the Japan Materials Testing Reactor (JMTR) and to install new tests rigs, in order to play an active role for solving irradiation related issues on plant aging and high-duty uses of the current LWRs and on development of next-generation reactors. New tests on fuel integrity under simulated abnormal transients and high-duty irradiation conditions are planned in the JMTR. Power ramp tests of newdesign fuel rods will also be performed in the first stage of the program, which is expected to start in year 2011 after refurbishment of the JMTR. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor (NSRR) and loss of coolant accident tests in hot laboratories would serve as the integrated fuel safety research on the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients. For the materials irradiation, fracture toughness of reactor vessel steels and stress corrosion cracking behavior of stainless steels are being studied in addition to basic irradiation behavior of nuclear materials such as hafnium. The irradiation studies would contribute not only to solve the current problems but also to identify possible seeds of troubles and to make proactive responses. (author)

  16. Post irradiation examination and experience

    International Nuclear Information System (INIS)

    1985-11-01

    The present meeting was scheduled by the International Atomic Energy Agency upon proposal from the members of the International Working Group on Water Reactor Fuel Performance and Technology. At the invitation of the Government of Japan, Japan Atomic Energy Research Institute and Nuclear Safety Research Association (of Japan), organized the meeting in Tokyo. 37 participants representing 13 countries and one international organization attended the meeting. 27 papers were presented in three sessions, namely: general fuel testing programme (3 papers), fuel performance study (10 papers), in-core, on-site and hot cell technique (8 papers). A separate abstract was prepared for each of these papers. Three syndicate meetings allowed participants to discuss the papers and to draw up conclusions and recommendations

  17. Microstructure of irradiated Inconel 706 fuel pin cladding

    International Nuclear Information System (INIS)

    Yang, W.J.S.; Makenas, B.J.

    1983-08-01

    A fuel pin from the HEDL-P-60 experiment with a cladding of solution-annealed Inconel 706 breached in an apparently brittle manner at a position 12.7 cm above the bottom of the fuel column with a crack of 5.72 cm in length after 5.0 atomic percent burnup in EBR-II. Temperatures (time-averaged midwall) and fast fluences for the fractured area range from 447 0 C and 5.5 x 10 22 n/cm 2 to 526 0 C and 6.1 x 10 22 n/cm 2 (E > 0.1 MeV). Specimens of the fractured fuel pin section were successfully prepared and examined in both a scanning electron microscope and a transmission electron microscope. The fracture surfaces of the breached section showed brittle intergranular fracture characteristics for both the axial and circumferential cracks. Formation of γ' in the matrix near the breach confirmed that the irradiation temperature at the breached area was below 500 0 C, in agreement with other estimates of the temperature for the area, 447 to 526 0 C. A hexagonal eta-phase, Ni 3 (Ti,Nb), precipitated at boundaries near the breach. A more extensive eta-phase coating at grain boundaries was found in a section irradiated at 650 0 C. The eta-phase plates at grain boundaries are expected to have a detrimental effect on alloy ductility. A plane of weakness in this region along the (111) slip planes will develop in Inconel 706 because the eta-plates have a (111) habit relationship with the matrix

  18. Characterization of irradiated fuel rods using pulsed eddy current techniques

    International Nuclear Information System (INIS)

    Martin, M.R.; Francis, W.C.

    1975-11-01

    A number of irradiated fuel rods and unfueled zircaloy cladding tubes (''water tubes'') were obtained from the Saxton reactor through arrangements with the Westinghouse Electric Corporation for use in subsequent irradiation effects and fuel behavior programs. A comprehensive nondestructive and corroborative destructive characterization program was undertaken on these fuel rods and tubes by ANC to provide baseline data on their characteristics prior to further testing and for comparison against post-post data. This report deals primarily with one portion of the NDT program performed remotely in the hot cells. The portion of interest in this paper is the pulsed eddy current inspection used in the nondestructive phase of the work. 6 references

  19. Heat and radiation analysis of NPP Krsko irradiated fuel

    International Nuclear Information System (INIS)

    Lalovic, M.

    1986-01-01

    Radioactive and heat potential for irradiated fuel in the region 2 with burnup of 13400 MWd/tHM, and in the region 4A with burnup of 9360 MWd/tHM for NPP KRSKO, was calculated. Computer code KORIGEN (Karlsruhe Oak Ridge Isotope Generation and Depletion Code) was used. The aspects of radiation (mainly gamma and neutrons) and of heat production was considered with respect to their impact on fuel handing and waste management. Isotopic concentrations for irradiated fuel was calculated and compared with Westinghouse data. (author)

  20. Analysis of irradiation temperature in fuel rods of OGL-1 fuel assembly

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Minato, Kazuo; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-10-01

    Irradiation temperature in the fuel rods of 5th OGL-1 fuel assembly was analysed by the system composed by STPDSP2 and TRUMP codes. As the measured input-data, following parameters were allowed for; circumferential heating distribution around the fuel rod, which was measured in the JMTR critical assembly, axial heating distribution through the fuel rod, ratio of peak heatings of three fuel rods, and pre- and post-irradiation outer radii of the fuel compacts and inner radii of the graphite sleeves, which had been measured in PIE of the 5th OGL-1 fuel assembly. In computation the axial distributions of helium coolant temperature through the fuel rod and the heating value of each fuel rod were, firstly, calculated as input data for TRUMP. The TRUMP calculation yielded the temperatures which were fitted in those measured by all of the thermo-couples installed in the fuel rods, by adjusting only the value of the surface heat transfer coefficient, and consequently, the temperatures in all portions of the fuel rod were obtained. The apparent heat transfer coefficient changed to 60% of the initial values in the middle period of irradiation. For this reduction it was deduced that shoot had covered the surface of the fuel rod during irradiation, which was confirmed in PIE. Beside it, several things were found in this analysis. (author)

  1. The irradiation behavior of atomized U-Mo alloy fuels at high temperature

    Science.gov (United States)

    Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

    2001-04-01

    Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

  2. Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

  3. Irradiation effects on thermal properties of LWR hydride fuel

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt, E-mail: terrani@berkeley.edu [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Balooch, Mehdi [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Carpenter, David; Kohse, Gordon [Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Keiser, Dennis; Meyer, Mitchell [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Olander, Donald [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States)

    2017-04-01

    Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH{sub 1.6}) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

  4. Irradiation of Superheater Test Fuel Elements in the Steam Loop of the R2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ravndal, F

    1967-12-15

    The design, fabrication, irradiation results, and post-irradiation examination for three superheater test fuel elements are described. During the spring of 1966 these clusters, each consisting of six fuel rods, were successfully exposed in the superheater loop No. 5 in the R2 reactor for a maximum of 24 days at a maximum outer cladding surface temperature of {approx} 650 deg C. During irradiation the linear heat rating of the rods was in the range 400-535 W/cm. The diameter of the UO{sub 2} pellets was 11.5 and 13.0 mm; the wall thickness of the 20/25 Nb and 20/35 cladding was in every case 0.4 mm. The diametrical gap between fuel and cladding was one of the main parameters and was chosen to be 0.05, 0.07 and 0.10 mm. These experiments, to be followed by one high cladding temperature irradiation ({approx} 750 deg C) and one long time irradiation ({approx} 6000 MWd/tU), were carried out to demonstrate the operational capability of short superheater test fuel rods at steady and transient operational environments for the Marviken superheater fuel elements and also to provide confirmation of design criteria for the same fuel elements.

  5. An Analysis of the Thermal and Structure Behaviour of the UO{sub 2}-PuO{sub 2}-Fuel in the Irradiation Experiment of the UO{sub 2}-PuO{sub 2}-Fuel in the Irradiation Experiment FR2 Capsule Test Series 5a; Analisis termico y estructural del combustible UO{sub 2}-PuO{sub 2} irradiado en el reactor FR2 dentro del experimento KVE-Vg.5a

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.; Helmut, E.

    1981-07-01

    In the Karlsruhe research reactor FR2 nine fuel pins were irradiated within three irradiation capsules in the course of the test series 5a. The pins contained UO{sub 2}-PuO{sub 2} fuel pellets. They reached bump values of about 6, 17 and 47 Mwd/Kg Me with linear rod powers of 400 to 600 W/cm and clad surface temperature between 500 and 700 degree centigree. A detailed analysis of the fuel structuration data (columnar-grain and equiaxed- -grain growth regions) have allowed to determine, with the help of physic-mathematical models, the radii of these regions and the heat transfer through the contact zone between fuel and clad depending on the bump. The results of the analysis showed that the fuel surface temperature rose with increasing burnup. (Author) 16 refs.

  6. Microstructural evolution and Am migration behaviour in Am-containing fuels at the initial stage of irradiation

    International Nuclear Information System (INIS)

    Tanaka, Kosuke; Miwa, Shuhei; Sato, Isamu; Osaka, Masahiko; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shin-ichi; Yoshimochi, Hiroshi; Tanaka, Kenya

    2010-01-01

    In order to investigate the effect of americium addition to MOX fuels on the irradiation behaviour, the 'Am-1' programme is being conducted in JAEA. The Am-1 programme consists of two short-term irradiation tests of 10-minute and 24-hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post-irradiation examinations (PIE) are in progress. The PIE for Am-containing MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation and the results to date are reported. The successful development of fabrication technology with remote handling and the evaluation of thermo-chemical properties based on the out-of-pile experiments are described with an emphasis on the effects of Am addition on the MOX fuel properties. (authors)

  7. MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

    1997-08-01

    Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

  8. Irradiation of Parts of the X-Gen Nuclear Fuel Assembly made by KNF in HANARO

    International Nuclear Information System (INIS)

    Choo, K. N.; Cho, M. S.; Shin, Y. T.; Kim, B. G.; Lee, S. H.; Eom, K. B.

    2008-01-01

    An instrumented capsule has been developed at HANARO (High flux Advanced Neutron Application ReactOr) for the neutron irradiation tests of materials. The capsule system has been actively utilized for the various material irradiation tests requested by users from research institutes, universities, and the industries. As a preliminary test, some specimens made of the parts of a nuclear fuel assembly were inserted in the 05M-07U instrumented capsule and successfully irradiated at HANARO. Based on the results and experience, a new irradiation capsule of 07M-13N was designed, fabricated, and irradiated at HANARO for the evaluation of the neutron irradiation properties of the parts of the X-Gen nuclear fuel assembly made by KNF (Korea Nuclear Fuel). Specimens such as bucking and spring test specimens of spacer grid, microstructure and tensile test specimens of welded parts, tensile, irradiation growth and spring test specimens made of HANA tube, Zirlo, Zircaloy-4 and Inconel-718 were placed in the capsule. The capsule was loaded into the CT test hole of HANARO of a 30MW thermal output and the specimens were irradiated at 295 - 460 .deg. C up to a fast neutron fluence of 1.2x10 21 (n/cm 2 ) (E>1.0MeV)

  9. Safety assessment of a dry storage container drop into irradiated fuel bays

    International Nuclear Information System (INIS)

    Parlatan, Y.; Oh, D.; Arguner, D.; Lei, Q.M.; Kulpa, T.; Bayoumi, M.H.

    2004-01-01

    In Pickering nuclear stations, Dry Storage Containers (DSCs) are employed to transfer used (irradiated) fuel from an irradiated fuel bay to a dry storage facility for interim storage. Each DSC is wet-loaded in the bay water with 4 fuel modules containing up to a total of 384 used fuel bundles that have been out of the reactor core for at least 10 years. Once the DSC is fully loaded, the crane in the bay raises the DSC for spray-wash such that the bottom of the DSC is never more than 2 m above the bay water surface. This paper presents a safety assessment of consequences of an unlikely event that a fully loaded DSC is accidentally dropped into an irradiated fuel bay from the highest possible elevation. Experiments and analyses performed elsewhere show that the DSC drop-generated shock waves will not threaten the structural integrity of an irradiated fuel bay. Therefore, this assessment only assesses the potential damage to the spent fuel bundles in the bay due to pressure transients generated by an accidental DSC drop. A bounding estimate approach has been used to calculate the upper limit of the pressure pulse and the resulting static and dynamic stresses on the fuel sheath. The bounding calculations and relevant experimental results demonstrate that an accidental drop of a fully loaded DSC into an irradiated fuel bay will not cause additional failures of the main fuel inventories stored in modules in the bay water, thus no consequential release of fission products into the bay water. (author)

  10. In-pile measurement of the thermal conductivity of irradiated metallic fuel

    International Nuclear Information System (INIS)

    Bauer, T.H.; Holland, J.W.

    1995-01-01

    Transient test data and posttest measurements from recent in-pile overpower transient experiments are used for an in situ determination of metallic fuel thermal conductivity. For test pins that undergo melting but remain intact, a technique is described that relates fuel thermal conductivity to peak pin power during the transient and a posttest measured melt radius. Conductivity estimates and their uncertainty are made for a database of four irradiated Integral Fast Reactor-type metal fuel pins of relatively low burnup (<3 at.%). In the assessment of results, averages and trends of measured fuel thermal conductivity are correlated to local burnup. Emphasis is placed on the changes of conductivity that take place with burnup-induced swelling and sodium logging. Measurements are used to validate simple empirically based analytical models that describe thermal conductivity of porous media and that are recommended for general thermal analyses of irradiated metallic fuel

  11. Irradiation-induced dimensional changes of fuel compacts and graphite sleeves of OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Minato, Kazuo; Kobayashi, Fumiaki; Tobita, Tsutomu; Kikuchi, Teruo; Kurobane, Shiro; Adachi, Mamoru; Fukuda, Kousaku

    1988-06-01

    Experimental data are summarized on irradiation-induced dimensional changes of fuel compacts and graphite sleeves of the first to ninth OGL-1 fuel assemblies. The range of fast-neutron fluence is up to 4 x 10 24 n/m 2 (E > 0.18 MeV); and that of irradiation temperature is 900 - 1400 deg C for fuel compacts and 800 - 1050 deg C for graphite sleeves. The dimensional change of the fuel compacts was shrinkage under these test conditions, and the shrinkage fraction increased almost linearly with fast-neutron fluence. The shrinkage fraction of the fuel compacts was larger by 20 % in the axial direction than in the radial direction. Influence of the irradiation temperature on the dimensional-change behavior of the fuel compacts was not observed clearly; presumably the influence was hidden by scatter of the data because of low level of the fast-neutron fluence and the resultant small dimensional changes. (author)

  12. Study of Irradiation Effect onto Uranium silicide Fuel

    International Nuclear Information System (INIS)

    Suparjo

    1998-01-01

    The irradiation effect onto the U 3 Si-Al and U 3 Si 2 -Al dispersion type of fuel element has been studied. The fuel material performs swelling during irradiation due to boehmite (Al 2 O 3 (H 2 O)) formation in which might occurs inside the meat and on the cladding surface, the interaction between the fuel and aluminium matrix that produce U(Al,Si) 3 phase, and the formation of fission gas bubble inside the fuel. At a constant fission density, the U 3 Si-Al fuel swelling is higher than that of U 3 Si 2 -Al fuel. The swellings of both fuels increase with the increasing of fission density. The difference of swelling behavior was caused by formation of large bubble gases generated from fission product of U 3 Si fuel and distributed non-uniformly over all of fuel zone. On the other hand, the U 3 Si 2 fission produced small bubble gases, and those were uniformly distributed. The growth rate of fission gas bubble in the U 3 Si fuel has shown high diffusivity, transformation into amorph material and thus decrease its mechanical strength

  13. Fuel utilization experience in Slovak Republic

    Energy Technology Data Exchange (ETDEWEB)

    Petenyi, V [Nuclear Regulatory Authority of the Slovak Republic, Bajkalska (Slovakia)

    1997-12-01

    The paper summarizes shortly the gained experience in utilization of the fuel in the four-year fuel cycles and describes the future activities in fuel management. The spent fuel management is also included. (author). 2 refs, 2 figs, 1 tab.

  14. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1993-03-01

    This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the Nuclear Regulatory Commission (NRC), and to meet the requirements of Public Law 96--295. The report provides a brief description of NRC authority for certain aspects of transporting spent fuel. It provides descriptive statistics on spent fuel shipments regulated by the NRC from 1979 to 1992. It also lists detailed highway and railway segments used within each state from October 1, 1987 through December 31, 1992

  15. Correlations between fuel pins irradiated in fast and thermal fluxes using the frump fuel pin modelling program

    International Nuclear Information System (INIS)

    Hayns, M.R.; Adam, J.

    1975-08-01

    There is no experimental facilities in which a fuel pin can be irradiated in a fast environment under well defined conditions of over power or flow run down. Consequently most of the infor mation which is being accumulated on the behaviour of fuel pins under severe conditions is obtained from either capsule or loop rigs in thermal reactors. It is the purpose of this paper to highlight the differences between the behaviour of fuel pins irradiated in a thermal flux and a fast flux. A typical set of conditions is taken from an overpower experiment in a thermal flux and the behaviour of the system is analysed using the fuel modelling program FRUMP. A second numerical experiment is then performed in which the same conditions prevail, except that a fast flux is assumed, the criterion for comparison being that the total power input to the system is the same in both cases. From the many possible correlations which result from such an exercise the fuel tempreature has been selected to highlight various important features of the two irradiations. It is demonstrated that the flux depression can cause differences in the pin behaviour, even to altering the order of events in a transient. For example fuel melting will occur at different times and at different positions in the fuel in the two cases. It is concluded that the techniques of fuel modelling, as typified in the program FRUMP can provide a very useful tool indeed for the analysis of such experiments and for guiding the establishment of the appropriate correlations for the extrapolation to the fast flux case. (author)

  16. Evaluation model for PWR irradiated fuel

    International Nuclear Information System (INIS)

    Gomes, I.C.

    1983-01-01

    The individual economic value of the plutonium isotopes for the recycle of the PWR reactor is investigated, assuming the existence of an market for this element. Two distinct market situations for the stages of the fuel cycle are analysed: one for the 1972 costs and the other for costs of 1982. Comparisons are made for each of the two market situations concerning enrichment of the U-235 in the uranium fuel that gives the minimum cost in the fuel cycle. The method adopted to establish the individual value of the plutonium isotopes consists on the economical analyses of the plutonium fuel cycle for four different isotopes mixtures refering to the uranium fuel cycle. (Author) [pt

  17. Irradiation behaviour of advanced fuel elements for the helium-cooled high temperature reactor (HTR)

    International Nuclear Information System (INIS)

    Nickel, H.

    1990-05-01

    The design of modern HTRs is based on high quality fuel. A research and development programme has demonstrated the satisfactory performance in fuel manufacturing, irradiation testing and accident condition testing of irradiated fuel elements. This report describes the fuel particles with their low-enriched UO 2 kernels and TRISO coating, i.e. a sequence of pyrocarbon, silicon carbide, and pyrocarbon coating layers, as well as the spherical fuel element. Testing was performed in a generic programme satisfying the requirements of both the HTR-MODUL and the HTR 500. With a coating failure fraction less than 2x10 -5 at the 95% confidence level, the results of the irradiation experiments surpassed the design targets. Maximum accident temperatures in small, modular HTRs remain below 1600deg C, even in the case of unrestricted core heatup after depressurization. Here, it was demonstrated that modern TRISO fuels retain all safety-relevant fission products and that the fuel does not suffer irreversible changes. Isothermal heating tests have been extended to 1800deg C to show performance margins. Ramp tests to 2500deg C demonstrate the limits of present fuel materials. A long-term programm is planned to improve the statistical significance of presently available results and to narrow remaining uncertainty limits. (orig.) [de

  18. Updated FY12 Ceramic Fuels Irradiation Test Plan

    International Nuclear Information System (INIS)

    Nelson, Andrew T.

    2012-01-01

    The Fuel Cycle Research and Development program is currently devoting resources to study of numerous fuel types with the aim of furthering understanding applicable to a range of reactors and fuel cycles. In FY11, effort within the ceramic fuels campaign focused on planning and preparation for a series of rabbit irradiations to be conducted at the High Flux Isotope Reactor located at Oak Ridge National Laboratory. The emphasis of these planned tests was to study the evolution of thermal conductivity in uranium dioxide and derivative compositions as a function of damage induced by neutron damage. Current fiscal realities have resulted in a scenario where completion of the planned rabbit irradiations is unlikely. Possibilities for execution of irradiation testing within the ceramic fuels campaign in the next several years will thus likely be restricted to avenues where strong synergies exist both within and outside the Fuel Cycle Research and Development program. Opportunities to augment the interests and needs of modeling, advanced characterization, and other campaigns present the most likely avenues for further work. These possibilities will be pursued with the hope of securing future funding. Utilization of synthetic microstructures prepared to better understand the most relevant actors encountered during irradiation of ceramic fuels thus represents the ceramic fuel campaign's most efficient means to enhance understanding of fuel response to burnup. This approach offers many of the favorable attributes embraced by the Separate Effects Testing paradigm, namely production of samples suitable to study specific, isolated phenomena. The recent success of xenon-imbedded thick films is representative of this approach. In the coming years, this strategy will be expanded to address a wider range of problems in conjunction with use of national user facilities novel characterization techniques to best utilize programmatic resources to support a science-based research program.

  19. SLIGHTLY IRRADIATED FUEL (SIF) INTERIM DISPOSITION PROJECT

    International Nuclear Information System (INIS)

    Norton, S.H.

    2010-01-01

    CH2M HILL Plateau Remediation Company (CH2M HILL PRC) is proud to submit the Slightly Irradiated Fuel (SIF) Interim Disposition Project for consideration by the Project Management Institute as Project of the Year for 2010. The SIF Project was a set of six interrelated sub-projects that delivered unique stand-alone outcomes, which, when integrated, provided a comprehensive and compliant system for storing high risk special nuclear materials. The scope of the six sub-projects included the design, construction, testing, and turnover of the facilities and equipment, which would provide safe, secure, and compliant Special Nuclear Material (SNM) storage capabilities for the SIF material. The project encompassed a broad range of activities, including the following: Five buildings/structures removed, relocated, or built; Two buildings renovated; Structural barriers, fencing, and heavy gates installed; New roadways and parking lots built; Multiple detection and assessment systems installed; New and expanded communication systems developed; Multimedia recording devices added; and A new control room to monitor all materials and systems built. Project challenges were numerous and included the following: An aggressive 17-month schedule to support the high-profile Plutonium Finishing Plant (PFP) decommissioning; Company/contractor changeovers that affected each and every project team member; Project requirements that continually evolved during design and construction due to the performance- and outcome-based nature ofthe security objectives; and Restrictions imposed on all communications due to the sensitive nature of the projects In spite of the significant challenges, the project was delivered on schedule and $2 million under budget, which became a special source of pride that bonded the team. For years, the SIF had been stored at the central Hanford PFP. Because of the weapons-grade piutonium produced and stored there, the PFP had some of the tightest security on the Hanford

  20. SLIGHTLY IRRADIATED FUEL (SIF) INTERIM DISPOSITION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    NORTON SH

    2010-02-23

    CH2M HILL Plateau Remediation Company (CH2M HILL PRC) is proud to submit the Slightly Irradiated Fuel (SIF) Interim Disposition Project for consideration by the Project Management Institute as Project of the Year for 2010. The SIF Project was a set of six interrelated sub-projects that delivered unique stand-alone outcomes, which, when integrated, provided a comprehensive and compliant system for storing high risk special nuclear materials. The scope of the six sub-projects included the design, construction, testing, and turnover of the facilities and equipment, which would provide safe, secure, and compliant Special Nuclear Material (SNM) storage capabilities for the SIF material. The project encompassed a broad range of activities, including the following: Five buildings/structures removed, relocated, or built; Two buildings renovated; Structural barriers, fencing, and heavy gates installed; New roadways and parking lots built; Multiple detection and assessment systems installed; New and expanded communication systems developed; Multimedia recording devices added; and A new control room to monitor all materials and systems built. Project challenges were numerous and included the following: An aggressive 17-month schedule to support the high-profile Plutonium Finishing Plant (PFP) decommissioning; Company/contractor changeovers that affected each and every project team member; Project requirements that continually evolved during design and construction due to the performance- and outcome-based nature ofthe security objectives; and Restrictions imposed on all communications due to the sensitive nature of the projects In spite of the significant challenges, the project was delivered on schedule and $2 million under budget, which became a special source of pride that bonded the team. For years, the SIF had been stored at the central Hanford PFP. Because of the weapons-grade piutonium produced and stored there, the PFP had some of the tightest security on the Hanford

  1. Design, Manufacturing and Irradiation Behaviour of Fast Reactor Fuel. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2013-04-01

    Fast reactors are vital for ensuring the sustainability of nuclear energy in the long term. They offer vastly more efficient use of uranium resources and the ability to burn actinides, which are otherwise the long-lived component of high level nuclear waste. These reactors require development, qualification, testing and deployment of improved and innovative nuclear fuel and structural materials having very high radiation resistance, corrosion/erosion and other key operational properties. Several IAEA Member States have made efforts to advance the design and manufacture of technologies of fast reactor fuels, as well as to investigate their irradiation behaviour. Due to the acute shortage of fast neutron testing and post-irradiation examination facilities and the insufficient understanding of high dose radiation effects, there is a need for international exchange of knowledge and experience, generation of currently missing basic data, identification of relevant mechanisms of materials degradation and development of appropriate models. Considering the important role of nuclear fuels in fast reactor operation, the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) proposed a Technical Meeting (TM) on 'Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels', which was hosted by the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russian Federation, from 30 May to 3 June 2011. The TM included a technical visit to the fuel production plant MSZ in Elektrostal. The purpose of the meeting was to provide a forum to share knowledge, practical experience and information on the improvement and innovation of fuels for fast reactors through scientific presentations and brainstorming discussions. The meeting brought together 34 specialists from national nuclear agencies, R and D and design institutes, fuel vendors and utilities from 10 countries. The presentations were structured into four sections: R and D Programmes on FR Fuel

  2. Reprocessing of irradiated fuel: pros and cons

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  3. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1982-06-01

    This publication is the third in a proposed series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. Subsequent issues in this series will update the information contained herein. This publication contains basically three kinds of information: (1) routes approved by the Commission for the shipment of irradiated reactor fuel, (2) information regarding any safeguards-significant incidents which have been reported to occur during shipments along such routes, and (3) cumulative amounts of material shipped

  4. Monitoring for fuel sheath defects in three shipments of irradiated CANDU nuclear fuel

    International Nuclear Information System (INIS)

    Johnson, H.M.

    1978-01-01

    Analyses of radioactive gases within the Pegase shipping flask were performed at the outset and at the completion of three shipments of irradiated nuclear fuel from the Douglas Point Generating Station to Whiteshell Nuclear Research Establishment. No increases in the concentration of active gases, volatiles or particulates were observed. The activity of the WR-1 bay water rose only marginally due to the storage of the fuel. Other tests indicated that minimal surface contamination was present. These data established that defects in fuel element sheaths did not arise during the transport or the handling of this irradiated fuel. The observation has significance for the prospect of irradiated nuclear fuel transfer and handling in preparation for storage or disposal. (author)

  5. Irradiation performance of uranium-molybdenum alloy dispersion fuels

    International Nuclear Information System (INIS)

    Almeida, Cirila Tacconi de

    2005-01-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm 3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm 3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  6. Basic properties of fuel determining its behavior under irradiation

    International Nuclear Information System (INIS)

    Konovalov, I.I.

    2000-01-01

    The theoretical model describing a swelling of nuclear fuel at low irradiation temperatures is considered. The critical physical parameters of substances determining behavior of point defects, gas fission atoms, dislocation density, nucleation and growth of gas-contained pores are determined. The correlation between meanings of critical parameters and physical properties of substance is offered. The accounts of swelling of various dense fuels with reference to work in conditions of research reactors are given. (author)

  7. What destiny could be given to the nuclear irradiated fuel

    International Nuclear Information System (INIS)

    Mundim, S.G.

    1985-01-01

    The uranium used in nuclear plants in the production of electric energy is not totally consumed. Part of the fuel that is left over is composed of radioactive material, that represents great danger to earth life. The destines that could be given to the irradiated fuel - reprocessing, provisional or definite storage - depend on the policy adopted by each country that enters the nuclear era, being involved in this increasing problem. (Author) [pt

  8. Clarification of dissolved irradiated light-water-reactor fuel

    International Nuclear Information System (INIS)

    Rodrigues, G.C.

    1983-02-01

    Bench-scale studies with actual dissolved irradiated light water reactor (LWR) fuels showed that continuous centrifugation is a practical clarification method for reprocessing. Dissolved irradiated LWR fuel was satisfactorily clarified in a bench-scale, continuous-flow bowl centrifuge. The solids separated were successfully reslurried in water. When the reslurried solids were mixed with clarified centrate, the resulting suspension behaved similar to the original dissolver solution during centrifugation. Settling rates for solids in actual irradiated fuel solutions were measured in a bottle centrifuge. The results indicate that dissolver solutions may be clarified under conditions achievable by available plant-scale centrifuge technology. The effective particle diameter of residual solids was calculated to be 0.064 microns for Oconee-1 fuel and 0.138 microns for Dresden-1 fuel. Filtration was shown unsuitable for clarification of LWR fuel solutions. Conventional filtration with filter aid would unacceptably complicate remote canyon operation and maintenance, might introduce dissolved silica from filter aids, and might irreversibly plug the filter with dissolver solids. Inertial filtration exhibited irreversible pluggage with nonradioactive stand-in suspensions under all conditions tested

  9. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.

    2010-06-17

    A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.

  10. Study on the irradiation swelling of U3Si2-Al dispersion fuel

    International Nuclear Information System (INIS)

    Xing Zhonghu; Ying Shihao

    2001-01-01

    The dominant modeling mechanisms on irradiation swelling of U 3 Si 2 -Al dispersion fuel are introduced. The core of dispersion fuel is looked to as micro-fuel elements of continuous matrix. The formation processes of gas bubbles in the fuel phase are described through the behavior mechanisms of fission gases. The swelling in the fuel phase causes the interaction between fuel particles and metal matrix, and the metal matrix can restrain the irradiation swelling of fuel particles. The developed code can predict irradiation-swelling values according to the parameters of fuel elements and irradiation conditions, and the predicted values are in agreement with the measured results

  11. Bilateral cooperation between Germany and Brazil on fuel irradiation

    International Nuclear Information System (INIS)

    Dias, J.W.

    1977-01-01

    Within the framework of the Government Agreement on Scientific and Technical Cooperation between the Federal Republic of Germany and Brazil, the Brazilian National Atomic Commission and the Juelich Nuclear Research Center (KFA) signed on 23rd April, 1971 an Agreement on Cooperation in the field of Nuclear Research and Reactor Technology. Projects have been elaborated in fields of mutual interest to share activities between the partner institutes in both countries. A typical project is the fuel irradiation programme jointly prepared by NUCLEBRAS and KFA-Juelich. Brazil is planning to use elements of its own production in nuclear power plants to be erected within the German-Brazilian Industrial Agreement. As no material test reactor is available in Brazil it is expedient to irradiate samples of Brazilian production in Germany. Brazilian collaborators will participate in the preparation, execution and post-irradiation examination. In this way an optimum transfer of all information and results is assured. In the first phase, sample rods manufactured in Brazil are irradiated in the FRJ-2 test reactor in Juelich. These rods are assembled under clean conditions in the NUCLEBRAS research centres. The first Brazilian test rods showed excellent in-pile behaviour even under very high fuel rod capacity. In the second phase, fuel rods of original length manufactured and assembled in Brazil will be irradiated in German power plants, and, at the same time, additional irradiations of small samples will be carried out in test reactors. In the third phase, rod clusters and complete fuel elements will be manufactured in Brazil and irradiated in German power plants until target burn-up. All the necessary prerequisites have been fulfilled to meet the above requirements, i.e. mutual interest, good infrastructure maintained by both partners, qualified personnel and last but not least unbureaucratic and effective help by the coordinating offices of NUCLEBRAS and KFA

  12. KNF's fuel service technologies and experiences

    International Nuclear Information System (INIS)

    Shin, Jung Cheol; Kwon, Jung Tack; Kim, Jaeik; Park, Jong Youl; Kim, Yong Chan

    2009-01-01

    In Korea, since 1978, the commercial nuclear power plant was operated. After 10 years, from 1988, the nuclear fuel was produced by KNF (Korea Nuclear Fuel). The Fuel Service Team was established at KNF in 1995. Through the technical self reliance periods in cooperate with advanced foreign companies for 5 years, KNF has started to carry out fuel service activities onsite in domestic nuclear power plants. By ceaseless improving and advancing our own methodologies, after that, KNF is able to provide the most safe and reliable fuel repair services and poolside examinations including the root cause analysis of failed fuels. Recently, KNF developed the fuel cleaning system using ultrasonic technique for crud removal, and the CANDU fuel sipping system to detect a failed fuel bundle in PHWR. In this paper, all of KNF's fuel service technologies are briefly described, and the gained experience in shown

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

    Science.gov (United States)

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

    2017-12-01

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

  14. Final safety analysis report for the irradiated fuels storage facility

    International Nuclear Information System (INIS)

    Bingham, G.E.; Evans, T.K.

    1976-01-01

    A fuel storage facility has been constructed at the Idaho Chemical Processing Plant to provide safe storage for spent fuel from two commercial HTGR's, Fort St. Vrain and Peach Bottom, and from the Rover nuclear rocket program. The new facility was built as an addition to the existing fuel storage basin building to make maximum use of existing facilities and equipment. The completed facility provides dry storage for one core of Peach Bottom fuel (804 elements), 1 1 / 2 cores of Fort St. Vrain fuel (2200 elements), and the irradiated fuel from the 20 reactors in the Rover program. The facility is designed to permit future expansion at a minimum cost should additional storage space for graphite-type fuels be required. A thorough study of the potential hazards associated with the Irradiated Fuels Storage Facility has been completed, indicating that the facility is capable of withstanding all credible combinations of internal accidents and pertinent natural forces, including design basis natural phenomena of a 10,000 year flood, a 175-mph tornado, or an earthquake having a bedrock acceleration of 0.33 g and an amplification factor of 1.3, without a loss of integrity or a significant release of radioactive materials. The design basis accident (DBA) postulated for the facility is a complete loss of cooling air, even though the occurrence of this situation is extremely remote, considering the availability of backup and spare fans and emergency power. The occurrence of the DBA presents neither a radiation nor an activity release hazard. A loss of coolant has no effect upon the fuel or the facility other than resulting in a gradual and constant temperature increase of the stored fuel. The temperature increase is gradual enough that ample time (28 hours minimum) is available for corrective action before an arbitrarily imposed maximum fuel centerline temperature of 1100 0 F is reached

  15. Precise measurement of fuel content of irradiated and nonirradiated materials

    International Nuclear Information System (INIS)

    Harker, Y.D.; Napper, P.R.; Proctor, A.E.

    1984-01-01

    This paper discusses the application of precise reactivity measurements in the Advanced Reactivity Measurement Facility at Idaho National Engineering Laboratory (INEL) to determine th fuel content in irradiated and nonirradiated materials. Different methods of reactivity measurements and examples of how they have been are presented, which provides an insight in capabilities available to analyze samples with different geometrical sizes from small volumes approx. 100 cc to 12 ft long fuel pins and also samples with different fuel content ranges from approx. 2 mg to approx. 600 g. The overall accuracy of these measurements is approx. 0.5% (1sigma)

  16. Behavior of pre-irradiated fuel under a simulated RIA condition. Results of NSRR Test JM-5

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Tanzawa, Sadamitsu; Ishijima, Kiyomi; Kobayashi, Shinsho; Kamata, Hiroshi; Homma, Kozo; Sakai, Haruyuki.

    1995-11-01

    This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-5, conducted in the Nuclear Safety Research Reactor (NSRR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and interpretations and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 25.7 MWd/kgU with average linear heat rate of 33.4 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a desposited energy of 223 ± 7 cal/g·fuel (0.93 ± 0.03 kJ/g·fuel) and a peak fuel enthalpy of 167 ± 5 cal/g·fuel (0.70 ± 0.02 kJ/g·fuel) under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The Test JM-5 resulted in cladding failure. More than twenty small cracks were found in the post-test cladding, and most of the defects located in pre-existing locally hydrided region. The result indicates an occurrence of fuel failure by PCMI (pellet/cladding mechanical interaction) in combination with decreased integrity of hydrided cladding. (author)

  17. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1983-07-01

    This publication contains basically three kinds of information: routes approved by the Commission for the shipment of irradiated reactor fuel, information regarding any safeguards-significant incidents which have been reported to occur during shipments along such routes, and cumulative amounts of material shipped

  18. Isotope correlation and mass spectrometry techniques for irradiated fuel assay

    International Nuclear Information System (INIS)

    Deron, S.

    1985-01-01

    This paper outlines the methods used to account for fissionable materials in irradiated nuclear fuel elements entering reprocessing plants. Verification is accomplished at three mass balance stations in the plant. Techniques employed fall into two categories: isotopic and isotope dilution analyses by mass spectometry and isotope correlation techniques. These methods are discussed in some detail

  19. Irradiation of defected SAP clad UO2 fuel in the X-7 organic loop

    International Nuclear Information System (INIS)

    Robertson, R.F.S.; Cracknell, A.G.; MacDonald, R.D.

    1961-10-01

    This report describes an experiment designed to test the behaviour under irradiation of a UO 2 fuel specimen clad in a defected SAP sheath and cooled by recirculating organic liquid. The specimen containing the defect was irradiated in the X-7 loop in the NRX reactor from the 25th of November until the 13th of December 1960. Up to the 13th of December the behaviour was analogous to that seen with defected UO 2 specimens clad in zircaloy which were irradiated in water loops. Reactor power transients resulted in peaking of gamma ray activities in the loop, but on steady operation these activities tended to fall to a steady state level, Over this period the pressure drop across the fuel increased by a factor of two, the increases occurring after reactor shut downs and start ups. On 13th December the pressure drop increased rapidly, after a reactor shut down and start up, to over five times its original value and the activities in the loop rose to a high level. The specimen was removed and examination showed that the sheath was very badly split and that the volume between the fuel and the sheath was filled with a hard black organic substance. This report gives full details of the irradiation and of the post -irradiation examination. Correlation of the observed phenomenon is attempted and a preliminary assessment of the problems which would be associated with defect fuel in an organic reactor is given. (author)

  20. Mechanisms of microstructural changes of fuel under irradiation

    International Nuclear Information System (INIS)

    Garcia, P.; Carlot, G.; Dorado, B.; Maillard, S.; Sabathier, C.; Martin, G.; Oh, J.Y.; Welland, M.J.

    2015-01-01

    Nuclear fuels are subjected to high levels of radiation damage mainly due to the slowing of fission fragments, which results in substantial modifications of the initial fuel microstructure. Microstructure changes alter practically all engineering fuel properties such as atomic transport or thermomechanical properties so understanding these changes is essential to predicting the performance of fuel elements. Also, with increasing burn-up, the fuel drifts away from its initial composition as the fission process produces new chemical elements. Because nuclear fuels operate at high temperature and usually under high-temperature gradients, damage annealing, foreign atom or defect clustering and migration occur on multiple time and length scales, which make long-term predictions difficult. The end result is a fuel microstructure which may show extensive differences on the scale of a single fuel pellet. The main challenge we are faced with is, therefore, to identify the phenomena occurring on the atom scale that are liable to have macroscopic effects that will determine the microstructure changes and ultimately the life-span of a fuel element. One step towards meeting this challenge is to develop and apply experimental or modelling methods capable of connecting events that occur over very short length and timescales to changes in the fuel microstructure over engineering length and timescales. In the first part of this chapter, we provide an overview of some of the more important microstructure modifications observed in nuclear fuels. The emphasis is placed on oxide fuels because of the extensive amount of data available in relation to these materials under neutron or ion irradiation. When possible and relevant, the specifics of other types of fuels such as metallic or carbide fuels are alluded to. Throughout this chapter but more specifically in the latter part, we attempt to give examples of how modelling and experimentation at various scales can provide us with

  1. Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

    Science.gov (United States)

    Mcdonald, G. E.

    1973-01-01

    A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

  2. DUPIC fuel irradiation test and performance evaluation; the performance analysis of pellet-cladding contact fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ho, K. I.; Kim, H. M.; Yang, K. B.; Choi, S. J. [Suwon University, Whasung (Korea)

    2002-04-01

    Thermal and mechanical models were reviewed, and selected for the analysis of nuclear fuel performance in reactor. 2 dimensional FEM software was developed. Thermal models-gap conductances, thermal conductivity of pellets, fission gas release, temperature distribution-were set and packaged into a software. Both thermal and mechanical models were interrelated to each other, and the final results, fuel performance during irradiation is obtained by iteration calculation. Also, the contact phenomena between pellet and cladding was analysed by mechanical computer software which was developed during this work. dimensional FEM program was developed which estimate the mechanical behavior and the thermal behaviors of nuclear fuel during irradiation. Since there is a importance during the mechanical deformation analysis in describing pellet-cladding contact phenomena, simplified 2 dimensional calculation method is used after the contact. The estimation of thermal fuel behavior during irradiation was compared with the results of other. 8 refs., 17 figs. (Author)

  3. Element bow profiles from new and irradiated CANDU fuel bundles

    International Nuclear Information System (INIS)

    Dennier, D.; Manzer, A.M.; Ryz, M.A.

    1996-01-01

    Improved methods of measuring element profiles on new CANDU fuel bundles were developed at the Sheridan Park Engineering Laboratory, and have now been applied in the hot cells at Whiteshell Laboratories. For the first time, the outer element profiles have been compared between new, out-reactor tested, and irradiated fuel elements. The comparison shows that irradiated element deformation is similar to that observed on elements in out-reactor tested bundles. In addition to the restraints applied to the element via appendages, the element profile appears to be strongly influenced by gravity and the end loads applied by local deformation of the endplate. Irradiation creep in the direction of gravity also tends to be a dominant factor. (author)

  4. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1991-01-01

    This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1989 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials. 11 figs., 3 tabs

  5. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1992-06-01

    The circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1991 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials

  6. Examples of remote handling of irradiated fuel assemblies in Germany

    International Nuclear Information System (INIS)

    Peehs, M.; Knecht, K.

    1999-01-01

    Examples for the remote handling of irradiated fuel in Germany are presented in the following areas: - fuel assembling pool service activities; - early encapsulation of spent fuel in the pool of a nuclear power plant (NPP) at the end of the wet storage period. All development in remote fuel assembly handling envisages minimization of the radioactive dose applied to the operating staff. In the service area a further key objective for applying advanced methods is to perform the work faster and at a higher quality standard. The early encapsulation is a new technology to provide the final packaging of spent fuel already in the pool of a NPP to ensure reliable handling for all further back end processes. (author)

  7. Examination in hot laboratories of irradiated fuels from fast reactors

    International Nuclear Information System (INIS)

    Clottes, G.; Peray, R.; Ratier, J.L.

    1980-05-01

    Low irradiation rate examinations were carried out soon after the Rapsodie, Rapsodie Fortissimo and Phenix reactors were started up for the first time in order to check the level of maximum temperatures reached and the radial migration of oxygen and plutonium and to assess the movements of fuels inside the cladding. The other examinations were effected at a high specific burnup in order to defines the limit specific burnup securing the integrity of the fuel pin claddings (distortion, ruptures and possible consequences). The examinations carried out so far on fuel elements coming from Phenix or Rapsodie have allowed good fuel surveillance to be undertaken and the acquisition of a large number of data, thanks to which the fuel characteristics of future reactors of the system have been developed [fr

  8. Irradiation performance of (Th,U)O2 fuel designed for advanced cycle applications

    International Nuclear Information System (INIS)

    Hastings, I.J.; Celli, A.; Onofrei, M.; Swanson, M.L.

    1982-06-01

    Our reference fabrication route for Advanced Cycle thoria-based fuel is conventional in that it produces cold-pressed and sintered pellets. However, we are also evaluating alternative fuels which offer the potential for simpler fabrication in a remote facility, and in some cases improved high burnup performance. These alternatives are impregnated, spherepac, and extruded thoria-based fuels. Spherepac fuel has been irradiated at a linear power of 50-60 kW/m to about 180 MW.h/kg H.E. There have been unexplained defects in fuel with both free-standing and collapsible cladding. Impegnated fuel has operated to 650 MW.h/kg H.E. at 50-60 kW/m. An experiment examining fuel from the sol-gel extrusion process has reached 450 Mw.h/kg H.E. at a maximum linear power of 60 kW/m. The latter two experiments have operated without defects and with fission gas release less than that for UO 2 under identical conditions. The extruded fuel has a pellet geometry similar to that for conventional fuel and is AECL's first practical demonstration of thoria-based fuel with the fissile component distributed homogeneously on an atomic scale

  9. EDF requirements for hot cells examinations on irradiated fuel

    International Nuclear Information System (INIS)

    Segura, J.C.; Ducros, G.

    2002-01-01

    The objectives of increasing French Nuclear Power Plants (NPP) availability while lengthening the fuel irradiation cycle and reaching higher burnups lead EDF to carry out on site and hot cell examinations. The data issued from such fuel behaviour monitoring programmes will be used to ascertain that the design criteria are met. Data are also needed for modelling, development and validation. The paper deals quickly with the logistics linked to the selection and transport of fuel rods from NPP to hot cell laboratory. Hot cell PIEs remain a valuable method to obtain data in such fields as PCI (Pellet-Cladding Interaction), internal pressure, FGR (Fission Gas Release), oxide thickness, metallurgical aspects. The paper introduces burnup determination methods, inner pressure evaluation, preparation of samples for further irradiation such as power ramps for PCI and RIA (Reactivity Initiated Accident) testing. The nuclear microprobe of Perre Suee laboratory is also presented. (author)

  10. Irradiation test plan of instrumented capsule(05F-01K) for nuclear fuel irradiation in Hanaro (Revision 1)

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jae Min; Kim, B. G.; Choi, M. H. (and others)

    2006-09-15

    An instrumented capsule was developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel pellet elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.84 full power days at 24 MW). In the year of 2004, 3 test fuel rods and the 03F-05K instrumented fuel capsule were designed and fabricated to measure fuel centerline temperature, internal pressure of fuel rod, and fuel axial deformation during irradiation test. Now, this capsule was successfully irradiated in the test hole OR5 of HANARO reactor from April 27, 2004 to October 1, 2004 (59.5 full power days at 24-30 MW). The capsule and fuel rods have been be dismantled and fuel rods have been examined at the hot cell of IMEF. The instrumented fuel capsule (05F-01K) was designed and manufactured for a design verification test of the dual instrumented fuel rods. The irradiation test of the 05F-01K instrumented fuel capsule will be carried out at the OR5 vertical experimental hole of HANARO.

  11. 78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

    Science.gov (United States)

    2013-08-19

    ... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Orders; rescission. SUMMARY... the NRC published a final rule, ``Physical Protection of Irradiated Fuel in Transit,'' on May 20, 2013... of Irradiated Reactor Fuel in Transit'' (RIN 3150-AI64; NRC-2009-0163). The final rule incorporates...

  12. Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K.; Ross Finlay, M.

    2012-06-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

  13. Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, Dennis D., E-mail: Dennis.Keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Ross Finlay, M. [Australian Nuclear Science and Technology Organization, PMB 1, Menai, NSW 2234 (Australia)

    2012-06-15

    The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

  14. The Canadian CANDU fuel development program and recent fuel operating experience

    International Nuclear Information System (INIS)

    Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

    1999-01-01

    This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the two-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These two programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

  15. Results of preliminary experiments on tritium decontamination by UV irradiation

    International Nuclear Information System (INIS)

    Oya, Yasuhisa; Shu, Wataru; O'hira, Shigeru; Hayashi, Takumi; Nishi, Masataka

    2000-03-01

    In the point of view of protection of workers from the radiation exposure and the limitation of the contamination with radioactive materials, it is important to decontaminate mobile tritium from plasma facing components of a nuclear fusion reactor at the beginning of their maintenance work. It is considered that the heating is the most effective method for decontamination. However, it is important to develop new decontamination method of adsorbed hydro-carbon based substances from the materials that cannot be heated or the inner pipe of double pipes. This report presents results of preliminary experiments performed for the development of the effective tritium decontamination technique pursuing under US/Japan collaborative program on technology for fusion-fuel processing (Annex IV). In the experiments, the effects of Ultra Violet (UV) irradiation on tritium removal from some kinds of materials, such as poly vinyl chloride -(CH 2 CHCl) n - film, polyethylene film and graphite samples coated by C 2 H 2 plasma were examined. As the result of UV irradiation, it was confined that hydrogen and carbon based compounds could be released from the specimen during UV irradiation. It is concluded that UV irradiation is one of the hopeful candidates for effective tritium decontamination. (author)

  16. Information for irradiation and post-irradiation of the silicide fuel element prototype P-07

    International Nuclear Information System (INIS)

    Abbate, Maximo J.; Sbaffoni, Maria M.

    2003-01-01

    Included in the 'Silicides' Project, developed by the Nuclear Fuels Department of the National Atomic Energy Commission (CNEA), it is foreseen the qualification of this type of fuel for research reactors in order to be used in the Argentine RA-3 reactor and to confirm the CNEA as an international supplier. The paper presents basic information on several parameters corresponding to the new silicide prototype, called P-07, to be taken into account for its irradiation, postirradiation and qualification. (author)

  17. Post-Irradiation Non-Destructive Analyses of the AFIP-7 Experiment

    Science.gov (United States)

    Williams, W. J.; Robinson, A. B.; Rabin, B. H.

    2017-12-01

    This article reports the results and interpretation of post-irradiation non-destructive examinations performed on four curved full-size fuel plates that comprise the AFIP-7 experiment. These fuel plates, having a U-10 wt.%Mo monolithic design, were irradiated under moderate operating conditions in the Advanced Test Reactor to assess fuel performance for geometries that are prototypic of research reactor fuel assemblies. Non-destructive examinations include visual examination, neutron radiography, profilometry, and precision gamma scanning. This article evaluates the qualitative and quantitative data taken for each plate, compares corresponding data sets, and presents the results of swelling analyses. These characterization results demonstrate that the fuel meets established irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  19. Fuel performance and operation experience of WWER-440 fuel in improved fuel cycle

    International Nuclear Information System (INIS)

    Gagarinski, A.; Proselkov, V.; Semchenkov, Yu.

    2007-01-01

    The paper summarizes WWER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation (Authors)

  20. Investigation of TIG welding characteristics with a dual cooled rod for the fuel irradiation test

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Kim, Hyung Kyu

    2008-01-01

    To establish the fabrication process, and for satisfying the requirements of the irradiation test, an TIG(Tungsten Inert Gas) welding machine for the dual cooled rods specimens was developed, and the preliminary welding experiments were performed to optimize the welding process conditions. Cladding tubes of 15.9 and 9 mm for the outer and inner diameters, respectively with a 0.57 mm thickness and end caps were used for the specimens. This paper describes the experimental results of the TIG welds and the micrograph examinations of the TIG welded specimens corresponding to various welding conditions for the dual cooled fuel irradiation test. The investigations revealed that the present TIG process satisfied the requirements for the fuel irradiation test in the HANARO research reactor

  1. Analysis of recent fuel-disruption experiments

    International Nuclear Information System (INIS)

    Kramer, J.M.; Kraft, T.E.; DiMelfi, R.J.; Fenske, G.R.; Gruber, E.E.

    1982-01-01

    Recent USDOE-sponsored DEH, FGR, and TREAT F series fuel-disruption experiments are analyzed with existing analytical models. The experiments are interpreted and the results used to evaluate the models. Calculations are presented using the FRAS3 fission-gas-behavior code and the DiMelfi-Deitrich fuel-response model

  2. An analysis of recent fuel disruption experiments

    International Nuclear Information System (INIS)

    Kramer, J.M.; Kraft, T.E.; Dimelfi, R.J.; Fenske, G.R.; Gruber, E.E.

    1982-01-01

    Recent USDOE-Sponsored DEH, FGR, and TREAT F series fuel disruption experiments are analyzed with existing analytical models. The experiments are interpreted and the results used to evaluate the models. Calculations are presented using the FRAS3 fission gas behavior code and the DiMelfi-Deitrich fuel response model

  3. Status on the construction of the fuel irradiation test facility

    International Nuclear Information System (INIS)

    Park, Kook Nam; Sim, Bong Shick; Lee, Chung Young; Yoo, Seong Yeon

    2005-01-01

    As a facility to examine general performance of nuclear fuel under irradiation condition in HANARO, Fuel Test Loop(FTL) has been developed which can accommodate 3 fuel pins at the core irradiation hole(IR1 hole) taking consideration user's test requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of- Pile System (OPS). Test condition in IPS such as pressure, temperature and the water quality, can be controlled by OPS. 3-Pin FTL Conceptual design was set up in 2001 and had completed detail design including a design requirement and basic Piping and Instrument Diagram (P and ID) in 2004. The safety analysis report was prepared and submitted in early 2005 to the regulatory body(KINS) for review and approval of FTL. In 2005, the development team is going to purchase and manufacture hardware and make a contract for construction work. In 2006, the development team is going to install an FTL system performance test shall be done as a part of commissioning. After a 3-Pin FTL development which is expected to be finished by the 2007, FTL will be used for the irradiation test of the new PWR-type fuel and the usage of HANARO will be enhanced

  4. On Cherenkov light production by irradiated nuclear fuel rods

    International Nuclear Information System (INIS)

    Branger, E.; Grape, S.; Svärd, S. Jacobsson; Jansson, P.; Sundén, E. Andersson

    2017-01-01

    Safeguards verification of irradiated nuclear fuel assemblies in wet storage is frequently done by measuring the Cherenkov light in the surrounding water produced due to radioactive decays of fission products in the fuel. This paper accounts for the physical processes behind the Cherenkov light production caused by a single fuel rod in wet storage, and simulations are presented that investigate to what extent various properties of the rod affect the Cherenkov light production. The results show that the fuel properties have a noticeable effect on the Cherenkov light production, and thus that the prediction models for Cherenkov light production which are used in the safeguards verifications could potentially be improved by considering these properties. It is concluded that the dominating source of the Cherenkov light is gamma-ray interactions with electrons in the surrounding water. Electrons created from beta decay may also exit the fuel and produce Cherenkov light, and e.g. Y-90 was identified as a possible contributor to significant levels of the measurable Cherenkov light in long-cooled fuel. The results also show that the cylindrical, elongated fuel rod geometry results in a non-isotropic Cherenkov light production, and the light component parallel to the rod's axis exhibits a dependence on gamma-ray energy that differs from the total intensity, which is of importance since the typical safeguards measurement situation observes the vertical light component. It is also concluded that the radial distributions of the radiation sources in a fuel rod will affect the Cherenkov light production.

  5. Post-irradiation examinations of THERMHET composite fuels for transmutation

    Science.gov (United States)

    Noirot, J.; Desgranges, L.; Chauvin, N.; Georgenthum, V.

    2003-07-01

    The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl 2O 4 spinel inert matrix and around 40% weight of UO 2 to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour.

  6. Post-irradiation examinations of THERMHET composite fuels for transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Noirot, J. E-mail: jnoirot@cea.fr; Desgranges, L.; Chauvin, N.; Georgenthum, V

    2003-07-01

    The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl{sub 2}O{sub 4} spinel inert matrix and around 40% weight of UO{sub 2} to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour.

  7. Post-irradiation examinations of THERMHET composite fuels for transmutation

    International Nuclear Information System (INIS)

    Noirot, J.; Desgranges, L.; Chauvin, N.; Georgenthum, V.

    2003-01-01

    The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl 2 O 4 spinel inert matrix and around 40% weight of UO 2 to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour

  8. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  9. Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation – Non-destructive analysis of the AFIP-1 fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Wachs, D.M., E-mail: daniel.wachs@inl.gov [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Robinson, A.B.; Rice, F.J. [Idaho National Laboratory, Characterization and Advanced PIE Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Kraft, N.C.; Taylor, S.C. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Lillo, M. [Idaho National Laboratory, Nuclear Systems Design and Analysis Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Woolstenhulme, N.; Roth, G.A. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2016-08-01

    Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008–2009. The irradiation conditions were: ∼250 W/cm{sup 2} peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm{sup 3} peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

  10. Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor

    International Nuclear Information System (INIS)

    Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon

    2005-01-01

    A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams

  11. Operating experience with gamma ray irradiators

    International Nuclear Information System (INIS)

    Fraser, F.M.; Ouwerkerk, T.

    1980-01-01

    The experience of Atomic Energy of Canada, Limited (AECL) with radioisotopes dates back to the mid-1940s when radium was marketed for medical purposes. Cobalt-60 came on the scene in 1949 and within a few years a thriving business in cancer teletherapy machines and research irradiators was developed. AECL's first full-scale cobalt-60 gamma ray sterilizer for medical products was installed in 1964. AECL now has over 50 plants and 30 million curies in service around the world. Sixteen years of design experience in cobalt-60 sources, radiation shielding, safety interlock systems, and source pass mechanisms have made gamma irradiators safe, reliable, and easy to operate. This proven technology is being applied in promising new fields such as sludge treatment and food preservation. Cesium-137 is expected to be extensively utilized as the gamma radiation source for these applications

  12. Pickering irradiated fuel transfer conveyor isolation

    Energy Technology Data Exchange (ETDEWEB)

    Koivisto, D J; Eijsermans, L J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1997-12-31

    Pickering A NGS has been in operation for 25 years and is one of the longest in service CANDU stations. Some underwater fuel handling equipment, notably the conveyor stops, have been without maintenance throughout that time. This paper describes the concept of a conveyor isolation system that permits draining of a single or multiple elevator columns and also the early stages of a development program for the elastomeric sealing element. The prototype seal element has been proven in lab tests to be capable of limiting leakage to 0.5 IGPM (imperial gallons per minute) at the design pressure of 6.5 psi. The design of a sealing element is particularly interesting because the conveyor tube is a square cross-section which contains an additional obstruction , a conveyor drive cable. A seal delivery, actuating and positioning system has been conceptually laid out and the design is proceeding, with projected implementation in 1998. (author). 8 figs.

  13. Pickering irradiated fuel transfer conveyor isolation

    International Nuclear Information System (INIS)

    Koivisto, D.J.; Eijsermans, L.J.

    1996-01-01

    Pickering A NGS has been in operation for 25 years and is one of the longest in service CANDU stations. Some underwater fuel handling equipment, notably the conveyor stops, have been without maintenance throughout that time. This paper describes the concept of a conveyor isolation system that permits draining of a single or multiple elevator columns and also the early stages of a development program for the elastomeric sealing element. The prototype seal element has been proven in lab tests to be capable of limiting leakage to 0.5 IGPM (imperial gallons per minute) at the design pressure of 6.5 psi. The design of a sealing element is particularly interesting because the conveyor tube is a square cross-section which contains an additional obstruction , a conveyor drive cable. A seal delivery, actuating and positioning system has been conceptually laid out and the design is proceeding, with projected implementation in 1998. (author). 8 figs

  14. In-pile instrumentation improvements for fuel irradiations in test reactor

    International Nuclear Information System (INIS)

    Blanc, J.Y.; Bernard, J.L.; Estrade, J.; Geoffroy, G.

    1996-01-01

    Knowledge of fuel limits and safety margins in normal and off-normal transients in nuclear power plants remains a constant preoccupation for electricity producers and fuel manufacturers. Accurate determination of such limits, through fuel irradiation testing in the OSIRIS reactor at Saclay is closely linked to the reliability of appropriate instrumentation techniques. Two paths are currently followed to obtain short experimental rods: segmented fuel coming directly from power plants, or re-fabrication of rods in hot cells with our FABRICE process. It can be associated with instrumentation such as fuel centerline thermocouple in annular pellets, pressure transducer or fission gas release measurement by gamma-spectrometry using helium sweeping, in analytic experiments. Our present development, to be implemented in 1993, is the the centerline instrumentation of a fuel column with solid pellets. Inserting the thermocouple requires a cold drilling machine, using CO 2 freezing of broken UO 2 (with liquid nitrogen). During the fuel rod irradiation itself, we try to lower the uncertainties associated to power determination, using thermal balance or neutronic calibration, or even gamma spectrometry. A description of the new test train designed for the ISABELLE water loop in OSIRIS is given, with special emphasis on instrumentation: a LVDT for measuring fuel rod elongation and eventual clad failure, and increased number and better localization of thermocouples and SPDN. The third part is devoted to the measurements by optical microdensitometry of neutron radiographs of the fuel pellet dish modification after irradiation. Dishes are generally disappearing through thermal and mechanical deformation of the pellet, and this can eventually be modelized to better understand pellet-cladding mechanical interaction. (author). 3 refs, 5 figs

  15. Fuel utilization potential in light water reactors with once-through fuel irradiation (AWBA Development Program)

    International Nuclear Information System (INIS)

    Rampolla, D.S.; Conley, G.H.; Candelore, N.R.; Cowell, G.K.; Estes, G.P.; Flanery, B.K.; Duncombe, E.; Dunyak, J.; Satterwhite, D.G.

    1979-07-01

    Current commercial light water reactor cores operate without recylce of fuel, on a once-through fuel cycle. To help conserve the limited nuclear fuel resources, there is interest in increasing the energy yield and, hence, fuel utilization from once-through fuel irradiation. This report evaluates the potential increase in fuel utilization of light water reactor cores operating on a once-through cycle assuming 0.2% enrichment plant tails assay. This evaluation is based on a large number of survey calculations using techniques which were verified by more detailed calculations of several core concepts. It is concluded that the maximum fuel utilization which could be achieved by practical once-through pressurized light water reactor cores with either uranium or thorium is about 17 MWYth/ST U 3 O 8 (Megawatt Years Thermal per Short Ton of U 3 O 8 ). This is about 50% higher than that of current commercial light water reactor cores. Achievement of this increased fuel utilization would require average fuel burnup beyond 50,000 MWD/MT and incorporation of the following design features to reduce parasitic losses of neutrons: reflector blankets to utilize neutrons that would otherwise leak out of the core; fuel management practices in which a smaller fraction of the core is replaced at each refueling; and neutron economic reactivity control, such as movable fuel control rather than soluble boron control. For a hypothetical situation in which all neutron leakage and parasitic losses are eliminated and fuel depletion is not limited by design considerations, a maximum fuel utilization of about 20 MWYth/ST U 3 O 8 is calculated for either uranium or thorium. It is concluded that fuel utilization for comparable reactor designs is better with uranium fuel than with thorium fuel for average fuel depletions of 30,000 to 35,000 MWD/MT which are characteristic of present light water reactor cores

  16. HTCAP-1: a program for calcuating operating temperatures in HFIR target irradiation experiments

    International Nuclear Information System (INIS)

    Kania, M.J.; Howard, A.M.

    1980-06-01

    The thermal modeling code, HTCAP-1, calculates in-reactor operating temperatures of fueled specimens contained in the High Flux Isotope Reactor (HFIR) target irradiation experiments (HT-series). Temperature calculations are made for loose particle and bonded fuel rod specimens. Maximum particle surface temperatures are calculated for the loose particles and centerline and surface temperatures for the fuel rods. Three computational models are employed to determine fission heat generation rates, capsule heat transfer analysis, and specimen temperatures. This report is also intended to be a users' manual, and the application of HTCAP-1 to the HT-34 irradiation capsule is presented

  17. Evolution of fuel rod support under irradiation impact on the mechanical behaviour of fuel assemblies

    International Nuclear Information System (INIS)

    Billerey, Antoine; Waeckel, Nicolas

    2005-01-01

    New fuel management targets imply to increase fuel assembly discharge burnup. Therefore, the prediction of the mechanical behaviour of the irradiated fuel assembly is essential such as excessive fuel assembly distortion induce incomplete Rod Cluster Control Assembly insertion problems (safety issue) or fuel rod vibration induced wear leading to leaking rods (plant operation problems). Within this framework, one of the most important parameter is the knowledge of the fuel rod support in the grid cell because it directly governs the mechanical behaviour of the fuel assembly and consequently allows to predict the behaviour of irradiated structures in terms of (1) axial and lateral deformation (global behaviour of the assembly) and (2) rod vibration induced wear (local behaviour of the rod). Generally, fuel rod support is provided by a spring-dimple system fixed to the grid. During irradiation, the spring force decreases and a gap between the rod and the spring may occur. This phenomenon is due to (1) stress relieving in the spring and in the dimples, (2) grid growth and (3) reduction of the rod diameter. Two models have been developed to predict the behaviour of the rod in the cell. The first model is dedicated to the evaluation of the spring force relaxation during irradiation. The second one can assess the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (1) the creep laws of the grid materials, (2) the growth law of the grid, (3) the evolution of rod diameter and (4) the design of the fuel rod support. The aim of this paper is to: (1) evaluate the consequences of grid support design modifications on the rod vibration sensitivity in terms of predicted rod to grid maximum gap during irradiation and time in operation with an open rod to grid gap, (2) evaluate, using a linear or non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the overall mechanical

  18. Method for monitoring irradiated nuclear fuel using cerenkov radiation

    International Nuclear Information System (INIS)

    Caldwell, J.T.; Dowdy, E.J.; Nicholson, N.

    1983-01-01

    A method is provided for monitoring irradiated nuclear fuel inventories located in a water-filled storage pond wherein the intensity of the cerenkov radiation emitted from the water in the vicinity of the nuclear fuel is measured. This intensity is then compared with the expected intensity for nuclear fuel having a corresponding degree of irradiation exposure and time period after removal from a reactor core. Where the nuclear fuel inventory is located in an assembly having fuel pins or rods with intervening voids, the cerenkov light intensity measurement is taken at selected bright spots corresponding to the water-filled interstices of the assembly in the water storage, the waterfilled interstices acting as cerenkov light channels so as to reduce cross-talk. On-line digital analysis of an analog video signal is possible, or video tapes may be used for later measurement using a video editor and an electrometer. Direct measurement of the cerenkov radiation intensity also is possible using spot photometers pointed at the assembly

  19. Microbial biofilm growth on irradiated, spent nuclear fuel cladding

    International Nuclear Information System (INIS)

    Bruhn, D.F.; Frank, S.M.; Roberto, F.F.; Pinhero, P.J.; Johnson, S.G.

    2009-01-01

    A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 x 10 3 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments

  20. LOFT instrumented fuel design and operating experience

    International Nuclear Information System (INIS)

    Russell, M.L.

    1979-01-01

    A summary description of the Loss-of-Fluid Test (LOFT) system instrumented core construction details and operating experience through reactor startup and loss-of-coolant experiment (LOCE) operations performed to date are discussed. The discussion includes details of the test instrumentation attachment to the fuel assembly, the structural response of the fuel modules to the forces generated by a double-ended break of a pressurized water reactor (PWR) coolant pipe at the inlet to the reactor vessel, the durability of the LOFT fuel and test instrumentation, and the plans for incorporation of improved fuel assembly test instrumentation features in the LOFT core

  1. Multi-physic simulations of irradiation experiments in a technological irradiation reactor; Modelisation pluridisciplinaire d'experiences d'irradiation dans un reacteur d'irradiation technologique

    Energy Technology Data Exchange (ETDEWEB)

    Bonaccorsi, Th

    2007-09-15

    A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

  2. Fuel-disruption experiments under high-ramp-rate heating conditions

    International Nuclear Information System (INIS)

    Wright, S.A.; Worledge, D.H.; Cano, G.L.; Mast, P.K.; Briscoe, F.

    1983-10-01

    This topical report presents the preliminary results and analysis of the High Ramp Rate fuel-disruption experiment series. These experiments were performed in the Annular Core Research Reactor at Sandia National Laboratories to investigate the timing and mode of fuel disruption during the prompt-burst phase of a loss-of-flow accident. High-speed cinematography was used to observe the timing and mode of the fuel disruption in a stack of five fuel pellets. Of the four experiments discussed, one used fresh mixed-oxide fuel, and three used irradiated mixed-oxide fuel. Analysis of the experiments indicates that in all cases, the observed disruption occurred well before fuel-vapor pressure was high enough to cause the disruption. The disruption appeared as a rapid spray-like expansion and occurred near the onset of fuel melting in the irradiated-fuel experiments and near the time of complete fuel melting in the fresh-fuel experiment. This early occurrence of fuel disruption is significant because it can potentially lower the work-energy release resulting from a prompt-burst disassembly accident

  3. Experimental fuel channel for samples irradiation at the RB reactor

    International Nuclear Information System (INIS)

    Pesic, M.; Markovic, H.; Sokcic-Kostic, M.; Miric, I.; Prokic, M.; Strugar, P.

    1984-12-01

    An 80% enriched UO 2 fuel channel at the RB nuclear reactor in the 'Boris Kidric' Institute of Nuclear Sciences is modified for samples irradiation by fast neutrons. Maximum sample diameter is 25 mm and length up to 1000 mm. Characteristics of neutron and gamma radiation fields of this new experimental channel are investigated. In the centre of the channel, the main contribution to the total neutron absorbed dose, i.e. 0.29 Gy/Wh of reactor operation, is due to the fast neutron spectrum component. Only 0.05 Gy and 0.07 Gy in the total neutron absorbed dose are due to intermediate and thermal neutrons, respectively. At the same time the gamma absorbed dose is 0.35 Gy. The developed experimental fuel channel, EFC, has wide possibilities for utilization, from fast neutron spectrum studies, electronic component irradiations, dosemeters testing, up to cross-section measurements. (author)

  4. Irradiation behavior of uranium-silicide dispersion fuels

    International Nuclear Information System (INIS)

    Hofman, G.L.; Neimark, L.A.

    1984-01-01

    This paper describes and analyzes the irradiation behavior of experimental fuel plates containing U 3 Si, U 3 Si-1.5 w/o Al, and U 3 Si 2 particulate fuel dispersed and clad in aluminum. The fuel is nominally 19.9%-enriched 235 U and the fuel volume fraction in the central ''meat'' section of the plates is approximately 33%. Sets of fuel plates were removed from the Oak Ridge Research reactor at burnup levels of 35, 83, and 94% 235 U depletion and examined at the Alpha-Gamma Hot-Cell Facility at Argonne National Laboratory. The results of the examination may be summarized as follows. The dimensional stability of the U 3 Si 2 and pure U 3 Si fuel was excellent throughout the entire burnup range, with uniform plate thickness increases up to a maximum of 4 mils at the highest burnup level (94% 235 U depletion). This corresponds to a meat volume increase of 11%. The swelling was partially due to solid fission products but to a larger extent to fission gas bubbles. The fission gas bubbles in U 3 Si 2 were small (submicrometer size) and very uniformly distributed, indicating great stability. To a large extent this was also the case for U 3 Si; however, larger bubbles ( 3 Si-1.5 w/o Al fuel became unstable at the higher burnup levels. Fission gas bubbles were larger than in the other two fuels and were present throughout the fuel particles. At 94% 235 U depletion, the formation of fission gas bubbles with diameters up to 20 mils caused the plates to pillow. It is proposed that aluminum in U 3 Si destabilizes fission gas bubble formation to the point of severe breakaway swelling in the prealloyed silicide fuel. (author)

  5. Equipment for detach the fuel elements of the irradiated candu fuel bundle

    International Nuclear Information System (INIS)

    Cojocaru, V.; Dinuta, G.

    2013-01-01

    Monitoring the behaviour of the fuel bundles during their combustion provides useful information for the operation of the nuclear power plant as well as for the fuel manufacturer. Before placing it inside the reactor, the fuel bundle is inspected visually, dimensionally and, during combustion in the reactor, its radioactive behaviour is monitored. The purpose of the presented equipment is to allow the visual external inspection of the damaged fuel bundle in order to identify visible defects and to detach the fuel element by breaking the welded connection between the cap and grid. These devices are operated using the handler devices already existing in the hot cells Post-Irradiation Examination Laboratory (LEPI). This equipment has been used successfully in the LEPI laboratory at SCN Pitesti to inspect the damaged fuel from Cernavoda NPP, in March 2013. (authors)

  6. Post-irradiation examination of U3SIX-AL fuel element manufactured and irradiated in Argentina

    International Nuclear Information System (INIS)

    Ruggirello, Gabriel; Calabroni, Hector; Sanchez, Miguel; Hofman, Gerard

    2002-01-01

    As a part of CNEA's qualification program as a supplier of low enriched Al-U 3 Si 2 dispersion fuel elements for research reactors, a post irradiation examination (PIE) of the first prototype of this kind, called P-04, manufactured and irradiated in Argentina, was carried out. The main purpose of this work was to set up various standard PIE techniques in the hot cell, looking forward to the next steps of the qualification program, as well as to acquire experience on the behaviour of this nuclear material and on the control of the manufacturing process. After an appropriate cooling period, on May 2000 the P-04 was transported to the hot cell in Ezeiza Atomic Centre. Non destructive and destructive tests were performed following the PIE procedures developed in Argonne National Laboratory (ANL), this mainly included dimensional measurement, microstructural observations and chemical burn-up analyses. The methodology and results of which are outlined in this report. The results obtained show a behaviour consistent with that of other fuel elements of the same kind, tested previously. On the other hand the results of this PIE, specially those concerning burn-up analysis and stability and corrosion behaviour of the fuel plates, will be of use for the IAEA Regional Program on the characterization of MTR spent fuel. (author)

  7. American proposals for long range storage of irradiated fuel

    International Nuclear Information System (INIS)

    Sugier, Annie

    1978-01-01

    The American politics of irradiated fuel management is reviewed, the short-range storage of huge amounts of wastes being the fundamental problem. Two steps are considered: the ''At the Reactor'' storage, ensured by the electricity companies, and the ''Away From Reactor'' storage on the DOE's responsibility. A technical and economical study has been carried out in order to estimate the cost of the AFR provisory storage and a project of taxation has been established on this basis [fr

  8. The permission of transport of irradiated nuclear fuel elements

    International Nuclear Information System (INIS)

    Klomberg, T.J.M.

    2000-01-01

    In July and October 2000 the Dutch government granted permits for the transportation of irradiated nuclear fuel elements. The environmental organization Greenpeace objected against the permit, but that was rejected by the Dutch Council of State. A brief overview is given of the judgements and the state-of-the-art with respect to the transportation of the elements from Dutch reactors and storage facilities in Petten, Dodewaard and Borssele to Cogema in La Hague, France and BNFL in Sellafield, England

  9. American proposals for long range storage of irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sugier, A [CEA, 75 - Paris (France). Dept. des Programmes

    1978-12-01

    The American politics of irradiated fuel management is reviewed, the short-range storage of huge amounts of wastes being the fundamental problem. Two steps are considered: the ''At the Reactor'' storage, ensured by the electricity companies, and the ''Away From Reactor'' storage on the DOE's responsibility. A technical and economical study has been carried out in order to estimate the cost of the AFR provisory storage and a project of taxation has been established on this basis.

  10. Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

    International Nuclear Information System (INIS)

    Wachs, G.W.

    1997-11-01

    The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

  11. Post irradiation examinations of uranium-plutonium mixed carbide fuels irradiated at low linear power rate

    International Nuclear Information System (INIS)

    Maeda, Atsushi; Sasayama, Tatsuo; Iwai, Takashi; Aizawa, Sakuei; Ohwada, Isao; Aizawa, Masao; Ohmichi, Toshihiko; Handa, Muneo

    1988-11-01

    Two pins containing uranium-plutonium carbide fuels which are different in stoichiometry, i.e. (U,Pu)C 1.0 and (U,Pu)C 1.1 , were constructed into a capsule, ICF-37H, and were irradiated in JRR-2 up to 1.0 at % burnup at the linear heat rate of 420 W/cm. After being cooled for about one year, the irradiated capsule was transferred to the Reactor Fuel Examination Facility where the non-destructive examinations of the fuel pins in the β-γ cells and the destructive ones in two α-γ inert gas atmosphere cells were carried out. The release rates of fission gas were low enough, 0.44 % from (U,Pu)C 1.0 fuel pin and 0.09% from (U,Pu)C 1.1 fuel pin, which is reasonable because of the low central temperature of fuel pellets, about 1000 deg C and is estimated that the release is mainly governed by recoil and knock-out mechanisms. Volume swelling of the fuels was observed to be in the range of 1.3 ∼ 1.6 % for carbide fuels below 1000 deg C. Respective open porosities of (U,Pu)C 1.0 and (U,Pu)C 1.1 fuel were 1.3 % and 0.45 %, being in accordance with the release behavior of fission gas. Metallographic observation of the radial sections of pellets showed the increase of pore size and crystal grain size in the center and middle region of (U,Pu)C 1.0 pellets. The chemical interaction between fuel pellets and claddings in the carbide fuels is the penetration of carbon in the fuels to stainless steel tubes. The depth of corrosion layer in inner sides of cladding tubes ranged 10 ∼ 15 μm in the (U,Pu)C 1.0 fuel and 15 #approx #25 μm in the (U,Pu)C 1.1 fuel, which is correlative with the carbon potential of fuels posibly affecting the amount of carbon penetration. (author)

  12. Technique of manufacturing specimen of irradiated fuel rods

    International Nuclear Information System (INIS)

    Min, Duck Seok; Seo, Hang Seok; Min, Duck Kee; Koo, Dae Seo; Lee, Eun Pyo; Yang, Song Yeol

    1999-04-01

    Technique of manufacturing specimen of irradiated fuel rods to perform efficient PIE is developed by analyzing the relation between requiring time of manufacturing specimen and manufacturing method in irradiated fuel rods. It takes within an hour to grind 1 mm of specimen thickness under 150 rpm in speed of grinding, 600 g gravity in force using no.120, no.240, no.320 of grinding paper. In case of no.400 of grinding paper, it takes more an hour to grind the same thickness as above. It takes up to a quarter to grind 80-130 μm in specimen thickness using no.400 of grinding paper. When grinding time goes beyond 15 minutes, the grinding thickness of specimen does not exist. The polishing of specimen with 150 Rpms in speed of grinding machine, 600 g gravity in force, 10 minutes in polishing time using diamond paste 15 μm on polishing cloths amounts to 50 μm in specimen thickness. In case of diamond paste 9 μm on polishing cloth, the polishing of specimen amounts to 20 μm. The polishing thickness of specimen with 15 minutes in polishing time using 6 μm, 3 μm, 1 μm, 1/4 μm does not exist. Technique of manufacturing specimen of irradiated fuel rods will have application to the destructive examination of PIE. (author). 6 refs., 1 tab., 10 figs

  13. Development of examination technique for oxide layer thickness measurement of irradiated fuel rods

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, S. W.; Kim, J. H.; Seo, H. S.; Min, D. K.; Kim, E. K.; Chun, Y. B.; Bang, K. S.

    1999-06-01

    Technique for oxide layer thickness measurement of irradiated fuel rods was developed to measure oxide layer thickness and study characteristic of fuel rods. Oxide layer thickness of irradiated fuels were measured, analyzed. Outer oxide layer thickness of 3 cycle-irradiated fuel rods were 20 - 30 μm, inner oxide layer thickness 0 - 10 μm and inner oxide layer thickness on cracked cladding about 30 μm. Oxide layer thickness of 4 cycle-irradiated fuel rods were about 2 times as thick as those of 1 cycle-irradiated fuel rods. Oxide layer on lower region of irradiated fuel rods was thin and oxide layer from lower region to upper region indicated gradual increase in thickness. Oxide layer thickness from 2500 to 3000 mm showed maximum and oxide layer thickness from 3000 to top region of irradiated fuel rods showed decreasing trend. Inner oxide layer thicknesses of 4 cycle-irradiated fuel rod were about 8 μm at 750 - 3500 mm from the bottom end of fuel rod. Outer oxide layer thickness were about 8 μm at 750 - 1000 mm from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel. Oxide layer thickness technique will apply safety evaluation and study of reactor fuels. (author). 6 refs., 14 figs

  14. Results of Am isotopic ratio analysis in irradiated MOX fuels

    Energy Technology Data Exchange (ETDEWEB)

    Koyama, Shin-ichi; Osaka, Masahiko; Mitsugashira, Toshiaki; Konno, Koichi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Kajitani, Mikio

    1997-04-01

    For analysis of a small quantity of americium, it is necessary to separate from curium which has similar chemical property. As a chemical separation method for americium and curium, the oxidation of americium with pentavalent bismuth and subsequent co-precipitation of trivalent curium with BIP O{sub 4} were applied to analyze americium in irradiated MOX fuels which contained about 30wt% plutonium and 0.9wt% {sup 241}Am before irradiation and were irradiated up to 26.2GWd/t in the experimental fast reactor Joyo. The purpose of this study is to measure isotopic ratio of americium and to evaluate the change of isotopic ratio with irradiation. Following results are obtained in this study. (1) The isotopic ratio of americium ({sup 241}Am, {sup 242m}Am and {sup 243}Am) can be analyzed in the MOX fuels by isolating americium. The isotopic ratio of {sup 242m}Am and {sup 243}Am increases up to 0.62at% and 0.82at% at maximum burnup, respectively, (2) The results of isotopic analysis indicates that the contents of {sup 241}Am decreases, whereas {sup 242m}Am, {sup 243}Am increase linearly with increasing burnup. (author)

  15. Nuclear spent fuel management. Experience and options

    International Nuclear Information System (INIS)

    1986-01-01

    Spent nuclear fuel can be stored safely for long periods at relatively low cost, but some form of permanent disposal will eventually be necessary. This report examines the options for spent fuel management, explores the future prospects for each stage of the back-end of the fuel cycle and provides a thorough review of past experience and the technical status of the alternatives. Current policies and practices in twelve OECD countries are surveyed

  16. Evolution of fuel rod support under irradiation consequences on the mechanical behavior of fuel assembly

    International Nuclear Information System (INIS)

    Billerey, A.; Bouffioux, P.

    2002-01-01

    The complete paper follows. According to the fuel management policy in French PWR with respect to high burn-up, the prediction of the mechanical behavior of the irradiated fuel assembly is required as far as excessive deformations of fuel assembly might lead to incomplete Rod Cluster Control Assembly insertion (safety problems) and fretting wear lead to leaking rods (plant operation problems). One of the most important parameter is the evolution of the fuel rod support in the grid cell as it directly governs the mechanical behavior of the fuel assembly and consequently allows to predict the behavior of irradiated structure in terms of (i) axial and lateral deformation (global behavior of the assembly) and (ii) fretting wear (local behavior of the rod). Fuel rod support is provided by a spring-dimple system fixed on the grid. During irradiation, the spring force decreases and a gap between the rod and the spring might open. This phenomenon is due to (i) irradiation-induced stress relaxation for the spring and for the dimples, (ii) grid growth and (iii) reduction of rod diameter. Two models have been developed to predict the behavior of the rod in the grid cell. The first model is able to evaluate the spring force relaxation during irradiation. The second one is able to evaluate the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (i) the creep laws of the grid materials, (ii) the growth law of the grid, (iii) the evolution of rod diameter and (iv) the design of the fuel rod support. The objectives of this paper are to: (i) evaluate the consequences of grid support design modifications on the fretting sensitivity in terms of predicted maximum gap during irradiation and operational time to gap appearance; (ii) evaluate, using a non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the mechanical behavior of the full assembly in terms of axial and

  17. Critical experiments with mixed oxide fuel

    International Nuclear Information System (INIS)

    Harris, D.R.

    1997-01-01

    This paper very briefly outlines technical considerations in performing critical experiments on weapons-grade plutonium mixed oxide fuel assemblies. The experiments proposed would use weapons-grade plutonium and Er 2 O 3 at various dissolved boron levels, and for specific fuel assemblies such as the ABBCE fuel assembly with five large water holes. Technical considerations described include the core, the measurements, safety, security, radiological matters, and licensing. It is concluded that the experiments are feasible at the Rensselaer Polytechnic Institute Reactor Critical Facility. 9 refs

  18. Designing the KNK II-TOAST irradiation experiment with the saturn-FS code

    International Nuclear Information System (INIS)

    Ritzhaupt-Kleissl, H.J.; Elbel, H.; Heck, M.

    1991-01-01

    In order to study the existing specification of FBR fuel with respect to allowable fabrication tolerances with the objective to reduce the expense of fabrication and quality control, the TOAST irradiation experiment will be carried out in the 3 rd core of the KNK II. This experiment shall investigate the influence of the following fuel specification parameters on the operational behaviour: - Fuel diameter - Stoichiometry - Sintering atmosphere - Fill gas in the fuel pin. The combination of these test parameters led to a fabrication of 6 types of fuel pellets, giving together with two fill gas mixtures a total of 9 fuel pin types. Design calculations in the frame of the standard licensing procedure have been performed with the SATURN-FS fuel pin behaviour code. These calculations have been done for the steady-state behaviour as well as for some defined design transients, such as startup procedures and overpower ramps

  19. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    International Nuclear Information System (INIS)

    Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

    1977-01-01

    Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

  20. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

    1977-04-01

    Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

  1. Test requirement for PIE of HANARO irradiated fuel rod

    International Nuclear Information System (INIS)

    Lim, I. C.; Cho, Y. G.

    2000-06-01

    Since the first criticality of HANARO reached in Feb. of 1995, the rod type U 3 Si-A1 fuel imported from AECL has been used. From the under-water fuel inspection which has been conducted since 1997, a ballooning-rupture type abnormality was observed in several fuel rods. In order to find the root cause of this abnormality and to find the resolution, the post irradiation examination(PIE) was proposed as the best way. In this document, the information from the under-water inspection as well as the PIE requirements are described. Based on the information in this document, a detail test plan will be developed by the project team who shall conduct the PIE

  2. The growth of intra-granular bubbles in post-irradiation annealed UO2 fuel

    International Nuclear Information System (INIS)

    White, R.J.

    2001-01-01

    Post-irradiation examinations of low temperature irradiated UO 2 reveal large numbers of very small intra-granular bubbles, typically of around 1 nm diameter. During high temperature reactor transients these bubbles act as sinks for fission gas atoms and vacancies and can give rise to large volumetric swellings, sometimes of the order of 10%. Under irradiation conditions, the nucleation and growth of these bubbles is determined by a balance between irradiation-induced nucleation, diffusional growth and an irradiation induced re-solution mechanism. This conceptual picture is, however, incomplete because in the absence of irradiation the model predicts that the bubble population present from the pre-irradiation would act as the dominant sink for fission gas atoms resulting in large intra-granular swellings and little or no fission gas release. In practice, large fission gas releases are observed from post-irradiation annealed fuel. A recent series of experiments addressed the issue of fission gas release and swelling in post-irradiation annealed UO 2 originating from Advanced Gas Cooled Reactor (AGR) fuel which had been ramp tested in the Halden Test reactor. Specimens of fuel were subjected to transient heating at ramp rates of 0.5 deg. C/s and 20 deg. C/s to target temperatures between 1600 deg. C and 1900 deg. C. The release of fission gas was monitored during the tests. Subsequently, the fuel was subjected to post-irradiation examination involving detailed Scanning Electron Microscopy (SEM) analysis. Bubble-size distributions were obtained from seventeen specimens, which entailed the measurement of nearly 26,000 intra-granular bubbles. The analysis reveals that the bubble densities remain approximately invariant during the anneals and the bubble-size distributions exhibit long exponential tails in which the largest bubbles are present in concentrations of 10 4 or 10 5 lower than the concentrations of the average sized bubbles. Detailed modelling of the bubble

  3. Method for the chemical reprocessing of irradiated nuclear fuels, in particular nuclear fuels containing uranium

    International Nuclear Information System (INIS)

    Koch, G.

    1976-01-01

    In the chemical processing of irradiated uranium-containing nuclear fuels which are hydrolyzed with aqueous nitric acid, a suggestion is made to use as quaternary ammonium nitrate trialkyl-methyl ammonium nitrates as extracting agent, in which the sum of C atoms is greater than 16. In the illustrated examples, tricaprylmethylammonium nitrate, trilaurylmethylammonium nitrate and tridecylmethylammonium nitrate are named. (HPH/LH) [de

  4. R and D for back-end options for irradiated research reactor fuel in Germany

    International Nuclear Information System (INIS)

    Bruecher, H.; Curtius, H.; Fachinger, J.

    2001-01-01

    Out of 11.5 t of irradiated fuel arising from German research reactors until the end of this decade, 3.9 t are intended to be returned to the USA, and 2.3 t are expected to be recycled for reuse of uranium. The remaining 5.3 t, as well as the fuel irradiated after the year 2010, will have to follow the domestic back-end option of extended dry interim storage in Castor-type casks, followed by disposal in a deep geological repository. R and D is going on in the Research Centre Juelich to investigate the long-term behaviour of U-Al based fuel in a salt repository. First results from leaching experiments show I) a fast dissolution of the fuel with mobilization of its radionuclide inventory, and 2) the following formation of amorphous Al-Mg-hydroxide phases. Long-lived actinides from the fuel were shown to be fixed in these phases and hence immobilized. Future R and D will be to investigate the nature and stability of these phases for long-term safety assessments. Investigations will have to be extended to cover alternative disposal sites (granite clay) as well as different (e.g. silicon based) fuels. (author)

  5. Design and manufacturing of 05F-01K instrumented capsule for nuclear fuel irradiation in Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, J. M.; Shin, Y. T.; Park, S. J. (and others)

    2007-07-15

    An instrumented capsule was developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel pellet elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in Hanaro. The instrumented capsule(02F-11K) for measuring and monitoring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. It was successfully irradiated in the test hole OR5 of Hanaro from March 14, 2003 to June 1, 2003 (53.84 full power days at 24 MW). In the year of 2004, 3 test fuel rods and the instrumented capsule(03F-05K) were designed and manufactured to measure fuel centerline temperature, internal pressure of fuel rod, and fuel axial deformation during irradiation test. This capsule was irradiated in the test hole OR5 of Hanaro reactor from April 26, 2004 to October 1, 2004 (59.5 EFPD at 24 {approx} 30 MW). The six typed dual instrumented fuel rods, which allow for two characteristics to be measured simultaneously in one fuel rod, have been designed and manufactured to enhance the efficiency of the irradiation test using the instrumented fuel capsule. The 05F-01K instrumented fuel capsule was designed and manufactured for a design verification test of the three dual instrumented fuel rods. The irradiation test of the 05F-01K instrumented fuel capsule will be carried out at the OR5 vertical experimental hole of Hanaro.

  6. Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

    Energy Technology Data Exchange (ETDEWEB)

    Boussard, F.; Huillery, R. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Combustibles; Averseng, J.L.; Serpantie, J.P. [Novatome Industries, 92 - Le Plessis-Robinson (France)

    1994-12-31

    This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs.

  7. Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

    International Nuclear Information System (INIS)

    Boussard, F.; Huillery, R.

    1994-01-01

    This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs

  8. Irradiation performance updates on Korean advanced fuels for PWRs

    International Nuclear Information System (INIS)

    Jang, Y.K.; Jeon, K.L.; Kim, Y.H.; Yoo, J.S.; Kim, J.I.; Shin, J.C.; Chung, J.G.; Park, J.R.; Chung, S.K.; Kim, T.W.; Yoon, Y.B.; Park, K.M.; Yoo, M.J.; Kim, M.S.; Lee, T.H.

    2010-01-01

    The developments of advanced nuclear fuels for PWRs were started in 1999 and in 2001, respectively: PLUS7 TM for eight operating optimized power reactors of 1000 MWe class (OPR1000) and four advanced power reactors of 1400 MWe class (APR1400) under construction, and 16ACE7 TM and 17ACE7 TM for an operating 16x16 Westinghouse type plant and six operating 17x17 Westinghouse type plants. The design targets were as follows: batch average burnup up to 55 GWD/MTU, over 10% thermal margin increase, improvement of the mechanical integrity of higher seismic capability, higher debris or grid fretting wear performance, higher control rod insertion capability, increase of neutron economy, improvement of manufacturability, solving incomplete rod insertion (IRI) issue and top nozzle screw failure issue, etc. in comparison of the existing nuclear fuels. The irradiation tests using each four LTAs (Lead Test Assemblies) during 3 cycles were completed in three Korean nuclear reactors until 2009. The eight irradiation performance items which are assembly growth, rod growth, grid width growth, assembly bow, rod bow, assembly twist, rod diameter and cladding oxidation were examined in pool-side after each cycle and evaluated. The irradiation tests could be continued by expecting the good performances for next cycle from the previous cycle. After 2 cycle irradiations, the region implementation could be started in 15 nuclear power plants. Even though the verifications using the LTAs were completed, each surveillance program was launched and the irradiation performance data were being updated during region implementation. In addition to pool-side examinations (PSEs) by assembly-wise during irradiation tests, six rod-wise performance items were also examined in pool-side using each LTA after discharge. All performance items met their design criteria as a result of the evaluation. Even though the interesting ones among the irradiation performance parameters were assembly and grid growths

  9. Data sheets of fission product release experiments for light water reactor fuel, (2)

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo; Yamamoto, Katsumune; Nakazaki, Chozaburo.

    1979-07-01

    This is the second data sheets of fission products (FP) release experiments for light water reactor fuel. Results of five FP release experiments from the third to the seventh are presented: results of pre-examinations of UO 2 pellets, photographs of parts of fuel rod assemblies for irradiation and the assemblies, operational conditions of JMTR and OWL-1, variations of radioiodine-131 level in the main loop coolant during experimental periods, and representative results of post-irradiation examinations of respective fuel rods. (author)

  10. The irradiation performance of austenitic stainless steel clade PWR fuel rods

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The steady state irradiation performance of austenitic stainless steel clad pressurized water reactor fuel rods is modeled with fuel performance codes of the FRAP series. These codes, originally developed to model the thermal-mechanical behavior of zircaloy clad fuel rods, are modified to model stainless steel clad fuel rods. The irradiation thermal-mechanical behavior of type 348 stainless steel and zircaloy fuel rods is compared. (author) [pt

  11. Fission gas release during post irradiation annealing of large grain size fuels from Hinkley point B

    International Nuclear Information System (INIS)

    Killeen, J.C.

    1997-01-01

    A series of post-irradiation anneals has been carried out on fuel taken from an experimental stringer from Hinkley Point B AGR. The stringer was part of an experimental programme in the reactor to study the effect of large grain size fuel. Three differing fuel types were present in separate pins in the stringer. One variant of large grain size fuel had been prepared by using an MgO dopant during fuel manufactured, a second by high temperature sintering of standard fuel and the third was a reference, 12μm grain size fuel. Both large grain size variants had similar grain sizes around 35μm. The present experiments took fuel samples from highly rated pins from the stringer with local burn-up in excess of 25GWd/tU and annealed these to temperature of up to 1535 deg. C under reducing conditions to allow a comparison of fission gas behaviour at high release levels. The results demonstrate the beneficial effect of large grain size on release rate of 85 Kr following interlinkage. At low temperatures and release rates there was no difference between the fuel types, but at temperatures in excess of 1400 deg. C the release rate was found to be inversely dependent on the fuel grain size. The experiments showed some differences between the doped and undoped large grains size fuel in that the former became interlinked at a lower temperature, releasing fission gas at an increased rate at this temperature. At higher temperatures the grain size effect was dominant. The temperature dependence for fission gas release was determined over a narrow range of temperature and found to be similar for all three types and for both pre-interlinkage and post-interlinkage releases, the difference between the release rates is then seen to be controlled by grain size. (author). 4 refs, 7 figs, 3 tabs

  12. Fission gas release during post irradiation annealing of large grain size fuels from Hinkley point B

    Energy Technology Data Exchange (ETDEWEB)

    Killeen, J C [Nuclear Electric plc, Barnwood (United Kingdom)

    1997-08-01

    A series of post-irradiation anneals has been carried out on fuel taken from an experimental stringer from Hinkley Point B AGR. The stringer was part of an experimental programme in the reactor to study the effect of large grain size fuel. Three differing fuel types were present in separate pins in the stringer. One variant of large grain size fuel had been prepared by using an MgO dopant during fuel manufactured, a second by high temperature sintering of standard fuel and the third was a reference, 12{mu}m grain size fuel. Both large grain size variants had similar grain sizes around 35{mu}m. The present experiments took fuel samples from highly rated pins from the stringer with local burn-up in excess of 25GWd/tU and annealed these to temperature of up to 1535 deg. C under reducing conditions to allow a comparison of fission gas behaviour at high release levels. The results demonstrate the beneficial effect of large grain size on release rate of {sup 85}Kr following interlinkage. At low temperatures and release rates there was no difference between the fuel types, but at temperatures in excess of 1400 deg. C the release rate was found to be inversely dependent on the fuel grain size. The experiments showed some differences between the doped and undoped large grains size fuel in that the former became interlinked at a lower temperature, releasing fission gas at an increased rate at this temperature. At higher temperatures the grain size effect was dominant. The temperature dependence for fission gas release was determined over a narrow range of temperature and found to be similar for all three types and for both pre-interlinkage and post-interlinkage releases, the difference between the release rates is then seen to be controlled by grain size. (author). 4 refs, 7 figs, 3 tabs.

  13. International experience in conditioning spent fuel elements

    International Nuclear Information System (INIS)

    Ashton, P.

    1991-04-01

    The purpose of this report is to compile and present in a clear form international experience (USA, Canada, Sweden, FRG, UK, Japan, Switzerland) gained to date in conditioning spent fuel elements. The term conditioning is here taken to mean the handling and packaging of spent fuel elements for short- or long-term storage or final disposal. Plants of a varying nature fall within this scope, both in terms of the type of fuel element treated and the plant purpose eg. experimental or production plant. Emphasis is given to plants which bear some similarity to the concept developed in Germany for direct disposal of spent fuel elements. Worldwide, however, relatively few conditioning plants are in existence or have been conceived. Hence additional plants have been included where aspects of the experience gained are also of relevance eg. plants developed for the consolidation of spent fuel elements. (orig./HP) [de

  14. Technical Meeting on Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels. Presentations

    International Nuclear Information System (INIS)

    2011-01-01

    The purpose of this meeting was to enable a rationalization and advancement of the design and manufacturing processes, a better selection of promising fuels, and a reduction of the time and costs currently required for R and D and testing, as well as to contribute to the improvement of the safety features of fuels under all operational states and accidental conditions. An overview of the status and perspective of the design, manufacturing and irradiation behaviour of fast reactors fuels were provided during this meeting. The main objectives are the following: Ensure sharing and dissemination of knowledge and expertise; Discuss specific features and issues of existing fuels; Improve knowledge and data for the design and engineering of fast reactor fuel and core structural materials; Discuss perspectives on advanced fuels; Consider modern technological, design and testing tools enabling reliable performance of fuels in current and planned operational environments; Establish international consensus in the developmental efforts on advanced fast reactor technologies, including collaborative programs and experiments. Contribute to the preparation and outline of the planned IAEA Coordinated Research Project on 'Examination of advanced fast reactor fuel and core structural materials. Each of the 24 presentations made at the meeting have been indexed separately

  15. Development, irradiation testing and PIE of UMo fuel at AECL

    International Nuclear Information System (INIS)

    Sears, D.F.

    2005-01-01

    This paper reviews recent U-Mo dispersion fuel development, irradiation testing and postirradiation examination (PIE) activities at AECL. Low-enriched uranium fuel alloys and powders have been fabricated at Chalk River Labs, with compositions ranging from U-7Mo to U-10Mo. The bulk alloys and powders were characterized using optical and scanning electron microscopy, chemical analysis, X-ray diffraction and neutron diffraction analysis. The analyses confirmed that the powders were of high quality, and in the desired gamma phase. Subsequently, kilogram quantities of DU-Mo and LEU-Mo powder have been manufactured for commercial customers. Mini-elements have been fabricated with LEU-7Mo and LEU-10Mo dispersed in aluminum, with a nominal loading of 4.5 gU/cm 3 . These have been irradiated in the NRU reactor at linear powers up to 100 kW/m. The mini-elements achieved 60 atom% 235 U burnup in 2004 March, and the irradiation is continuing to a planned discharge burnup of 80 atom% 235 U. Interim PIE has been conducted on mini-elements that were removed after 20 atom% 235 U burnup. The PIE results are presented in this paper. (author)

  16. Microstructure and elemental distribution of americium containing MOX fuel under the short term irradiation tests

    International Nuclear Information System (INIS)

    Tanaka, Kosuke; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shin Ichi; Yoshimochi, Hiroshi; Tanaka, Kenya

    2008-01-01

    In order to investigate the effect of americium addition to MOX fuels on the irradiation behavior, the 'Am-1' program is being conducted in JAEA. The Am-1 program consists of two short term irradiation tests of 10-minute and 24 hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. The PIEs for Am-containing MOX fuels focused on the microstructural evolution and redistribution behavior of Am at the initial stage of irradiation and the results to date are reported

  17. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1988-04-01

    This circular has been prepared in response to numerous requests for information regarding routes for the shipment of irradiated reactor (spent) fuel subject to regulation by the Nuclear Regulatory Commission (NRC). The NRC staff approves such routes prior to their use, in accordance with the regulatory provisions of 10 CFR Part 73.37. The objective of the safeguards regulations contained in 10 CFR Part 73.37 is to provide protection against radioactive dispersal caused by malevolent acts by persons. The design and construction of the casks used to ship the spent fuel provide adequate radiological protection of the public health and safety against accidents. Therfore, transporting appropriately packaged spent fuel over existing rail systems and via any highway system is radiologically safe without specific NRC approval of the route. However, to assure adequate planning for protection against actual or attempted acts of radiological sabotage, the NRC requires advance route approval. This approval is given on a shipment-by-shipment or series basis, it is not general approval of the route for subsequent spent fuel shipments. Spent fuel shipment routes, primarily for road transportation, but also including three rail routes, are indicated on reproductions of road maps. Also included are the amounts of material shipped during the approximate 8-year period that safeguards regulations have been effective. This information is current as of September 30, 1987

  18. Irradiated fuel by-product separation research in Canada

    International Nuclear Information System (INIS)

    Burston, M.

    1984-01-01

    Although no decision has been made to reprocess irradiated CANDU fuel, by-product separation research has recently been initiated in Canada because of its potential importance to Canadian research programs in advanced fuel cycles (especially U/Pu cycle development in the near term) and nuclear waste management. In addition, separated by-products could have a significant commercial potential. Demonstrated applications include: heat sources, gamma radiation sources, light sources, new materials for productions of other useful isotopes, etc. For illustrative purposes the calculated market value of by-products currently stored in irradiated CANDU fuel is approximately $210/kgU. Ontario Hydro has initiated a program to study the application of new separation technolgies, such as laser-based techniques and the plasma ion cyclotron resonance separation technique, to either augment and/or supplant the chemical extraction methods. The main goal is to develop new, more economical extraction methods in order to increase the magnitude of the advantages resulting from this approach to reprocessing. (author)

  19. Monitoring of releases from an irradiated fuel reprocessing plant

    International Nuclear Information System (INIS)

    Fitoussi, L.

    1978-01-01

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

  20. Thermophysical properties of the irradiated uranium-zirconium fuel

    International Nuclear Information System (INIS)

    Gajduchenko, A.B.

    2008-01-01

    The dependence of the thermophysical properties of metallic nuclear fuel, i.e. Zr alloy 40U, in a wide temperature range as a function of accumulated fission products amount is presented. Both non-irradiated and irradiated test pieces with different degrees of accumulation of fission products, i.e. 0.4, 0.6, and 0.9 g/cm 3 , are investigated. The specific heat is measured in the range of 50-1000 deg C, the thermal diffusivity is measured in the range 300-1000 deg C, and the variation of the dimensions and density of the samples on heating is also investigated. The thermal conductivity in the range of 50-1000 deg C is calculated on the basis of the experimental data [ru

  1. Cracking and bulk movement in irradiated uranium oxide fuel elements

    International Nuclear Information System (INIS)

    Bain, A.S.

    1963-09-01

    UO 2 pellets were fabricated with simulated circumferential or diametral cracks, and with voids formed by drilling axial or radial holes. Under irradiation the cracks healed in a region extending out slightly beyond the area of discernible grain growth. Cracks in the cooler outer annulus formed early and remained during the irradiation. Similarly voids in the outer annulus were unchanged, whereas those in the grain-growth region closed. Tungsten wire markers stayed in their original positions, demonstrating that the surrounding columnar grains in the UO 2 had not formed during the solidification of a melt. Decreases in diameter of 1 mm thick Zircaloy-2 sheathing assembled with large fuel/sheath diametral clearances were due to multi-axial stresses arising from axial elongation and the lack of diametral restraint. (author)

  2. Cracking and bulk movement in irradiated uranium oxide fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Bain, A S

    1963-09-15

    UO{sub 2} pellets were fabricated with simulated circumferential or diametral cracks, and with voids formed by drilling axial or radial holes. Under irradiation the cracks healed in a region extending out slightly beyond the area of discernible grain growth. Cracks in the cooler outer annulus formed early and remained during the irradiation. Similarly voids in the outer annulus were unchanged, whereas those in the grain-growth region closed. Tungsten wire markers stayed in their original positions, demonstrating that the surrounding columnar grains in the UO{sub 2} had not formed during the solidification of a melt. Decreases in diameter of 1 mm thick Zircaloy-2 sheathing assembled with large fuel/sheath diametral clearances were due to multi-axial stresses arising from axial elongation and the lack of diametral restraint. (author)

  3. SILICON CARBIDE GRAIN BOUNDARY DISTRIBUTIONS, IRRADIATION CONDITIONS, AND SILVER RETENTION IN IRRADIATED AGR-1 TRISO FUEL PARTICLES

    Energy Technology Data Exchange (ETDEWEB)

    Lillo, T. M.; Rooyen, I. J.; Aguiar, J. A.

    2016-11-01

    Precession electron diffraction in the transmission electron microscope was used to map grain orientation and ultimately determine grain boundary misorientation angle distributions, relative fractions of grain boundary types (random high angle, low angle or coincident site lattice (CSL)-related boundaries) and the distributions of CSL-related grain boundaries in the SiC layer of irradiated TRISO-coated fuel particles. Two particles from the AGR-1 experiment exhibiting high Ag-110m retention (>80%) were compared to a particle exhibiting low Ag-110m retention (<19%). Irradiated particles with high Ag-110m retention exhibited a lower fraction of random, high angle grain boundaries compared to the low Ag-110m retention particle. An inverse relationship between the random, high angle grain boundary fraction and Ag-110m retention is found and is consistent with grain boundary percolation theory. Also, comparison of the grain boundary distributions with previously reported unirradiated grain boundary distributions, based on SEM-based EBSD for similarly fabricated particles, showed only small differences, i.e. a greater low angle grain boundary fraction in unirradiated SiC. It was, thus, concluded that SiC layers with grain boundary distributions susceptible to Ag-110m release were present prior to irradiation. Finally, irradiation parameters were found to have little effect on the association of fission product precipitates with specific grain boundary types.

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  5. An evaluation of the results of the HTR fuel programme conducted in the Dragon reactor experiment

    International Nuclear Information System (INIS)

    Shepherd, L.R.

    1982-01-01

    The Dragon Reactor Experiment was used over a period of ten years to investigate the behaviour of HTR fuel elements under realistic service conditions. The purpose of the work was to develop fuel capable of meeting the requirements of commercial power reactors. The studies divided into areas concerned with the mechanical behaviour of the graphite core structure under fast neutron irradiation and the ability of the coated particle fuel to retain fissile products over commercially viable life-cycles. (author)

  6. Fission gas induced deformation model for FRAP-T6 and NSRR irradiated fuel test simulations

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Sasajima, Hideo; Fuketa, Toyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hosoyamada, Ryuji; Mori, Yukihide

    1996-11-01

    Pulse irradiation tests of irradiated fuels under simulated reactivity initiated accidents (RIAs) have been carried out at the Nuclear Safety Research Reactor (NSRR). Larger cladding diameter increase was observed in the irradiated fuel tests than in the previous fresh fuel tests. A fission gas induced cladding deformation model was developed and installed in a fuel behavior analysis code, FRAP-T6. The irradiated fuel tests were analyzed with the model in combination with modified material properties and fuel cracking models. In Test JM-4, where the cladding temperature rose to higher temperatures and grain boundary separation by the pulse irradiation was significant, the fission gas model described the cladding deformation reasonably well. The fuel had relatively flat radial power distribution and the grain boundary gas from the whole radius was calculated to contribute to the deformation. On the other hand, the power density in the irradiated LWR fuel rods in the pulse irradiation tests was remarkably higher at the fuel periphery than the center. A fuel thermal expansion model, GAPCON, which took account of the effect of fuel cracking by the temperature profile, was found to reproduce well the LWR fuel behavior with the fission gas deformation model. This report present details of the models and their NSRR test simulations. (author)

  7. Calculation and experimental estimation of the equation of state of irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bober, M; Breitung, W; Karow, H U; Schumacher, G [Gesellschaft fuer Kernforschung mbH, INR Kernforschungszentrum, Karlsruhe (Germany)

    1977-07-01

    The gas pressure development in an irradiated mixed oxide fuel is mainly influenced by fission gases and volatile fission products in the temperature range below the melting point and by the fuel material itself and the less volatile fission products in the temperature region above 4000 K. Besides the temperature the important factors for the vapor pressure are the oxygen potential of the fuel and the concentration of fission products in the fuel. As demonstrated previously the oxygen potential influences strongly the pressure of vapor species above (U Pu)O{sub 2}. The pressure of the species U, UO, UO{sub 2}, Pu, PuO, PuO{sub 2} varies over a range of more than five orders of magnitude by variation of the oxygen potential at 2000 K. Similar effects were observed with oxides of the fission products. Fission products dissolved in mixed oxide fuel on the other hand can influence significantly the oxygen potential of the irradiated mixed oxide. In the first paragraph of the paper an attempt is made to calculate oxygen potentials of mixed oxides containing dissolved fission products. The model used is based on the equilibrium of oxygen defects in the mixed oxide. The chemical state and distribution of fission products is a further behavior that should be considered in calculation of the local and overall pressures and behavior of the fuel. Fission products were transported during the irradiation time and collect at different positions within the fuel pin. This process can produce high local concentrations of fission products, thus enabling elements with low overall concentrations to reach their saturation pressure. The distribution of fission products and their behavior in irradiated mixed oxide fuel is described in the second paragraph. The third paragraph deals with the calculation of vapor pressures that has been conducted using a model described for uranium-plutonium mixed oxides. This model is based on the law of mass action and provides vapor pressures as a

  8. The irradiation induced creep in fuel compact materials for H.T.R. applications

    International Nuclear Information System (INIS)

    Veringa, H.; Blackstone, R.; Loelgen, R.

    1976-01-01

    Restrained shrinkage experiments up to 3 x 10 21 ncm -2 (DNE) in the temperature range of 600-1,200 0 C on three different dummy coated particle fuel compact materials were performed in the High Flux Reactor at Petten, the Netherlands. The data were evaluated to obtain the steady state irradiation creep coefficient of the compacts. It was found that for the materials investigated, the creep coefficient is temperature-dependent, but no clear relationship to the Young's modulus could be established. Under certain conditions, this irradiation-induced plasticity influences the elastic properties, while also the creep coefficient increases. This effect coincides with the formation and further opening of cracks due to stresses caused by irradiation shrinkage of the matrix material. (orig.) [de

  9. Determination of fuel irradiation parameters. Required accuracies and available methods

    International Nuclear Information System (INIS)

    Mas, P.

    1977-01-01

    This paper reports on the present point of some main methods to determine the nuclear parameters of fuel irradiation in testing reactors (nuclear power, burn up, ...) The different methods (theoretical or experimental) are reviewed: neutron measurements and calculations, gamma scanning, heat balance, ... . The required accuracies are reviewed: they are of 3-5 % on flux, fluences, nuclear power, burn-up, conversion factor. These required accuracies are compared with the real accuracies available which are the present time of order of 5-20 % on these parameters

  10. Irradiation behavior of U 6Mn-Al dispersion fuel elements

    Science.gov (United States)

    Meyer, M. K.; Wiencek, T. C.; Hayes, S. L.; Hofman, G. L.

    2000-02-01

    Irradiation testing of U 6Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U 6Mn in an unrestrained plate configuration performs similarly to U 6Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3×10 27 m-3. Fuel plate failure occurs by fission gas pressure driven `pillowing' on continued irradiation.

  11. Modified fuel channel for sample irradiation at the RB reactor

    International Nuclear Information System (INIS)

    Pesic, M.; Markovic, H.; Sokcic, M.; Miric, I.; Prokic, M.; Strugar, P.

    1983-01-01

    Fuel channel of 80% enriched UO 2 at RB reactor in Boris Kidric Institute of nuclear sciences is modified for sample irradiation in the fast neutron field. Maximum sample diameter is 25 mm and length up to 100 mm. Characteristics of neutron as well as gamma radiation fields of this new experimental channel are investigated. In the center of channel, the main contribution to the total neutron absorbed dose i.e. 0.29 Gy per 1 Wh of reactor operation, is due to the fast neutron spectrum component. Only 0.05 Gy and 0.07 Gy in the total neutron absorbed dose are due to epithermal and thermal neutrons respectively. At the same time gamma absorption dose is 0.35 Gy. The development of experimental fuel channel GRK has wide possibility for utilization, from electronic components fast neutron studies, dosimeters testing, to cross section measurements for fast neutron energies. (author)

  12. Safety aspects of advanced fuels irradiations in EBR-II

    International Nuclear Information System (INIS)

    Lehto, W.K.

    1975-09-01

    Basic safety questions such as MFCI, loss-of-Na bond, pin behavior during design basis transients, and failure propagation were evaluated as they pertain to advanced fuels in EBR-II. With the exception of pin response to the unlikely loss-of-flow transient, the study indicates that irradiation of significant numbers of advanced fueled subassemblies in EBR-II should pose no safety problems. The analysis predicts, however, that Na boiling may occur during the postulated design basis unlikely loss-of-flow transient in subassemblies containing He-bonded fuel pins with the larger fuel-clad gaps. The calculations indicate that coolant temperatures at top of core in the limiting S/A's, containing the He bonded pins, would reach approximately 1480 0 F during the transient without application of uncertainty factors. Inclusion of uncertainties could result in temperature predictions which approach coolant boiling temperatures (1640 0 F). Further analysis of He-bonded pins is being done in this potential problem area, e.g., to apply best estimates of uncertainty factors and to determine the sensitivity of the preliminary results to gap conductance

  13. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    Science.gov (United States)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  14. BWR fuel experience with zinc injection

    International Nuclear Information System (INIS)

    Levin, H.A.; Garcia, S.E.

    1995-01-01

    In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

  15. POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

    Directory of Open Access Journals (Sweden)

    H.J. RYU

    2013-12-01

    Full Text Available Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4–5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr, additional protective coatings (silicide or nitride, and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

  16. Gadolinia experience and design for PWR fuel cycles

    International Nuclear Information System (INIS)

    Stephenson, L. C.

    2000-01-01

    The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

  17. Computer-controlled gamma-ray scanner for irradiated reactor fuel

    International Nuclear Information System (INIS)

    Mandler, J.W.; Coates, R.A.; Killian, E.W.

    1979-01-01

    Gamma-ray scanning of irradiated fuel is an important nondestructive technique used in the thermal fuels behavior program currently under way at the Idaho National Engineering Laboratory. This paper is concerned with the computer-controlled isotopic gamma-ray-scanning system developed for postirradiation examination of fuel and includes a brief discussion of some scan results obtained from fuel rods irradiated in the Power-Burst Facility to illustrate gamma-ray spectrometry for this application. Both burnup profiles and information concerning fission-product migration in irradiated fuel are routinely obtained with the computer-controlled system

  18. SPHERE: Irradiation of sphere-pac fuel of UPuO2−x containing 3% Americium

    International Nuclear Information System (INIS)

    D’Agata, E.; Hania, P.R.; McGinley, J.; Somers, J.; Sciolla, C.; Baas, P.J.; Kamer, S.; Okel, R.A.F.; Bobeldijk, I.; Delage, F.; Bejaoui, S.

    2014-01-01

    Highlights: • SPHERE is designed to check the behaviour of MADF sphere-pac concept. • MADF sphere-pac are compared with MADF pellet. • Swelling, helium release and restructuring behaviour will be the main output of the experiment. • An experiment to check sphere-pac MABB fuel behaviour is now under design. - Abstract: Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like 241 Am is therefore an option for the reduction of radiotoxicity of waste packages to be stored in a repository. The SPHERE irradiation experiment is the latest of a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS) performed in the HFR (High Flux Reactor). The SPHERE experiment is carried out in the framework of the 4-year project FAIRFUELS of the EURATOM 7th Framework Programme (FP7). During the past years of experimental works in the field of transmutation and tests of innovative nuclear fuels, the release or trapping of helium as well as helium induced fuel swelling have been shown to be the key issues for the design of Am-bearing targets. The main objective of the SPHERE experiment is to study the in-pile behaviour of fuel containing 3% of americium and to compare the behaviour of sphere-pac fuel to pellet fuel, in particular the role of microstructure and temperature on fission gas release (mainly He) and on fuel swelling. The SPHERE experiment is being irradiated since September 2013 in the HFR in Petten (The Netherlands) and is expected to be terminated in spring 2015. The experiment has been designed to last up to 18 reactor cycles (corresponding to 18 months) but may reach its target earlier. This paper discusses the rationale and objective of the SPHERE experiment and provides a general description of its design

  19. 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

  20. Behavior of solid fission products in irradiated fuel

    International Nuclear Information System (INIS)

    Song, Ung Sup; Jung, Yang Hong; Kim, Hee Moon; Yoo, Byun Gok; Kim, Do Sik; Choo, Yong Sun; Hong, Kwon Pyo

    2004-01-01

    Many fission products are generated by fission events in UO 2 fuel under irradiation in nuclear reactor. Concentration of each fission product is changed by conditions of neutron energy spectrum, fissile material, critical thermal power, irradiation period and cooling time. Volatile materials such as Cs and I, the fission products, degrade nuclear fuel rod by the decrease of thermal conductivity in pellet and the stress corrosion cracking in cladding. Metal fission products (white inclusion) make pellet be swelled and decrease volume of pellet by densification. It seems that metal fission products are filled in the pore in pellet and placed between UO 2 lattices as interstitial. In addition, metal oxide state may change structural lattice volume. Considering behavior of fission products mentioned above, concentration of them is important. Fission products could be classified as bellows; solid solution in matrix : Sr, Zr, Nb, Y, La, Ce, Pr, Nd, Pm, Sm - metal precipitates : Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sb, Te - oxide precipitates : Ba, Zr, Nb, Mo, (Rb, Cs, Te) - volatile and gases : Kr, Xe, Br, I, (Rb, Cs, Te)

  1. Post-irradiation examination of overheated fuel bundles

    International Nuclear Information System (INIS)

    Sears, D.F.; Primeau, M.F.; Leach, D.A.

    1995-01-01

    Post-irradiation examinations (PIE) were conducted on prototype 43-element CANDU fuel bundles that overheated during test irradiations in the NRU reactor. PIE revealed that the bundles remained physically intact, but on several elements the Zr-4 sheath collapsed into axial gaps between the pellet stack and end caps, between adjacent pellets within the stacks, and into missing pellet chips and cracks. Helium pressurization tests showed that none of the collapsed elements leaked. Hydride blisters were discovered on a few elements, but the source of the hydrogen was not linked to a breach of the cladding or end caps. These defects were attributed to primary hydriding. Microstructural changes in the fuel and cladding indicate that the cladding-was briefly exposed to temperatures in the range 600-800 o C and pressures above 11.2 MPa. The results show that Zr-4 cladding behaves in a highly ductile manner during such transient, high-temperature and high-pressure excursions. (author)

  2. Post-irradiation examination of overheated fuel bundles

    International Nuclear Information System (INIS)

    Sears, D.F.; Primeau, M.F.; Leach, D.A.

    1997-08-01

    Post-irradiation examinations (PIE) were conducted on prototype 43-element CANDU fuel bundles that overheated during test irradiations in the NRU reactor. PIE revealed that the bundles remained physically intact, but on several elements the Zr-4 sheath collapsed into axial gaps between the pellet stack and end caps, between adjacent pellets within the stacks, and into missing pellet chips and cracks. Helium pressurization tests showed that none of the collapsed elements leaked. Hydride blisters were discovered on a few elements, but the source of the hydrogen was.not linked to a breach of the cladding or end caps. These defects were attributed to primary hydriding. Microstructural changes in the fuel and cladding indicate that the cladding was briefly exposed to temperatures in the range 600-800 o C and pressures above 11.2MPa. The results show that Zr-4 cladding behaves in a highly ductile manner during such transient, high-temperature and high-pressure excursions. (author)

  3. Design verification test of instrumented capsule (02F-11K) for nuclear fuel irradiation in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, J. M.; Oh, J. M. [and others

    2004-01-01

    An instrumented capsule is being developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. The instrumented capsule includes three test fuel rods installed thermocouple to measure fuel centerline temperature and three SPNDs (Self-Powered Neutron Detector) to monitor the neutron flux. Its stability was verified by out-of-pile performance test, and its safety evaluation was also shown that the safety requirements were satisfied. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.8 full power days at 24 MWth). During irradiation, the centerline temperature of PWR UO{sub 2} fuel pellets fabricated by KEPCO Nuclear Fuel Company and the neutron flux were continuously measured and monitored. The test fuel rods were irradiated at less than 350 W/cm to 5.13 GWD/MTU with fuel centerline peak temperature below 1,375 .deg. C. The structural stability of the capsule was satisfied by the naked eye in service pool of HANARO. The capsule and test fuel rods were dismantled and test fuel rods were examined at the hot cell of IMEF (Irradiated Material Examination Facility)

  4. Design and manufacturing of instrumented capsule(03F-05K) for nuclear fuel irradiation in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, J. M.; Shin, Y. T. [and others

    2004-06-01

    An instrumented capsule is being developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule(02F-11K) for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. The instrumented capsule includes three test fuel rods installed thermocouple to measure fuel centerline temperature and three SPNDs (self-powered neutron detector) to monitor the neutron flux. Its stability was verified by out-of-pile performance test, and its safety evaluation was also shown that the safety requirements were satisfied. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.8 full power days at 24 MWth). During irradiation, the centerline temperature of PWR UO{sub 2} fuel pellets fabricated by KEPCO Nuclear Fuel Company and the neutron flux were continuously measured and monitored. In the year of 2004, 3 test fuel rods and the instrumented capsule(03F-05K) were designed and fabricated to measure fuel centerline temperature, internal pressure of fuel rod, and fuel axial deformation during irradiation test. This capsule is being irradiated in the test hole OR5 of HANARO reactor from April 26, 2004.

  5. The Analysis of RSG-GAS Spent Fuel Elements Utilization as a Gamma Irradiator

    International Nuclear Information System (INIS)

    Pudjijanto MS; Setiyanto

    2004-01-01

    A gamma irradiator using RSG-GAS spent fuels was analyzed. The cylindrical geometry of the irradiator was designed using spent fuels placed in the cylindrical periphery. The analysis especially was focused to evaluate the feasibilities of the irradiator for foods and non-foods which need not too high dose rates. While the spent fuels activities were calculated by ORIGEN2 code, the dose rates at the irradiation positions were determined by linear attenuation model with transport coefficient. The evaluated results showed that the cylindrical geometry of the irradiator with diameter around 1-1.5 m gave the effective dose rate for irradiation needs the dose rate about 2 kGy/hr. Regarding this work, it can be concluded that one can use the unutilized spent fuels effectively as a gamma irradiator for certain applications. (author)

  6. Development of an End-plug Welding Technology for an Instrumented Fuel Irradiation Test

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Lee, Chul Yong; Shin, Yoon Taek; Choo, Kee Nam

    2010-01-01

    The irradiation test of end-plug specimens was planned for the evaluation of nuclear fuels performance. To establish the fabrication process, and for satisfying the requirements of the irradiation test, an orbital-GTA weld machine for the specimens of the dual rods was developed, and the preliminary welding experiments for optimizing the process conditions of the specimens of the dual rods were performed. Dual rods with a 9.5mm diameter and a 0.6mm wall thickness of the cladding tubes and end-plugs have been used and the optimum conditions of the pin-hole welding have also been selected. This paper describes the experimental results of the GTA welds of the specimens of the dual rods and the metallography examinations of the GTA welded specimens for various welding conditions for the instrumented fuel irradiation test. These investigations satisfied the requirements of the instrumented irradiation test and the GTA welds for the specimens of the dual rods at the HANARO research reactor

  7. The development of the neutron flux measurement technology using SPNDs during nuclear fuel irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B. G.; Kang, Y. H.; Cho, M. S.; Joo, K. N.; Choi, M. H.; Park, S. J.; Shin, Y. T.; Oh, J. M.; Kim, Y. J

    2004-03-01

    As a part of the development of instrumentation technologies for a nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), a study is performed to measure and evaluate the neutron flux at the same position as the nuclear fuel during irradiation test using the SPND(Self Powered Neutron Detector). To perform this study, rhodium type SPNDs and amplifier are selected suitable to irradiation test, and the selected SPNDs are installed in instrumented fuel capsule(02F-11K). The irradiation test using a instrumented fuel capsule are performed in the OR5 vertical hole of HANARO for about 54 days, and SPND output signals are acquired successfully during irradiation test. Acquired SPND signals are analyzed and evaluated as a reliable data by COSMOS Code. This will be utilized for the fuel related research together with fuel center temperature and reactor operation data.

  8. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Provost, Jean-Luc; Debes, Michel

    2005-01-01

    Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

  9. Post-irradiation examinations of inert matrix nitride fuel irradiated in JMTR (01F-51A capsule)

    International Nuclear Information System (INIS)

    Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Honda, Junichi; Hatakeyama, Yuichi; Ono, Katsuto; Matsui, Hiroki; Arai, Yasuo

    2007-03-01

    A plutonium nitride fuel pin containing inert matrix such as ZrN and TiN was encapsulated in 01F-51A and irradiated in JMTR. Minor actinides are surrogated by plutonium. Average linear powers and burnups were 408W/cm, 30000MWd/t(Zr+Pu) [132000MWd/t-Pu] for (Zr,Pu)N and 355W/cm, 38000MWd/t(Ti+Pu) [153000MWd/t-Pu] for (TiN,PuN). The irradiated capsule was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pin. Very low fission gas release rate of about 1.6% was measured. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

  10. Separation of Plutonium from Irradiated Fuels and Targets

    Energy Technology Data Exchange (ETDEWEB)

    Gray, Leonard W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Holliday, Kiel S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Murray, Alice [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Thompson, Major [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Thorp, Donald T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Yarbro, Stephen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Venetz, Theodore J. [Hanford Site, Benton County, WA (United States)

    2015-09-30

    Spent nuclear fuel from power production reactors contains moderate amounts of transuranium (TRU) actinides and fission products in addition to the still slightly enriched uranium. Originally, nuclear technology was developed to chemically separate and recover fissionable plutonium from irradiated nuclear fuel for military purposes. Military plutonium separations had essentially ceased by the mid-1990s. Reprocessing, however, can serve multiple purposes, and the relative importance has changed over time. In the 1960’s the vision of the introduction of plutonium-fueled fast-neutron breeder reactors drove the civilian separation of plutonium. More recently, reprocessing has been regarded as a means to facilitate the disposal of high-level nuclear waste, and thus requires development of radically different technical approaches. In the last decade or so, the principal reason for reprocessing has shifted to spent power reactor fuel being reprocessed (1) so that unused uranium and plutonium being recycled reduce the volume, gaining some 25% to 30% more energy from the original uranium in the process and thus contributing to energy security and (2) to reduce the volume and radioactivity of the waste by recovering all long-lived actinides and fission products followed by recycling them in fast reactors where they are transmuted to short-lived fission products; this reduces the volume to about 20%, reduces the long-term radioactivity level in the high-level waste, and complicates the possibility of the plutonium being diverted from civil use – thereby increasing the proliferation resistance of the fuel cycle. In general, reprocessing schemes can be divided into two large categories: aqueous/hydrometallurgical systems, and pyrochemical/pyrometallurgical systems. Worldwide processing schemes are dominated by the aqueous (hydrometallurgical) systems. This document provides a historical review of both categories of reprocessing.

  11. HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.; Chrisensen, Cad L.

    2016-11-01

    High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Test Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two

  12. Multi-physic simulations of irradiation experiments in a technological irradiation reactor

    International Nuclear Information System (INIS)

    Bonaccorsi, Th.

    2007-09-01

    A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

  13. Comparison of silver release predictions using PARFUME with results from the AGR-2 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P.; Demkowicz, Paul A.; Baldwin, Charles A.; Harp, Jason M.; Hunn, John D.

    2016-11-01

    The PARFUME (PARticle FUel ModEl) code was used to predict silver release from tristructural isotropic (TRISO) coated fuel particles and compacts during the second irradiation experiment (AGR-2) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-2 experiment used the fuel compact volume average temperature for each of the 559 days of irradiation to calculate the release of fission product silver from a representative particle for a select number of AGR-2 compacts and individual fuel particles containing either mixed uranium carbide/oxide (UCO) or 100% uranium dioxide (UO2) kernels. Post-irradiation examination (PIE) measurements were performed to provide data on release of silver from these compacts and individual fuel particles. The available experimental fractional releases of silver were compared to their corresponding PARFUME predictions. Preliminary comparisons show that PARFUME under-predicts the PIE results in UCO compacts and is in reasonable agreement with experimental data for UO2 compacts. The accuracy of PARFUME predictions is impacted by the code limitations in the modeling of the temporal and spatial distributions of the temperature across the compacts. Nevertheless, the comparisons on silver release lie within the same order of magnitude.

  14. Fabrication of Non-instrumented capsule for DUPIC simulated fuel irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.G.; Kang, Y.H.; Park, S.J.; Shin, Y.T. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    In order to develope DUPIC nuclear fuel, the irradiation test for simulated DUPIC fuel was planed using a non-instrumented capsule in HANARO. Because DUPIC fuel is highly radioactive material the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO was designed to remotely assemble and disassemble in hot cell. And then, according to the design requirements the non-instrumented DUPIC capsule was successfully manufactured. Also, the manufacturing technologies of the non-instrumented capsule for irradiating the nuclear fuel in HANARO were established, and the basic technology for the development of the instrumented capsule technology was accumulated. This report describes the manufacturing of the non-instrumented capsule for simulated DUPIC fuel. And, this report will be based to develope the instrumented capsule, which will be utilized to irradiate the nuclear fuel in HANARO. 26 refs., 4 figs. (Author)

  15. Post-irradiation examinations of uranium-plutonium mixed nitride fuel irradiated in JMTR (89F-3A capsule)

    International Nuclear Information System (INIS)

    Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Arai, Yasuo; Kimura, Yasuhiko; Nagashima, Hisao; Sekita, Noriaki

    2000-03-01

    Two helium-bonded fuel pins filled with uranium-plutonium mixed nitride pellets were encapsulated in 89F-3A and irradiated in JMTR up to 5.5% FIMA at a maximum linear power of 73 kW/m. The capsule cooled for ∼5 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pins. Very low fission gas release rate of about 2 ∼ 3% was observed, while the diametric increase of fuel pin was limited to ∼0.4% at the position of maximum reading. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

  16. Vented fuel experiment for gas-cooled fast reactor application

    International Nuclear Information System (INIS)

    Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

    1976-01-01

    A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100MWd/kg-heavy metal. The fuel is a sol-gel-derived 88 at.% uranium (approximately 9% 235 U) and 12 at.% plutonium oxide, and the cladding is 20% cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70MWd/kg. The fuel has been operated at linear power rates of 39 and 44kW/m, and peak outer cladding temperature of 565 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48kW/m (685 0 C). Helium gas sweeps through any portion of the three regions of the fuel rod, namely: fuel, blanket, and charcoal trap. The charcoal trap is operated at about 300 0 C. An on-line Ge(Li) detector is used to analyse release rates of several gamma-emitting noble gas isotopes. Analyses are performed primarily on sweep gas flowing through the entire fuel rod, and for sweeps over the top of the charcoal trap. Sweep gas samples are analyzed for stable noble gas isotopes. Results in the form of ratios of release rate over birth rate (R/B) and venting rate over birth rate (V/B) are derived. R/B rates range from 10 -4 % to 30% while V/B ranges from 10 -6 % to 30%. Flow conductance in the capsule was monitored by recording the flow rate and pressure drop across the fuel rod and inlet sweep line. The flow conductance has been falling with increasing burnup, currently restricting the flow to about 20ml (s.t.p.)/min at a pressure difference of about 1.5MPa. Venting rates of the gaseous fission products as a function of gas pressure in the range 6.9 to 1.4MPa have also been measured. Planned future experiments include the monitoring of tritium release, venting and cladding permeation rates, and its molecular form. First measurements have been made. A simulated leak experiment will determine the mixture of fission gases as a function of flow rate and the most

  17. MOX fuel transport: the French experience

    International Nuclear Information System (INIS)

    Sanchis, H.; Verdier, A.; Sanchis, H.

    1999-01-01

    In the back-end of the fuel cycle, several leading countries have chosen the Reprocessing, Conditioning, Recycling (RCR) option. Plutonium recycling in the form of MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants an several European countries. The COGEMA Group has developed the industrialized products to master the RCR operation including transport COGEMA subsidiary, TRANSNUCLEAIRE has been operating MOX fuel transports on an industrial scale for more than 10 years. In 1998, around 200 transports of Plutonium materials have been organised by TRANSNUCLEAIRE. These transports have been carried out by road between various facilities in Europe: reprocessing plants, manufacturing plants and power plants. The materials transported are either: PuO 2 and MOX powder; BWR and PWR MOX fuel rods; BWR and PWR MOX fuel assemblies. Because MOX fuel transport is subject to specific safety, security and fuel integrity requirements, the MOX fuel transport system implemented by TRANSNUCLEAIRE is fully dedicated. Packaging have been developed, licensed and manufactured for each kind of MOX material in compliance with relevant regulations. A fleet of vehicles qualified according to existing physical protection regulations is operated by TRANSNUCLEAIRE. TRANSNUCLEAIRE has gained a broad experience in MOX transport in 10 years. Technical and operational know-how has been developed and improved for each step: vehicles and packaging design and qualification; vehicle and packaging maintenance; transport operations. Further developments are underway to increase the payload of the packaging and to improve the transport conditions, safety and security remaining of course top priority. (authors)

  18. Development of oxygen sensing technology in an irradiated fuel rod. Characteristic test of oxygen sensor

    International Nuclear Information System (INIS)

    Saito, Junichi; Hoshiya, Taiji; Sakurai, Fumio; Sakai, Haruyuki

    1996-03-01

    At the Department of JMTR (Japan Materials Test Reactor), the re-instrumentation technologies to a high burnup fuel rod irradiated in an LWR have been developed to study irradiation behavior of the fuel during power transient. It has been progressed developing a chemical sensor as one of the re-instrumentation technologies. This report summarizes the results of characteristic tests of an oxygen sensor made of Yttria Stabilized Zirconia (YSZ) as a solid electrolyte. Several kinds of experiments were carried out to evaluate the electromotive force (emf) performance, stability and lifetime of the oxygen sensor with Ni/NiO, Cr/Cr 2 O 3 and Fe/FeO, respectively as a reference electrode. From the experimental data, it is suggested that the reference electrode of Ni/NiO reveals the most appropriate characteristic of the sensor to measure the partial oxygen pressure in a fuel rod. It is the final goal of this development to clarify the change of oxygen chemical potential in a fuel rod during power transient. (author)

  19. A model to predict failure of irradiated U–Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Senor, David J.; Casella, Andrew M.

    2016-12-15

    Highlights: • Simple model to predict failure of dispersion fuel meat designs. • Evaluated as a function of fabrication parameters and irradiation conditions. • Predictions compare well with experimental measurements of miniature fuel plates. • Interaction layer formation reduces matrix strength and increases temperature. • Si additions to the matrix appear effective only at moderate heat flux and burnup. - Abstract: Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium–molybdenum (U–Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO{sub 2}-stainless steel dispersion fuels and uses currently available thermal–mechanical property information for the materials of interest in the currently proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as onset of pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the {sup 235}U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of

  20. Improving the AGR fuel testing power density profile versus irradiation-time in the advanced test reactor

    International Nuclear Information System (INIS)

    Chang, Gray S.; Lillo, Misti A.; Maki, John T.; Petti, David A.

    2009-01-01

    The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235 U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235 U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250degC throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235 U

  1. LWR mox fuel experience in Belgium and France with special emphasis on results obtained in BR3

    International Nuclear Information System (INIS)

    Bairiot, H.; Haas, D.; Lippens, M.; Motte, F.; Lebastard, G.; Marin, J.F.

    1986-09-01

    The course of the paper reflects two main topics: LWR MOX fuel experience in Belgium and France, summarizing the fabrication techniques, the references, the underlying MOX fuel technology and the current R and D programs for expanding the data base; behaviour of MOX fuel rods irradiated under steady state and transient operating conditions, focusing on MOX fuel technology features acquired through the irradiations performed in the BR3 PWR, supplemented by tests in the BR2 MTR. This paper focuses on the thermomechanical behaviour of LWR MOX fuel rods, which is intimately related to the fabrication technique and vice-versa. 22 refs

  2. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  3. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    International Nuclear Information System (INIS)

    Cowell, B.S.; Fisher, S.E.

    1999-01-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

  4. Survey of post-irradiation examinations made of mixed carbide fuels

    International Nuclear Information System (INIS)

    Coquerelle, M.

    1997-01-01

    Post-irradiation examinations on mixed carbide, nitride and carbonitride fuels irradiated in fast flux reactors Rapsodie and DFR were carried out during the seventies and early eighties. In this report, emphasis was put on the fission gas release, cladding carburization and head-end gaseous oxidation process of these fuels, in particular, of mixed carbides. (author). 8 refs, 16 figs, 3 tabs

  5. Preliminary test results for post irradiation examination on the HTTR fuel

    International Nuclear Information System (INIS)

    Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

    2007-01-01

    The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

  6. Evaluation of the problems associated with ENEL's irradiated fuel management

    International Nuclear Information System (INIS)

    Sani, L.; Belelli, U.

    1978-01-01

    The present difficulties in the reprocessing field, wich will continue for a period of time not easily definable, impose on the nuclear utilities the research of adequate solutions to solve the problem associated with storage of the fuels discharged by the nuclear power stations in operation. In this context, the paper examines the technical aspects of the various possible alternatives and on the basis of the Italian nuclear program, possible storage strategies are discussed in the assumption of the development of an adequate oxide-fuel reprocessing capacity. The problems related to storage in pools at the station or away from reactor are described in the light of the recent experience made in Italy with the adaptation of an old pool reactor out of service, that is, the Avogadro reactor in Saluggia, to store the fuels of the stations now in operation. Finally some economic aspects are considered in order to ascertain their impact on the cost of energy produced. (author)

  7. VHTR-fuel irradiation capsules for VT-1 hole of JRR-2

    International Nuclear Information System (INIS)

    Kikuchi, Teruo; Kikuchi, Akira; Tobita, Tsutomu; Kashimura, Satoru; Miyasaka, Yasuhiko

    1977-02-01

    Irradiations of VHTR fuels were made in the VT-1 irradiation hole of JRR-2. Three capsules, VP-1, VP-2 and VP-4, which contained fuel compacts, were irradiated for 300 hr at temperatures of 950 0 , 1370 0 and 1500 0 C up to the estimated burn-ups of 0.74, 0.87 and 0.80%FIMA, respectively. And, to study the amoeba effect of fuel particles, two capsules, VP-3 and VP-5, were irradiated for 300 hr at temperatures of 1650 0 and 1670 0 C up to the estimated burn-ups of 0.38 and 0.33%FIMA, respectively. (auth.)

  8. Performance test of the I and C system (GSF - 2002) for the irradiation tests using a fuel capsule

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Park, S. J.; Kim, B. G.; Ahn, D. H

    2004-12-01

    HANARO is a very important facility in Korea. It offers various types of irradiation tests of nuclear fuels and materials. With the various applications of the HANARO capsule for the academic and industrial applications, new technologies and relevant facilities will become more important especially for the advanced nuclear fuels and materials development. A new I and C system for an irradiation test using an instrumented fuel capsule have been designed and manufactured to provide more qualified data to fuel developer. The performance test which started in 2004, was done to investigate the thermal response of the capsule connected to the gas mixing system of the new I and C system(GSF-2002) in the cold test loop under the HANARO hydraulic operational condition. Main test parameters are mass flow rate of 25, 50 and 100 cc/min of He/Ne gas, gas pressure of 1 to 3 kg/cm{sup 2}, heater power of 1 to 3.4kW and different gas mixing ratios of He to Ne. From the out-pile tests, it was confirmed that the I and C system(GSF-2002) would be feasible for the fuel irradiation tests. Both analytical and test data prepared by this study are directly used for the fuel experiments related to advanced fuel development program.

  9. Dry Storage at long term of nuclear fuels: Influence of the fuel design and commercial irradiation conditions

    International Nuclear Information System (INIS)

    Marino, Armando C

    2009-01-01

    The BaCo code was applied to simulate the behaviour for a PHWR fuel under storage conditions showing a strong dependence on the original design of the fuel and the irradiation history. In particular, the results of the statistical analysis of BaCo indicate that the integrity of the fuel is influenced by the manufacture tolerances and the solicitations during the NPP irradiation. The main conclusion of the present study is that the fuel temperature of the device should be carefully controlled in order to ensure safe storage conditions. [es

  10. Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Khericha, S.T.

    2002-06-30

    This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to {approx}42 GWd/MT burnup (+ 2.5%) as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: {approx}50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies ({at} {approx}40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches {approx}40 GWd/MT burnup per MCNP-predicted values.

  11. Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

    International Nuclear Information System (INIS)

    Thoms, K.R.

    1975-04-01

    Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 (Ta-8 percent W-2 percent Hf) and uranium dioxide (UO 2 ) fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1 percent Zr. A total of nine fuel pins was irradiated (four containing porous UN, two containing dense, nonporous UN, and three containing dense UO 2 ) at average cladding temperatures ranging from 931 to 1015 0 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO 2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of 10 -5 . (U.S.)

  12. Release behavior of fission products from irradiated dispersion fuels at high temperatures

    International Nuclear Information System (INIS)

    Iwai, Takashi; Shimizu, Michio; Nakagawa, Tetsuya

    1990-02-01

    As a framework of reduced enrichment fuel program of JMTR Project, the measurements of fission products release rates at high temperatures (600degC - 1100degC) were performed in order to take the data to use for safety evaluation of LEU fuel. Three type miniplates of dispersion silicide and aluminide fuel, 20% enrichment LEU fuel with 4.8 gU/cc (U 3 Si 2 90 %, USi 10 % and U 3 Si 2 50 %, U 3 Si 50 % dispersed in aluminium) and 45 % enrichment MEU fuel with 1.6 gU/cc, were irradiated in JMTR. The burnups attained by one cycle (22 days) irradiation were within 21.6 % - 22.5 % of initial 235 U. The specimens cut down from miniplates were measured on fission products release rates by means of new apparatus specially designed for this experiment. The specimens were heated up within 600degC - 1100degC in dry air. Then fission products such as 85 Kr, 133 Xe, 131 I, 137 Cs, 103 Ru, 129m Te were collected at each temperature and measured on release rates. In the results of measurement, the release rates of 85 Kr, 133 Xe, 131 I, 129m Te from all specimens were slightly less than that of G.W. Parker's data on U-Al alloy fuel. For 137 Cs and 103 Ru from a silicide specimen (U 3 Si 2 90 %, USi 10 % dispersed in aluminium) and 137 Cs from an aluminide specimen, the release rates were slightly higher than that of G.W. Parker's. (author)

  13. Formulation of engineering design principles for the treatment of irradiated fuel and associated radioactive waste

    International Nuclear Information System (INIS)

    Banford, A.W.; Hanson, B.C.; Scully, P.J.; Taylor, R.

    2007-01-01

    The industrial scale treatment of irradiated fuel in the UK has resulted in BNFL developing extensive experience of the process design, build, commissioning, and operation necessary for successful nuclear processing plant. Much of the design experience now resides in Nexia Solutions (formally BNFL Research and Development Division) who have always defined and undertaken the extensive development programmes necessary to underpin the design at all stages of the project life-cycle. Since the 1990's, Nexia Solutions has built up a large portfolio of plant designs for a range of spent fuel applications, from fuel conditioning to partitioning and transmutation. In addition, by investigation of a large and diverse portfolio of technologies Nexia Solutions has developed innovative concepts for plant design that could present significant economic savings on conventional approaches. Using this experience and the lessons learned, we have developed and refined our own engineering design principles necessary for the successful design of commercial spent fuel and waste treatment plant. Our approach is to advocate an integral concept, with both science and engineering designs working in parallel during development. 4 foundation principles for success have been identified: -) understand the strategic objective, -) adopt a risk driven programme, -) engage in engineering activities early, and -) timely application of appropriate engineering methodologies. 2 Case studies presented in this paper: first, the BNFL segregated effluent treatment plant and secondly, the selection of a pyrochemical process for recycle of fast reactor, demonstrate how this approach has been adopted and the benefits that have been gained

  14. Program description for the qualification of CNEA - Argentina as a supplier of LEU silicide fuel and post-irradiation examinations plan for the first prototype irradiated in Argentina

    International Nuclear Information System (INIS)

    Rugirello, Gabriel; Adelfang, Pablo; Denis, Alicia; Zawerucha, Andres; Marco, Agustin di; Guillaume, Eduardo; Sbaffoni, Monica; Lacoste, Pablo

    1998-01-01

    In this report we present a description of the ongoing and future stages of the program for the qualification of CNEA, Argentina, as a supplier of low enriched uranium silicide fuel elements for research reactor. Particularly we will focus on the characteristics of the future irradiation experiment on a new detachable prototype, the post-irradiation examinations (PIE) plan for the already irradiated prototype PO4 and an overview of the recently implemented PIE facilities and equipment. The program is divided in several steps, some of which have been already completed. It concludes: development of the uranium silicide fissile material, irradiation and PIE of several full-scale prototypes. Important investments have been already carried out in the facilities for the FE production and PIE. (author)

  15. 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.

  16. Evaluation of the effect of probe design parameters on ECT signal and development of eddy current probe for irradiated fuel rods

    International Nuclear Information System (INIS)

    Kwank, S. W.; Han, Y. K.; Woo, S. K.; Kim, T. W.; Park, J. Y.; Kim, B. J.; Park, J. Y.

    1999-01-01

    Eddy current test(ECT) is used to inspect not only the failed fuel rods but also peripheral rods during repairing of the failed fuel rods, to detect internal defects in irradiated fuel rods which could not be detected by ultrasonic test and visual test, and to obtain the data for determining the root cause of fuel rod failure. This study evaluates the effect of properties of test article, irradiated fuel rods, on the impedance diagram in order to reduce the difficulty of ECT signal analysis. The optimum eddy current probe design conditions for inspecting the irradiated fuel rods, is estimate by using experimental equations and the probe is manufactured based on the estimated conditions. The performance of developed eddy current probe and the optimum conditions is proved through characteristic comparison experiment with the probe purchased from the foreign vendor

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  18. Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

    International Nuclear Information System (INIS)

    Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

    1982-01-01

    The document reports in detail the fuel pin fabrication data and describes the irradiation conditions and history. All the relevant results of the non-destructive and destructive post-irradiation examinations are reported. They include: visual inspection and chemical analysis of crud; length and diameter measurements; neutron radiography and gamma scanning; juncture tests and fission gas analysis (including residual gas in fuel samples); microscopy and alpha + beta/gamma autoradiography; microprobe investigations; burn-up and isotopic analysis; and hydrogen analysis in clad. The data and observations obtained are discussed in detail and conclusions are given. The irradiation and post-irradiation examinations of the R-109 pins have shown the safe, pre-calculable performance of LWR fuel pins containing mixed-oxide sphere-pac fuel with the fissile material mainly present in the large spheres

  19. Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu

    2008-01-01

    U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during irradiation tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particle size was increases. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth. (author)

  20. Irradiation behavior of modified high-performance nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jungwirth, Rainer

    2011-11-03

    To overcome the problem of UMo/Al fuel swelling, four different possibilities have been identified: (i) the modification of the Al matrix by adding diffusion limiting elements (ii) the insertion of a diffusion barrier at the interface UMo-Al (iii) further alloying the UMo with a third element to stabilize the γ-UMo phase (iv) a combination of means (i)-(iii). In consequence, 20 different UMoX/AlY (X=Si, Ti, Mg, Bi, with and without oxidation layer; Y=Nb, Ti, Pt) samples have been examined before and after irradiation with Iodine at 80MeV. First it has been shown, that a protective oxidation layer on the UMo grains does not prevent the formation of a interdiffusion layer. In contrast, additions to the Al matrix can be reduced to the self-acting formation of a protective layer at the UMo/Al interface. Additions to the UMo to stabilize the γ-UMo upon heating are of minor importance since irradiation reverses the phase decomposition of UMo.

  1. Evaluation of microstructure of irradiated fuel channel components of PHWR

    International Nuclear Information System (INIS)

    Ramadasan, E.

    2005-01-01

    Performance evaluation and failure analysis of irradiated reactor structural components such as those in-core and PHT circuit components necessitate metallographic evaluation using special metallographic specimen preparation techniques due to the radiation dose and contamination levels involved in handling the specimens. The metallographic specimen preparation techniques that are resorted to involve use of fully automatic and semi automatic machines, shielded metallographic microscope and specialised equipment developed for lead-cell metallography. The techniques used and the results obtained in the metallographic studies on irradiated fuel channel components such as pressure tubes and garter springs of various Indian PHWRs at RAPS, NAPS and MAPS are presented as case studies in the paper. The evaluation of oxidation and hydriding behaviour of zircaloy-2 pressure tubes and garter springs are presented. The paper also gives in detail the microstructural evaluation of hydride blistering seen at the PT-CT contact location of the pressure tubes of RAPS-2. The evaluation revealed that the hydride blisters was small compared to their length, unlike the hydride blisters seen in the CANDU pressure tube G-16 of Pickering-2. This could be attributed to be due to the difference in the annulus conditions between the two types of reactors. The hydride blisters in J-07 pressure tube of RAPS-2 had ductile material adjacent to them. The paper also gives the hydride blistering observation on irradiated Zr-2.5% Nb-0.5% Cu garter springs of RAPS-2. It was seen that there was only negligible hydriding of the garter springs during service through they showed presence of benign hydride blisters in them. The general hydriding observations made on the pressure tubes of Indian PHWRs under different conditions are also presented. (author)

  2. Electron Microscopic Examination of Irradiated TRISO Coated Particles of Compact 6-3-2 of AGR-1 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Van Rooyen, Isabella Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riesterer, Jessica Lori [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, Brandon Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott Arden [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2012-12-01

    The electron microscopic examination of selected irradiated TRISO coated particles of the AGR-1 experiment of fuel compact 6-3-2 are presented in this report. Compact 6-3-2 refers to the compact in Capsule 6 at level 3 of Stack 2. The fuel used in capsule 6 compacts, are called the “baseline” fuel as it is fabricated with refined coating process conditions used to fabricate historic German fuel, because of its excellent irradiation performance with UO2 kernels. The AGR-1 fuel is however made of low-enriched uranium oxycarbide (UCO). Kernel diameters are approximately 350 µm with a U-235 enrichment of approximately 19.7%. Compact 6-3-2 has been irradiated to 11.3% FIMA compact average burn-up with a time average, volume average temperature of 1070.2°C and with a compact average fast fluence of 2.38E21 n/cm

  3. Post-irradiation analysis of low enriched U-Mo/Al dispersions fuel miniplate tests, RERTR 4 and 5

    International Nuclear Information System (INIS)

    Hofman, G.L.; Finlay, M.R.; Kim, Y.S.

    2005-01-01

    Interpretation of the post irradiation data of U-Mo/Al dispersion fuel mini plates irradiated in the Advanced Test Reactor to a maximum U-235 burn up of 80% are presented. The analyses addresses fuel swelling and porosity formation as these fuel performance issues relate to fuel fabrication and irradiation parameters. Specifically, mechanisms involved in the formation of porosity observed in the U-Mo/Al interaction phase are discussed and, means of mitigating or eliminating this irradiation phenomenon are offered. (author)

  4. AGR-1 Fuel Compact 6-3-2 Post-Irradiation Examination Results

    Energy Technology Data Exchange (ETDEWEB)

    Paul demkowicz; jason Harp; Scott Ploger

    2012-12-01

    Destructive post-irradiation examination was performed on fuel Compact 6-3-2, which was irradiated in the AGR-1 experiment to a final compact average burnup of 11.3% FIMA and a time-average, volume-average temperature of 1070°C. The analysis of this compact was focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, measurement of fuel burnup by several methods, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy. A single particle with a defective SiC layer was identified during deconsolidation-leach-burn-leach analysis, which is in agreement with previous measurements showing elevated cesium in the Capsule 6 graphite fuel holder associated with this fuel compact. The fraction of the compact europium inventory released from the particles and retained in the matrix was relatively high (approximately 6E-3), indicating release from intact particle coatings. The Ag-110m inventory in individual particles exhibited a very broad distribution, with some particles retaining =80% of the predicted inventory and others retaining less than 25%. The average degree of Ag-110m retention in 60 gamma counted particles was approximately 50%. This elevated silver release is in agreement with analysis of silver on the Capsule 6 components, which indicated an average release of 38% of the Capsule 6 inventory from the fuel compacts. In spite of the relatively high degree of silver release from the particles, virtually none of the Ag-110m released was found in the compact matrix, and presumably migrated out of the compact and was deposited on the irradiation capsule components. Release of all other fission products from the particles appears to be less than a single

  5. KUCA critical experiments using MEU fuel (II)

    International Nuclear Information System (INIS)

    Kanda, Keiji; Hayashi, Masatoshi; Shiroya, Seiji; Kobayashi, Keiji; Fukui, Hiroshi; Mishima, Kaichiro; Shibata, Toshikazu

    1983-01-01

    Due to mutual concerns in the USA and Japan about the proliferation potential of highly-enriched uranium (HEU), a joint study program I was initiated between Argonne National Laboratory (ANL and Kyoto University Research Reactor Institute (KURRI) in 1978. In accordance with the reduced enrichment for research and test reactor (RERTR) program, the alternatives were studied for reducing the enrichment of the fuel to be used in the Kyoto University High Flux Reactor (KUHFR). The KUHFR has a distinct feature in its core configuration it is a coupled-core. Each annular shaped core is light-water-moderated and placed within a heavy water reflector with a certain distance between them. The phase A reports of the joint ANL-KURRI program independently prepared by two laboratories in February 1979, 3,4 concluded that the use of medium-enrichment uranium (MEU, 45%) in the KUHFR is feasible, pending results of the critical experiments in the Kyoto University Critical Assembly (KUCA) 5 and of the burnup test in the Oak Ridge Research Reactor 6 (ORR). An application of safety review (Reactor Installation License) for MEU fuel to be used in the KUCA was submitted to the Japanese Government in March 1980, and a license was issued in August 1980. Subsequently, the application for 'Authorization before Construction' was submitted and was authorized in September 1980. Fabrication of MEU fuel-elements for the KUCA experiments by CERCA in France was started in September 1980, and was completed in March 1981. The critical experiments in the KUCA with MEU fuel were started on a single-core in May 1981 as a first step. The first critical state of the core using MEU fuel was achieved at 312 p.m. in May 12, 1981. After that, the reactivity effects of the outer side-plates containing boron burnable poison were measured. At Munich Meeting in Sept., 1981, we presented a paper on critical mass and reactivity of burnable poison in the MEU core. Since then we carried out the following experiments

  6. KUCA critical experiments using MEU fuel (II)

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, Keiji; Hayashi, Masatoshi; Shiroya, Seiji; Kobayashi, Keiji; Fukui, Hiroshi; Mishima, Kaichiro; Shibata, Toshikazu [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka (Japan)

    1983-09-01

    Due to mutual concerns in the USA and Japan about the proliferation potential of highly-enriched uranium (HEU), a joint study program I was initiated between Argonne National Laboratory (ANL and Kyoto University Research Reactor Institute (KURRI) in 1978. In accordance with the reduced enrichment for research and test reactor (RERTR) program, the alternatives were studied for reducing the enrichment of the fuel to be used in the Kyoto University High Flux Reactor (KUHFR). The KUHFR has a distinct feature in its core configuration it is a coupled-core. Each annular shaped core is light-water-moderated and placed within a heavy water reflector with a certain distance between them. The phase A reports of the joint ANL-KURRI program independently prepared by two laboratories in February 1979, 3,4 concluded that the use of medium-enrichment uranium (MEU, 45%) in the KUHFR is feasible, pending results of the critical experiments in the Kyoto University Critical Assembly (KUCA) 5 and of the burnup test in the Oak Ridge Research Reactor 6 (ORR). An application of safety review (Reactor Installation License) for MEU fuel to be used in the KUCA was submitted to the Japanese Government in March 1980, and a license was issued in August 1980. Subsequently, the application for 'Authorization before Construction' was submitted and was authorized in September 1980. Fabrication of MEU fuel-elements for the KUCA experiments by CERCA in France was started in September 1980, and was completed in March 1981. The critical experiments in the KUCA with MEU fuel were started on a single-core in May 1981 as a first step. The first critical state of the core using MEU fuel was achieved at 312 p.m. in May 12, 1981. After that, the reactivity effects of the outer side-plates containing boron burnable poison were measured. At Munich Meeting in Sept., 1981, we presented a paper on critical mass and reactivity of burnable poison in the MEU core. Since then we carried out the following experiments

  7. Fuel Performance Modeling of U-Mo Dispersion Fuel: The thermal conductivity of the interaction layers of the irradiated U-Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mistarhi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    U-Mo/Al dispersion fuel performed well at a low burn-up. However, higher burn-up and higher fission rate irradiation testing showed enhanced fuel meat swelling which was caused by high interaction layer growth and pore formation. The performance of the dispersion type fuel in the irradiation and un-irradiation environment is very important. During the fabrication of the dispersion type fuel an Interaction Layer (IL) is formed due to the inter-diffusion between the U-Mo fuel particles and the Al matrix which is an intermetallic compound (U,Mo)Alx. During irradiation, the IL becomes amorphous causing a further decrease in the thermal conductivity and an increase in the centerline temperature of the fuel meat. Several analytical models and numerical methods were developed to study the performance of the unirradiated U-Mo/Al dispersion fuel. Two analytical models were developed to study the performance of the irradiated U-Mo/Al dispersion fuel. In these models, the thermal conductivity of the IL was assumed to be constant. The properties of the irradiated U-Mo dispersion fuel have been investigated recently by Huber et al. The objective of this study is to develop a correlation for IL thermal conductivity during irradiation as a function of the temperature and fission density from the experimentally measured thermal conductivity of the irradiated U-Mo/Al dispersion fuel. The thermal conductivity of IL during irradiation was calculated from the experimentally measured data and a correlation was developed from the thermal conductivity of IL as a function of T and fission density.

  8. Individual and collective doses associated with the transport of irradiated magnox fuel within the UK

    International Nuclear Information System (INIS)

    Macdonald, H.F.; Mairs, J.H.

    1978-12-01

    A method is described of evaluating the individual and collective doses arising during the transport of irradiated fuel from a system of nuclear power stations to a central reprocessing plant. The doses associated with irradiated Magnox fuel movements in the UK are estimated and compared with those resulting from other phases of the nuclear fuel cycle. In addition, the individual and collective doses implied by the accidental activity release limits contained within the 1973 IAEA Tranport Regulations are discussed. (author)

  9. A Study on Structural Strength of Irradiated Spacer Grid for PWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Y. G.; Baek, S. J.; Kim, D. S.; Yoo, B. O.; Ahn, S. B.; Chun, Y. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, J. I.; Kim, Y. H.; Lee, J. J. [KEPCO NF, Daejeon (Korea, Republic of)

    2014-10-15

    A fuel assembly consists of an array of fuel rods, spacer grids, guide thimbles, instrumentation tubes, and top and bottom nozzles. In PWR (Pressurized light Water Reactor) fuel assemblies, the spacer grids support the fuel rods by the friction forces between the fuel rods and springs/dimples. Under irradiation, the spacer grids supporting the fuel rods absorb vibration impacts due to the reactor coolant flow, and also bear static and dynamic loads during operation inside the nuclear reactor and transportation for spent fuel storage. Thus, it is important to understand the characteristics of deformation behavior and the change in structural strength of an irradiated spacer grid.. In the present study, the static compression test of a spacer grid was conducted to investigate the structural strength of the irradiated spacer grid in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. To evaluate the structural strength of an irradiated spacer grid, hot cell tests were carried out at IMEF of KAERI. The fuel assembly was dismantled and the irradiated spacer grid was obtained for the compression test. The apparatus for measuring the compression strength of the irradiated spacer grid was developed and installed successfully in the hot cell.

  10. Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

    International Nuclear Information System (INIS)

    Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

    1981-04-01

    Three fuel pin bundles, R-109/1, 2 and 3, were irradiated in a PWR loop in the HFR at Petten during respectively 131, 57 and 57 effective full power days at average powers of approximately 39 kW.m -1 and at peak powers of approximately 60 kW.m -1 . The results of the post-irradiation examinations of these fuel bundles are presented. (Auth.)

  11. Shielding calculations for ships carrying irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Burstall, R.F.; Dean, M.H.

    1983-01-01

    A number of ships have been constructed to carry irradiated fuel from Japan to the UK and France, for reprocessing. About twenty transport flasks may be carried on each voyage. Permanent shielding must be provided on the ships to ensure that no member of the crew receives an annual dose rate greater than a specified limit. As the fuel is of varying type and radiation history, and as flasks of differing designs are used, many calculations are needed. There are a number of difficulties in making shielding calculations for the ships. The geometry is complex, dimensions are large, and considerable air spaces are involved. The paper considers possible methods of calculation. The line-of-sight method is chosen for most of the calculations, for both gamma radiation and neutrons. The basic data which is used in the calculations is described. As the methods of calculation are somewhat approximate, it is necessary to provide confirmation that they are sufficiently accurate. Validation has been provided in two ways. First, measurements have been made on board the ships, and these have been checked against calculation. Second, a simplified model of the flasks and ship has been set up, and calculations checked against more sophisticated methods. Results of the validation checks are presented, and it is shown that adequate accuracy is achieved. (author)

  12. Shielding calculations for ships carrying irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Dean, M.H.

    1985-01-01

    A number of ships have been constructed to carry irradiated fuel from Japan to the U.K. and France, for reprocessing. About 20 transport flasks may be carried on each voyage. Permanent shielding must be provided on the ships to ensure that no member of the crew receives an annual dose greater than a specified limit. As the fuel is of varying type and radiation history, and as flasks of differing designs are used, many shielding calculations are needed. There are a number of difficulties in making shielding calculations for the ships. The geometry is complex, dimensions are large and considerable air spaces are involved. The paper considers possible methods of calculation. The line-of-sight method is chosen for most of the calculations, for both γ-radiation and neutrons. The basic data which is used in the calculations is described. As the methods of calculation are somewhat approximate, it is necessary to provide confirmation that they are sufficiently accurate. Validation has been provided in two ways. First, measurements have been made on board one of the ships, Pacific Crane, and these have been checked against calculation. Second, a simplified model of the flasks and ship has been set up, and calculations checked against more sophisticated methods. Results of the validation checks are presented, and it is shown that adequate accuracy is achieved. (author)

  13. Design, irradiation, and post-irradiation examination of the UC and (U,Pu)C fuel rods of the test groups Mol-11/K1 and Mol-11/K2

    International Nuclear Information System (INIS)

    Freund, D.; Elbel, H.; Steiner, H.

    1976-06-01

    The test groups K1 and K2 of the irradiation experiment Mol-11 are reported. Design, irradiation, and post-irradiation examination of the fuel rods irradiated are described. Mol-11/K1 consisted of one fuel rod with UC of 94% T.D. and helium bonding. This test group was intended to prove the high power irradiation capsule in pile. Mol-11/K2 consists of three fuel rods in total. One of these is presently still in the reactor. In this test group mixed carbide fuel of 83% T.D. and 15% Pu content under helium bonding is irradiated. The fuel rod K2-2 was provided with a capillary tube for the continuous measurement of fission gas pressure built up. 1.4988 stainless steel was chosen as cladding material. The final burnup lies between 35 and 70 MWd/kg M. Post-irradiation examination of the two test groups covers a theoretical analysis of the irradiation behaviour. (orig./GSCH) [de

  14. Irradiation-induced creep in fuel compacts for high-temperature reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Veringa, H; Blackstone, R [Stichting Energieonderzoek Centrum Nederland, Petten; Loelgen, R

    1977-01-01

    Restrained shrinkage experiments at neutron fluences up to 3 x 10/sup 21/ n cm/sup -2/ DNE in the temperature range 600 to 1200/sup 0/C were performed on three different dummy coated-particle fuel compacts in the high-flux reactor at Petten. The data were evaluated to obtain the steady-state radiation creep coefficient of the compacts. It was found that, for the materials investigated, the creep coefficient is temperature dependent, but no clear relationship with Young's modulus could be established. Under certain conditions this irradiation-induced plasticity influences the elastic properties, with the concomitant increase of the creep coefficient. This effect coincides with the formation and further opening up of cracks due to stresses caused by irradiation-induced shrinkage of matrix material.

  15. Irradiation-induced creep in fuel compacts for high-temperature reactor applications

    International Nuclear Information System (INIS)

    Veringa, H.; Blackstone, R.; Loelgen, R.

    1977-01-01

    Restrained shrinkage experiments at neutron fluences up to 3 x 10 21 n cm -2 DNE in the temperature range 600 to 1200 0 C were performed on three different dummy coated-particle fuel compacts in the high-flux reactor at Petten. The data were evaluated to obtain the steady-state radiation creep coefficient of the compacts. It was found that, for the materials investigated, the creep coefficient is temperature dependent, but no clear relationship with Young's modulus could be established. Under certain conditions this irradiation-induced plasticity influences the elastic properties, with the concomitant increase of the creep coefficient. This effect coincides with the formation and further opening up of cracks due to stresses caused by irradiation-induced shrinkage of matrix material. (author)

  16. Simulation of fuel rod irradiation capsules in water loops by electric heater rods

    International Nuclear Information System (INIS)

    Lopez, J.; Montes, M.; Serrano, J.; Haefner, H.E.

    1984-01-01

    The out of pile simulation of irradiation devices was carried out by J.E.N. in the frame of the KfK-JEN joint experiment for irradiation of fast reactor fuel rods (IVO-FR2-Vg7). A typical single-wall-Nak (22% Na, 78% K) electrical heated capsule was fabricated and hydraulical tests were done. The capsule was instrumented with 10 thermocouples in order to obtain the radial temperature profile into the capsule in function of the electrical rod power (max. 215 w/cm), flow rate (max. 2,4 m 3 /h) and coolant temperature (max. 60degC). The experimental values are compared to the Tecap-Code results. (author)

  17. Food irradiation and Hungarian experiences with it

    Energy Technology Data Exchange (ETDEWEB)

    Farkas, Jozsef; Kiss, Istvan

    1988-01-01

    Among the different types of food irradiation techniques using ionizing radiations the methods serving - in a direct or indirect way - the purposes of health protection are reviewed. Based on foreign and Hungarian examples radurization of vegetables (potatoes, onions, garlics) by low irradiation doses, radiosterilization of corn, tropic and subtropic fruits, radicidation of dry additivies (spice, medicinal herb) and foodstuffs of animal origin (meat, poultry, egg, fish, shrimp, etc.) are discussed. The general situation of food irradiation in Hungary inlcuding licensing procedures is evaluated. (V.N.) 32 refs.; 3 figs.

  18. Waste management implications of irradiated nuclear fuel storage

    International Nuclear Information System (INIS)

    Larrick, A.P.; Schneidmiller, D.

    1977-01-01

    Irradiated nuclear fuel is stored underwater in large recirculating basins prior to reprocessing in chemical separations plants. A small number of the fuel rods contain minor defects which allow fission products to seep into the basin water. The predominant radionuclides leached into the water are dependent upon the decay time after removal from the reactor. Freshly discharged fuel releases short half-lived radioiodine which presents exposure and airborne release problems on a short-term basis but does not impose significant long-term waste management problems. After a reasonable decay period, the major radionuclides present are 137 Cs, 144 Ce, 90 Sr, 3 H, and 239 / 240 Pu. Each of the radionuclides imposes specific waste management problems which require individual consideration in their control, packaging and shipment, and permanent disposal. These considerations are reviewed in this paper for general industry practices and specific illustrative examples are presented. Control of radionuclide concentrations and water purity are normally achieved by filtration and ion exchange treatment. Cartridge-type filters are the most common but improved filtration techniques that minimize personnel exposure, particularly during filter replacements, are now being adopted. Mixed bed ion exchange resins are most commonly utilized for soluble radionuclide removal, particularly for basins filled with demineralized water. Cesium-specific exchange media are employed at basins where demineralized water is not employed; these media operate for very long periods of time since they are not depleted by the normal dissolved non-radioactive water impurities. The resins are either buried when depleted or regenerated and the regeneration solutions concentrated for burial. Resin run lengths are usually determined by ionic or radiochemical depletion of the resin or in some cases by limiting radionuclide concentrations specified by shipping regulations or established ion column dose rates

  19. Post irradiation examination of HANARO nucler mini-element fuel (metallographic and density test)

    International Nuclear Information System (INIS)

    Yoo, Byung Ok; Hong, K. P.; Park, D. G.; Choo, Y. S.; Baik, S. J.; Kim, K. H.; Kim, H. C.; Jung, Y. H.

    2001-05-01

    The post irradiation examination of a HANARO mini-element nuclear fuel, KH96C-004, was done in June 6, 2000. The purpose of this project is to evaluate the in-core performance and reliability of mini-element nuclear fuel for HANARO developed by the project T he Nuclear Fuel Material Development of Research Reactor . And, in order to examine the performance of mini-element nuclear fuel in normal output condition, the post irradiation examination of a nuclear fuel bundle composed by 6 mini nuclear fuel rods and 12 dummy fuel rods was performed. Based on these examination results, the safety and reliability of HANARO fuel and the basic data on the design of HANARO nuclear fuel can be ensured and obtained,

  20. The Thai multipurpose food irradiation and experience in food irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Banditsing, C; Pringsulka, V; Sutantawong, M [and others

    1986-12-31

    Losses of agricultural produce in Thailand are due to the hot climate accelerating the ripening of fruits and sprouting of vegetables, spoilage microorganisms, pathogenic microorganisms, and insect infestation. Losses amount to as much as 30%. Onion, garlic and potato which have a short shelf-life, for instance, cannot be stored long enough for off-season domestic consumption. The annual