WorldWideScience

Sample records for fuel samples irradiated

  1. 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

  2. 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)

  3. 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)

  4. Recent activity on the post-irradiation analyses of nuclear fuels and actinide samples at JAERI

    International Nuclear Information System (INIS)

    Shinohara, Nobuo; Nakahara, Yoshinori; Kohno, Nobuaki; Tsujimoto, Kazufumi

    2003-01-01

    Radiochemical analyses of spent fuels have been carried out at JAERI for contributing to the development of nuclear technologies, where several samples from research reactors and nuclear power plants were analyzed to obtain isotopic compositions and burnups. The history and procedures of the radiochemical analyses are depicted and some recent results are given in this paper. (author)

  5. Irradiation of mixed UO2-PuO2 oxide samples for fast neutron reactor fuel elements

    International Nuclear Information System (INIS)

    Mikailoff, H.; Mustelier, J.; Bloch, J.; Conte, M.; Hayet, L.; Lauthier, J.C.; Leclere, J.

    1968-01-01

    Thermal flux irradiation testings of small mixed oxide pellets UPuO 2 fuel elements were performed in support of the fuel reference design for the Phenix fast reactor. The effects of different parameters (stoichiometry, pellet density, pellet clad gap). on the behaviour of the oxide (temperature distribution, microstructural changes, fission gas release) were investigated in various irradiation conditions. In particular, the effect of fuel density decrease and power rate increase on thermal performances were determined on short term irradiations of porous fuels. (authors) [fr

  6. 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

  7. 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.)

  8. 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

  9. 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)

  10. Benchmark reference data on post irradiation analysis of light water reactor fuel samples

    International Nuclear Information System (INIS)

    Guardini, S.; Guzzi, G.

    1983-01-01

    The structure of the present report is as follows: in section I the benchmark activity (BM) is described in detail; characteristics of the reactors and fuel assemblies examinated are given, and the technical aspects of the chemical and analytical processes are discussed. In section II all the techniques used to certify the analytical data are presented, together with a discussion of evaluated random and systematic uncertainties. A comparison with the calculated values and the interpretation with ICT (Isotopic Correlation Techniques) is also presented in this section. Section III presents the results. In practice the complete sets of results referring to all JRC measurements are given here for the sake of the completeness and consistency of this final report

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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)

  16. Temperature measurement of the reactor materials samples irradiated in the fuel channels of the RA reactor - Annex 16

    International Nuclear Information System (INIS)

    Nikolic, M.; Djalovic, M.

    1964-01-01

    Reactor materials as graphite, stainless steel, magnox, zirconium alloys, etc. were exposed to fast neutron flux inside the fuel elements specially adapted for this purpose. Samples in the form ampoules were placed in capsules inside the fuel channels and cooled by heavy water which cools the fuel elements. In order to monitor the samples temperature 42 thermocouples were placed in the samples. That was necessary for reactor safety reasons and for further interpretation of measured results. Temperature monitoring was done continuously by multichannel milivoltmeters. This paper describes the technique of introducing the thermocouples, compensation instruments, control of the cold ends and adaptation of the instruments for precision (0.5%) temperature measurement in the range 30 deg - 130 deg C; 30 deg - 280 deg C and 30 deg - 80 deg C [sr

  17. 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

  18. 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.)

  19. 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)

  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. 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.)

  2. 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

  3. 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)

  4. 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

  5. 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.)

  6. 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)

  7. 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

  8. Tests on CANDU fuel elements sheath samples

    International Nuclear Information System (INIS)

    Ionescu, S.; Uta, O.; Mincu, M.; Prisecaru, I.

    2016-01-01

    This work is a study of the behavior of CANDU fuel elements after irradiation. The tests are made on ring samples taken from fuel cladding in INR Pitesti. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory. By metallographic and ceramographic examination we determinate that the hydride precipitates are orientated parallel to the cladding surface. A content of hydrogen of about 120 ppm was estimated. After the preliminary tests, ring samples were cut from the fuel rod, and were subject of tensile test on an INSTRON 5569 model machine in order to evaluate the changes of their mechanical properties as consequence of irradiation. Scanning electron microscopy was performed on a microscope model TESCAN MIRA II LMU CS with Schottky FE emitter and variable pressure. The analysis shows that the central zone has deeper dimples, whereas on the outer zone, the dimples are tilted and smaller. (authors)

  9. 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.

  10. 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

  11. 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)

  12. 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%.

  13. 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)

  14. 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

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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

  1. 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)

  2. 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)

  3. 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

  4. 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)

  5. 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)

  6. Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

    International Nuclear Information System (INIS)

    Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V.; Boulyga, S.F.; Becker, J.S.

    2005-01-01

    An analytical method is described for the estimation of uranium concentrations, of 235 U/ 238 U and 236 U/ 238 U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10 -9 g/g to 2.0 x 10 -6 g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing 235 U/ 238 U and 236 U/ 238 U isotope ratios and the average value amounted to 9.4±0.3 MWd/(kg U). (orig.)

  7. 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).

  8. 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)

  9. 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

  10. 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

  11. Spent nuclear fuel sampling strategy

    International Nuclear Information System (INIS)

    Bergmann, D.W.

    1995-01-01

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

  12. 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

  13. 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

  14. FY 2013 Summary Report: Post-Irradiation Examination of Zircaloy-4 Samples in Target Capsules and Initiation of Bending Fatigue Testing for Used Nuclear Fuel Vibration Integrity Investigations

    Energy Technology Data Exchange (ETDEWEB)

    Howard, Richard H [ORNL; Yan, Yong [ORNL; Wang, Jy-An John [ORNL; Ott, Larry J [ORNL; Howard, Rob L [ORNL

    2013-10-01

    This report documents ongoing work performed at Oak Ridge National Laboratory (ORNL) for the Department of Energy, Office of Fuel Cycle Technology Used Fuel Disposition Campaign (UFDC), and satisfies the deliverable for milestone M2FT-13OR0805041, “Data Report on Hydrogen Doping and Irradiation in HFIR.” This work is conducted under WBS 1.02.08.05, Work Package FT-13OR080504 ST “Storage and Transportation-Experiments – ORNL.” The objectives of work packages that make up the S&T Experiments Control Account are to conduct the separate effects tests (SET) and small-scale tests that have been identified in the Used Nuclear Fuel Storage and Transportation Data Gap Prioritization (FCRD-USED-2012-000109). In FY 2013, the R&D focused on cladding and container issues and small-scale tests as identified in Sections A-2.9 and A-2.12 of the prioritization report.

  15. 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.)

  16. 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

  17. 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

    for dimensional changes, blisters, or any other visible defect. After inspection the experiment is reassembled and reinserted into the reactor for further irradiation. At completion of the specified in reactor exposure of the miniplates, the experiment will be disassembled, and the irradiated samples will be returned to the respective fuel plate fabricators for detailed postirradiation examination. Ultimately, fuel plate types with suitable characteristics will be manufactured into full-sized plate-type fuel elemental suitable for testing in the ORR. (author)

  18. 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

  19. 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)

  20. 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)

  1. 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

  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. 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

  4. 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.

  5. 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

  6. 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

  7. Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

    Energy Technology Data Exchange (ETDEWEB)

    Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V. [Inst. of Radiobiology, Minsk Univ. (Belarus); Boulyga, S.F. [Inst. of Inorganic Chemistry and Analytical Chemistry, Johannes Gutenberg-Univ. Mainz, Mainz (Germany); Becker, J.S. [Central Div. of Analytical Chemistry, Research Centre Juelich, Juelich (Germany)

    2005-07-01

    An analytical method is described for the estimation of uranium concentrations, of {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10{sup -9}g/g to 2.0 x 10{sup -6}g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and the average value amounted to 9.4{+-}0.3 MWd/(kg U). (orig.)

  8. 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)

  9. 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.

  10. 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)

  11. 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

  12. 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.

  13. 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.

  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. Tube closure device, especially for sample irradiation

    International Nuclear Information System (INIS)

    Klahn, F.C.; Nolan, J.H.; Wills, C.

    1979-01-01

    Device for closing the outlet of a bore and temporarily locking components in this bore. Specifically, it concerns a device for closing a tube containing a set of samples for monitoring irradiation in a nuclear reactor [fr

  17. 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

  18. 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)

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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.

  4. 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)

  5. 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)

  6. 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

  7. 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

  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. Irradiation chamber and sample changer for biological samples

    International Nuclear Information System (INIS)

    Kraft, G.; Daues, H.W.; Fischer, B.; Kopf, U.; Liebold, H.P.; Quis, D.; Stelzer, H.; Kiefer, J.; Schoepfer, F.; Schneider, E.

    1980-01-01

    This paper describes an irradiaton system with which living cells of different origin are irradiated with heavy ion beams (18 <- Z <- 92) at energies up to 10 MeV/amu. The system consists of a beam monitor connected to the vacuum system of the accelerator and the irradiation chamber, containing the biological samples under atmospheric pressure. The requirements and aims of the set up are discussed. The first results with saccharomyces cerevisiae and Chinese Hamster tissue cells are presented. (orig.)

  10. 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

  11. 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)

  12. 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)

  13. 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)

  14. 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)

  15. 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

  16. 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)

  17. 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.

  18. 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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. 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)

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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)

  10. 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)

  11. SEM Characterization of an Irradiated Monolithic U-10Mo Fuel Plate

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B.

    2010-01-01

    Results of scanning electron microscopy (SEM) characterization of irradiated U-7Mo dispersion fuel plates with differing amounts of matrix Si have been reported. However, to date, no results of SEM analysis of irradiated U-Mo monolithic fuel plates have been reported. This paper describes the first SEM characterization results for an irradiated monolithic U-10Mo fuel plate. Two samples from this fuel plate were characterized. One sample was produced from the low-flux side of the fuel plate, and another was produced at the high-flux side of the fuel plate. This characterization focused on the microstructural features present at the U-10Mo foil/cladding interface, particularly the interaction zone that had developed during fabrication and irradiation. In addition, the microstructure of the foil itself was investigated, along with the morphology of the observed fission gas bubbles. It was observed that a Si-rich interaction layer was present at the U-10Mo foil/cladding interface that exhibited relatively good irradiation behavior, and within the U-10Mo foil the microstructural features differed in some respects from what is typically seen in the U-Mo powders of an irradiated dispersion fuel.

  12. SEM characterization of an irradiated monolithic U-10Mo fuel plate

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B.; Finlay, M.R.

    2010-01-01

    Results of scanning electron microscopy (SEM) characterization of irradiated U-7Mo dispersion fuel plates with differing amounts of matrix Si have been reported. However, to date, no results of SEM analysis of irradiated U-Mo monolithic fuel plates have been reported. This paper describes the first SEM characterization results for an irradiated monolithic U-10Mo fuel plate. Two samples from this fuel plate were characterized. One sample was produced from the low-flux side of the fuel plate, and another was produced at the high-flux side of the fuel plate. This characterization focused on the microstructural features present at the U-10Mo foil/AA6061 cladding interface, particularly the interaction zone that had developed during fabrication and any continued development during irradiation. In addition, the microstructure of the foil itself was investigated, along with the morphology of the observed fission gas bubbles. It was observed that a Si-rich interaction layer was present at the U-10Mo foil/cladding interface that exhibited relatively good irradiation behavior, and within the U-10Mo foil the microstructural features differed in some respects from what is typically seen in the U-7Mo powders of an irradiated dispersion fuel. (author)

  13. Device for sampling HTGR recycle fuel particles

    International Nuclear Information System (INIS)

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

    1977-03-01

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

  14. 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

  15. 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)

  16. 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)

  17. 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)

  18. 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

  19. 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)

  20. 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.

  1. 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.

  2. 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.

  3. 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)

  4. 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.)

  5. 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

  6. 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)

  7. 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)

  8. 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)

  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. Inert medium (helium) irradiation testing of pressure tube samples

    International Nuclear Information System (INIS)

    Ancuta, M.; Radu, V.; Stefan, V.; Preda, M.

    2001-01-01

    Irradiation tests currently performed in C-5 capsule aim at obtaining data and information concerning behavior to irradiation of pressure tubes of CANDU type fuel channel, to evidence the factors limiting operation life span. A calculation code for analysis and prediction of pressure tube behavior should be based upon periodical inspection results, post irradiation examination of the removed from reactor pressure tubes as well as on the experimental results obtained with materials subjected to irradiation conditions identical with the operational ones. Mechanical behavior analysis should focus both complex thermal-mechanical type stresses and mechanical properties alteration under irradiation. The experimental results should be applied: - to evaluate the irradiation effects upon mechanical properties of Zr-2.5% Nb exposed to fluences up to 10 21 n·cm -2 ; - to gather data concerning the real stress / real deformation characteristic from which characteristic quantities can be deduced as, for instance, elasticity modulus, plasticity modulus, exponent of stress term in the Tsu-Berteles relation, to be used within the CANTUP simulation code describing pressure tube behavior, currently developed at INR Pitesti; - to develop prediction methods of pressure tube behavior and merging with in-service inspection procedure in order to forecast the life span and the proper timing for replacement before major failures occur. The samples irradiated in C-5 capsule were extracted from the ends of Zr-2.5% Nb pressure tubes resulting from Cernavoda NPP Unit 1. The samples for tensile tests were extracted on longitudinal and transversal directions of the pressure tube. The tests were carried out under following conditions: - test environment temperature, 260 - 280 deg.C; - testing medium, helium at 1 - 6 b pressure; - neutron flux (E n > 1 MeV), 1 - 2 · 10 13 ncm -2 s -1 ; - neutron fluence (E n > 1 MeV), 4 · 10 20 ncm -2 . The following characteristics were obtained from tensile

  11. Application of FE-SEM with elemental analyzer for irradiated fuel materials

    International Nuclear Information System (INIS)

    Sasaki, Shinji; Maeda, Koji; Yamada, A.

    2012-01-01

    It is important to study the irradiation behavior of the uranium-plutonium mixed oxide fuels (MOX fuels) for development of fast reactor fuels. During irradiation in a fast reactor, the changes of microstructures and the changes of element distributions along radial direction occur in the MOX fuels because of a radial temperature gradient. In order to make detailed observations of microstructure and elemental analyses of fuel samples, a field emission scanning electron microscope (FE-SEM) equipped with a wavelength-dispersive X-ray spectrometer (WDX) and an energy-dispersive X-ray spectrometer (EDX) were installed in a hot laboratory. Because fuel samples have high radioactivities and emit α-particles, the instrument was modified correspondingly. The notable modified points were as follows. 1) To prevent leakage of radioactive materials, the instrument was attached to a remote control air-tight sample transfer unit between a shielded hot cell and the FE-SEM. 2) To protect operators and the instruments from radiation, the FE-SEM was installed in a lead shield box and the control unit was separately located outside the box. After the installation, the microscopy and elemental analyses were made on low burnup fuel samples. High resolution images were obtained on the fuel sample surface. The characteristic X-rays (U, Pu) emitted from the fuel sample surface measured along radial direction successfully. Thereby, it was able to grasp the change of U, Pu radial distribution after irradiation. The technique has the great advantage of being able to evaluate the changes of microstructures and the changes of element distributions of MOX fuels due to irradiation. In future work, samples of even higher radioactivity will be observed and analyzed. (author)

  12. 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)

  13. 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

  14. Characterization of fission gas bubbles in irradiated U-10Mo fuel

    Energy Technology Data Exchange (ETDEWEB)

    Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.; Buck, Edgar C.

    2017-09-01

    Irradiated U-10Mo fuel samples were prepared with traditional mechanical potting and polishing methods with in a hot cell. They were then removed and imaged with an SEM located outside of a hot cell. The images were then processed with basic imaging techniques from 3 separate software packages. The results were compared and a baseline method for characterization of fission gas bubbles in the samples is proposed. It is hoped that through adoption of or comparison to this baseline method that sample characterization can be somewhat standardized across the field of post irradiated examination of metal fuels.

  15. Microstructure of the irradiated U 3Si 2/Al silicide dispersion fuel

    Science.gov (United States)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J.-F.; Robinson, A. B.; Madden, J. W.; Medvedev, P. G.; Wachs, D. M.

    2011-12-01

    The silicide dispersion fuel of U 3Si 2/Al is recognized as the best performance fuel for many nuclear research and test reactors with up to 4.8 gU/cm 3 fuel loading. An irradiated U 3Si 2/Al dispersion fuel ( 235U ˜ 75%) from the high-flux side of a fuel plate (U0R040) from the Reduced Enrichment for Research and Test Reactors (RERTR)-8 test was characterized using transmission electron microscopy (TEM). The fuel was irradiated in the Advanced Test Reactor (ATR) for 105 days. The average irradiation temperature and fission density of the U 3Si 2 fuel particles for the TEM sample are estimated to be approximately 110 °C and 5.4 × 10 27 f/m 3. The characterization was performed using a 200-kV TEM. The U/Si ratio for the fuel particle and (Si + Al)/U for the fuel-matrix-interaction layer are approximately 1.1 and 4-10, respectively. The estimated average diameter, number density and volume fraction for small bubbles (<1 μm) in the fuel particle are ˜94 nm, 1.05 × 10 20 m -3 and ˜11%, respectively. The results and their implication on the performance of the U 3Si 2/Al silicide dispersion fuel are discussed.

  16. 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)

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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.)

  2. 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)

  3. 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)

  4. 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

  5. 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

  6. 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)

  7. 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)

  8. 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)

  9. 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.

  10. 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)

  11. 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.

  12. 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

  13. 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.

  14. 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)

  15. 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)

  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. 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

  18. Calorimetric assay of HTGR fuel samples

    International Nuclear Information System (INIS)

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

    1979-04-01

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

  19. 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

  20. 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

  1. 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

  2. Transmission electron microscopy characterization of irradiated U-7Mo/Al-2Si dispersion fuel

    International Nuclear Information System (INIS)

    Gan, J.; Keiser, D.D.; Wachs, D.M.; Robinson, A.B.; Miller, B.D.; Allen, T.R.

    2010-01-01

    The plate-type dispersion fuels, with the atomized U(Mo) fuel particles dispersed in the Al or Al alloy matrix, are being developed for use in research and test reactors worldwide. It is found that the irradiation performance of a plate-type dispersion fuel depends on the radiation stability of the various phases in a fuel plate. Transmission electron microscopy was performed on a sample (peak fuel mid-plane temperature ∼109 deg. C and fission density ∼4.5 x 10 27 f m -3 ) taken from an irradiated U-7Mo dispersion fuel plate with Al-2Si alloy matrix to investigate the role of Si addition in the matrix on the radiation stability of the phase(s) in the U-7Mo fuel/matrix interaction layer. A similar interaction layer that forms in irradiated U-7Mo dispersion fuels with pure Al matrix has been found to exhibit poor irradiation stability, likely as a result of poor fission gas retention. The interaction layer for both U-7Mo/Al-2Si and U-7Mo/Al fuels is observed to be amorphous. However, unlike the latter, the amorphous layer for the former was found to effectively retain fission gases in areas with high Si concentration. When the Si concentration becomes relatively low, the fission gas bubbles agglomerate into fewer large pores. Within the U-7Mo fuel particles, a bubble superlattice ordered as fcc structure and oriented parallel to the bcc metal lattice was observed where the average bubble size and the superlattice constant are 3.5 nm and 11.5 nm, respectively. The estimated fission gas inventory in the bubble superlattice correlates well with the fission density in the fuel.

  3. Analysis of irradiated Magnox fuel for fission product and actinide elements

    International Nuclear Information System (INIS)

    Foster, E.; Fudge, A.J.; Glover, K.M.; Wiltshire, R.A.P.

    1980-03-01

    This report describes the methods used for, and the results obtained from, chemical and isotopic analysis of four Magnox fuel samples of known irradiation history. The analysis of this material was carried out in conjunction with CEGB, Berkeley Nuclear Laboratories and BNFL. The data obtained was used to evaluate the accuracy of fuel composition as predicted by computer codes such as RICE (Reactor Inventory Code). (author)

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-04-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  7. 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.

  8. Irradiation project of SiC/SiC fuel pin 'INSPIRE': Status and future plan

    International Nuclear Information System (INIS)

    Kohyama, Akira; Kishimoto, Hirotatsu

    2015-01-01

    After the March 11 Disaster in East-Japan, Research and Development towards Ensuring Nuclear Safety Enhancement for LWR becomes a top priority R and D in nuclear energy policy of Japan. The role of high temperature non-metallic materials, such as SiC/SiC, is becoming important for the advanced nuclear reactor systems. SiC fibre reinforced SiC composite has been recognised to be the most attractive option for the future, now, METI fund based project, INSPIRE, has been launched as 5-year termed project at OASIS in Muroran Institute of Technology aiming at early realisation of this system. INSPIRE is the irradiation project of SiC/SiC fuel pins aiming to accumulate material, thermal, irradiation effect data of NITE-SiC/SiC in BWR environment. Nuclear fuel inserted SiC/SiC fuel pins are planned to be installed in the Halden reactor. The project includes preparing the NITE-SiC/SiC tubes, joining of end caps, preparation of rigs to control the irradiation environment to BWR condition and the instruments to measure the condition of rigs and pins in operation. Also, basic neutron irradiation data will be accumulated by SiC/SiC coupon samples currently under irradiation in BR2. The output from this project may present the potentiality of NITE-SiC/SiC fuel cladding with the first stage fuel-cladding interaction. (authors)

  9. Study of candu fuel elements irradiated in a nuclear power plant

    International Nuclear Information System (INIS)

    Ionescu, S.; Uta, O.; Mincu, M.; Anghel, D.; Prisecaru, I.

    2015-01-01

    The object of this work is the behaviour of CANDU fuel elements after service in nuclear power plant. The results are analysed and compared with previous result obtained on unirradiated samples and with the results obtained on samples irradiated in the TRIGA reactor of INR Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor, the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) of INR Pitesti on samples from fuel elements after they were removed out of the nuclear power plant: - dimensional and macrostructural characterization; - microstructural characterization by metallographic analyses; - determination of mechanical properties; - fracture surface analysis by scanning electron microscopy (SEM). A full set of non-destructive and destructive examinations concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the cladding was performed. The obtained results are typical for CANDU 6-type fuel. The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for the improvement of the CANDU fuel. (authors)

  10. 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

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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)

  16. 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)

  17. 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)

  18. 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

  19. 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)

  20. 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

  1. 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.

  2. 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

  3. Proposal for a project to irradiation of fuel rods in hydraulic rabbit loop of R-2 reactor, Studsvik, Sweden

    International Nuclear Information System (INIS)

    Almeida, M.R.; Ambrozio, F.; Carvalho, S.R.B.

    1974-01-01

    It is planning to perform irradiation test on 21 little samples during a short time. These samples are formed by fuel rods covered with zircaloy 4. It contains UO sub(2) natural pellets and UO sub(2) enriched pellets. The type of fuel and the effects of some parameters, such as diametral gap, irradiation time, power liner density, thermal treatment, over the performance of the samples, will be studied. The parameters variations will be measured by pre and post-irradiation tests. Computer codes has been used in order to simulate the samples performance. The safety test performance via those simulations leads to conclusions that irradiation does not cause damage to integrity of the fuel rods. (M.I.)

  4. 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.

  5. Nondestructive examination of Oconee 1 fuel assemblies after four cycles of irradiation

    International Nuclear Information System (INIS)

    Pyecha, T.D.; Mayer, J.T.; Guthrie, B.A. III; Riordan, J.E.

    1980-12-01

    Five B and W Mark B (15 x 15) pressurized water reactor fuel assemblies were nondestructively examined after four cycles of irradiation in the Oconee 1 reactor. Four of the five assemblies examined had a burnup of 40,000 MWd/mtU; the fifth assembly had a burnup of 36,800 MWd/mtU. This effort is part of a Department of Energy program to improve uranium utilization by extending the burnup of light water reactor fuel. The examinations were conducted in the Oconee 1 and 2 spent fuel storage pool. Data obtained included fuel assembly and fuel rod dimensions, water channel spacings, spacer grid and holddown spring forces, fuel column stack and axial gap lengths, and crud samples. The results indicate that the assemblies performed well through four cycles of operation; all of the data were within design limits

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    Energy Technology Data Exchange (ETDEWEB)

    Collette, R. [Colorado School of Mines, Nuclear Science and Engineering Program, 1500 Illinois St, Golden, CO 80401 (United States); King, J., E-mail: kingjc@mines.edu [Colorado School of Mines, Nuclear Science and Engineering Program, 1500 Illinois St, Golden, CO 80401 (United States); Buesch, C. [Oregon State University, 1500 SW Jefferson St., Corvallis, OR 97331 (United States); Keiser, D.D.; Williams, W.; Miller, B.D.; Schulthess, J. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2016-07-15

    The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. The results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program. - Highlights: • Automated image processing is used to extract fission gas bubble data from irradiated U−Mo fuel samples. • Verification and validation tests are performed to ensure the algorithm's accuracy. • Fission bubble parameters are predictably difficult to compare across samples of varying compositions. • The 2-D results suggest the need for more homogenized fuel sampling in future studies. • The results also demonstrate the value of 3-D reconstruction techniques.

  14. 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.

  15. New experimental space for irradiating samples by RA reactor fast neutron flux at temperatures up to 100 deg C

    International Nuclear Information System (INIS)

    Pavicevic, M.; Novakovic, M.; Zecevic, V.

    1961-01-01

    The objective of this paper is to present adaptation of the RA reactor which would enable samples irradiation by fast neutrons and describe new experimental possibilities. New experimental space was achieved using hollow fuel elements which have been reconstructed to enable placement of irradiation capsules inside the tube. This paper includes thermal analysis and describes problems related to operation, safety and radiation protection issues which arise from using reconstructed fuel elements

  16. Determination of burn-up of irradiated nuclear fuels using mass spectrometry

    International Nuclear Information System (INIS)

    Jagadish Kumar, S.; Telmore, V.M.; Shah, R.V.; Sasi Bhushan, K.; Paul, Sumana; Kumar, Pranaw; Rao, Radhika M.; Jaison, P.G.

    2017-01-01

    Burn-up defined as the atom percent fission, is a vital parameter used for assessing the performance of nuclear fuel during its irradiation in the reactor. Accurate data on the actinide isotopes are also essential for the reliable accountability of nuclear materials and for nuclear safeguards. Both destructive and non-destructive methods are employed in the post-irradiation analysis for the burn-up measurements. Though non-destructive methods are preferred from the point view of remote handling of irradiated fuels with high radioactivity, they do not provide the high accuracy as achieved by the chemical analysis methods. Thus destructive radiochemical and chemical analyses are still the established reference methods for accurate and reliable burn-up determination of irradiated nuclear fuels. In the destructive method, burn-up of irradiated nuclear fuel is determined by correlating the amount of a fission product formed during irradiation with that of heavy elements. Thus the destructive experimental determination of burn-up involves the dissolution of irradiated fuel samples followed by the separation and determination of heavy elements and fission product(s) to be used as burn-up monitor(s). Another approach for the experimental determination of burn-up is based on the changes in the abundances of the heavy element isotopes. A widely accepted method for burn-up determination is based on stable "1"4"8Nd and "1"3"9La as burn-up monitors. Several properties such as non-volatility, nearly same yields for thermal fissions of "2"3"5U and "2"3"9Pu etc justifies the selection of "1"4"8Nd as a burn-up monitor

  17. 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)

  18. 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

  19. 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)

  20. Examination of the creep behaviour of ceramic fuel elements under neutron irradiation

    International Nuclear Information System (INIS)

    Brucklacher, D.

    1978-01-01

    This paper examines the creeping of UO 2 , UO 2 -PuO 2 and UN under neutron irradiation. It starts with the experimental results about the relation between the thermal creep rate and the load, the temperature, as well as characteristic material values, stoichiometry, grain size and porosity. These correlation are first qualitatively discussed and then compared with the statements of actual quantitative equations. From the models and theories on which these equations are based a modified Nabarro-Heering-equation results for the correlation between the creep rate of ceramic fuels, stress, temperature and the fission rate. In the experimental part of the examination, length-changes of creep samples of UO 2 , (U,Pu)O 2 and UN were measured in specially developed irradiation creep casings in different reactors. The measuring data were corrected and evaluated considering the thermal expansion effects, irregular temperature distribution and swelling effects in such a way that the dependences of the creep rate of UO 2 , UO 2 -PuO 2 and UN under irradiation on stress, temperature, fission rate, burn-up and porosity is obtained. It shows that creeping of fuels under irradiation at high temperatures is equivalent to thermally activated creeping, while at low temperature the creep rate induced by irradiation is much higher than the condition without irradiation. The increment of oxidic nuclear fuels is greater than in UN, the stress dependence on low burn-up is proportional in both cases, and the influence of temperature is quite small. (orig.) [de

  1. 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

  2. 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

  3. 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)

  4. 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

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

    International Nuclear Information System (INIS)

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

    1974-01-01

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

  6. 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.

  7. Use of TRIGA flip fuel for improved in-core irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Whittemore, W L [General Atomic Co., San Diego, CA (United States)

    1974-07-01

    Use of standard TRIGA fuel (20% enriched uranium) in a reactor provides a suitable facility for in-core irradiations. However, large numbers of in-core samples irradiated for long periods (many months) can be handled more economically with a TRIGA loaded with FLIP fuel. As an example, ten or more in-core thermionic devices (each worth 50 to 80 cents with respect to a water-filled position) were irradiated in the Mark III TRIGA at General Atomic Company for 18 months with only a modest change in excess reactivity due to core burnup. A core loading of FLIP fuel has been added to the General Atomic Mark F reactor in order to provide numerous in-core irradiation sites for the production of radioisotopes. Since the worth of a 500-gram sample of a molybdenum compound (used for the production of {sup 99}Mo) is about 25 to 50 cents with respect to a water-filled position, use of a FLIP- TRIGA core will permit the irradiation of more than 5 kilograms of a molybdenum compound. A procedure is under development for the production of {sup 99}Mo with relatively high specific activity. Several techniques to concentrate {sup 99}Mo have been tested experimentally. The results will be reported. (author)

  8. 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...

  9. 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)

  10. Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Brémier, S., E-mail: stephan.bremier@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Inagaki, K. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Capriotti, L.; Poeml, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Ogata, T.; Ohta, H. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany)

    2016-11-15

    The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15–20 μm and migration of cladding elements to the fuel. - Highlights: • Electron Probe MicroAnalysis of a UPuZr metallic fuel alloy irradiated to 7.0 at.% burn-up. • Significant redistribution of the fuel components along the fuel radius, nearly complete depletion of Zr in the central part of the fuel. • Interactions between the fuel and the cladding with occurrence of a number of intermediary layers and migration of cladding elements to the fuel. • Safe irradiation behaviour of the base alloy fuel.

  11. Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

    Science.gov (United States)

    Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

    2002-12-01

    This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium).

  12. 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)

  13. 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

  14. Development of a hot cell for post-irradiation analysis of nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Selma S.C.; Silva Junior, Silverio Ferreira da; Loureiro, Joao Roberto M. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: selmasallam@yahoo.com.br, e-mail: silvasf@cdtn.br, e-mail: jrmattos@cdtn.br

    2009-07-01

    Post irradiation examinations of nuclear fuels are performed in order to verify their in-service behavior. Examinations are conducted in compact structures called hot cells, designed to attend the different types of tests and analysis for fuel's characterization. The characterization of fuel microstructure is an activity performed in hot cells. Usually, hot cells for microstructural fuel analysis are designed to allow the metallographic and ceramographic samples preparation and after that, microscopical analysis of the fuel's microstructure. Due to the complexity of the foreseen operations, the severe limitations imposed by the available space into the hot cells, the capabilities of the remote manipulation devices, the procedures of radiological protection and the needs to obtain samples with an adequate surface quality for microscopic analysis, the design of a hot cell for fuel samples preparation presents a high level of complexity. In this paper, the methodology used to develop a hot cell facility for nuclear fuel's metallographic and ceramographic samples preparation is presented. Equipment, devices and systems used in conventional sample preparation processes were evaluated during bench tests. After the necessary adjustments and processes adaptations, they were assembled in a mock-up of the respective hot cell, where they were tested in conditions as realistic as possible, in order to improve the operations and processes to be performed at the real hot cells. (author)

  15. Development of a hot cell for post-irradiation analysis of nuclear fuels

    International Nuclear Information System (INIS)

    Silva, Selma S.C.; Silva Junior, Silverio Ferreira da; Loureiro, Joao Roberto M.

    2009-01-01

    Post irradiation examinations of nuclear fuels are performed in order to verify their in-service behavior. Examinations are conducted in compact structures called hot cells, designed to attend the different types of tests and analysis for fuel's characterization. The characterization of fuel microstructure is an activity performed in hot cells. Usually, hot cells for microstructural fuel analysis are designed to allow the metallographic and ceramographic samples preparation and after that, microscopical analysis of the fuel's microstructure. Due to the complexity of the foreseen operations, the severe limitations imposed by the available space into the hot cells, the capabilities of the remote manipulation devices, the procedures of radiological protection and the needs to obtain samples with an adequate surface quality for microscopic analysis, the design of a hot cell for fuel samples preparation presents a high level of complexity. In this paper, the methodology used to develop a hot cell facility for nuclear fuel's metallographic and ceramographic samples preparation is presented. Equipment, devices and systems used in conventional sample preparation processes were evaluated during bench tests. After the necessary adjustments and processes adaptations, they were assembled in a mock-up of the respective hot cell, where they were tested in conditions as realistic as possible, in order to improve the operations and processes to be performed at the real hot cells. (author)

  16. Container for gaseous samples for irradiation at accelerators

    International Nuclear Information System (INIS)

    Kupsch, H.; Riemenschneider, J.; Leonhardt, J.

    1985-01-01

    The invention concerns a container for gaseous samples for the irradiation at accelerators especially to generate short-lived radioisotopes. The container is also suitable for storage and transport of the target gas and can be multiply reused

  17. 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)

  18. 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

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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

  4. 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

  5. Burn-up determination of irradiated thoria samples by isotope dilution-thermal ionisation mass spectrometry

    International Nuclear Information System (INIS)

    Aggarwal, S.K.; Jaison, P.G.; Telmore, V.M.; Shah, R.V.; Sant, V.L.; Sasibhushan, K.; Parab, A.R.; Alamelu, D.

    2010-03-01

    Burn-up was determined experimentally using thermal ionization mass spectrometry for two samples from ThO 2 bundles irradiated in KAPS-2. This involved quantitative dissolution of the irradiated fuel samples followed by separation and determination of Th, U and a stable fission product burn-up monitor in the dissolved fuel solution. Stable fission product 148 Nd was used as a burn-up monitor for determining the number of fissions. Isotope Dilution-Thermal Ionisation Mass Spectrometry (ID-TIMS) using natural U, 229 Th and enriched 142 Nd as spikes was employed for the determination of U, Th and Nd, respectively. Atom % fission values of 1.25 ± 0.03 were obtained for both the samples. 232 U content in 233 U determined by alpha spectrometry was about 500 ppm and this was higher by a factor of 5 compared to the theoretically predicted value by ORIGEN-2 code. (author)

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

    International Nuclear Information System (INIS)

    Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

    2009-01-01

    The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

  11. 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

  12. Temperature measurement of the reactor materials samples irradiated in the fuel channels of the RA reactor - Annex 16; Prilog 16 - Merenje temperature uzoraka reaktorskih materijala ozracivanih u gorivnim kanalima reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Nikolic, M; Djalovic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    Reactor materials as graphite, stainless steel, magnox, zirconium alloys, etc. were exposed to fast neutron flux inside the fuel elements specially adapted for this purpose. Samples in the form ampoules were placed in capsules inside the fuel channels and cooled by heavy water which cools the fuel elements. In order to monitor the samples temperature 42 thermocouples were placed in the samples. That was necessary for reactor safety reasons and for further interpretation of measured results. Temperature monitoring was done continuously by multichannel milivoltmeters. This paper describes the technique of introducing the thermocouples, compensation instruments, control of the cold ends and adaptation of the instruments for precision (0.5%) temperature measurement in the range 30 deg - 130 deg C; 30 deg - 280 deg C and 30 deg - 80 deg C. Ozracivanje uzoraka materijala za izgradnju reaktora kao sto su grafit, nerdjajuci celik, magnox, legure cirkonijuma, aluminijuma itd. vrseno je u fluksu brzih neutrona unutar samih gorivnih elemenata koji su specijalno adaptirani za ovu svrhu. Uzorci u vidu ampula smesteni su u kapsulu od aluminijuma i postavljeni unutar kanala gde su hladjeni cirkulacijom teske vode koja hladi same gorivne elemente. U cilju kontrole temperature uzoraka radi bezbednosti samog reaktora, kao i radi kasnije interpretacije rezultata ispitivanja radijacionog ostecenja materijala, ugradjeno je 42 termopara u uzorke. Kontrola temperature je vrsena kontinualno visekanalnim registratorima. U radu je prikazana tehnika izvodjenja termoparova, kompenzacionih vodova, kontrola hladnih krajeva i prilagodjenje instrumentacije za merenje i registraciju temperature sa tacnoscu 0,5% u opsezima 30 deg - 130 deg C; 30 deg - 280 deg C i 30 deg - 80 deg C (author)

  13. 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

  14. 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

  15. 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

  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 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.

  18. 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)

  19. 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)

  20. Advanced post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2002-03-01

    The purpose of the meeting was to provide and overview of the status of post-irradiation examination (PIE) techniques for water cooled reactor fuel assemblies and their components with emphasis given to advanced PIE techniques applied to high burnup fuel. Papers presented at the meeting described progress obtained in non-destructive (e.g. dimensional measurements, oxide layer thickness measurements, gamma scanning and tomography, neutron and X-ray radiography, etc.) and destructive PIE techniques (e.g. microstructural studies, elemental and isotopic analysis, measurement of physical and mechanical properties, etc.) used for investigation of water reactor fuel. Recent practice in high burnup fuel investigation revealed the importance of advanced PIE techniques, such as 3-D tomography, secondary ion mass spectrometry, laser flash, high resolution transmission and scanning electron microscopy, image analysis in microstructural studies, for understanding mechanisms of fuel behaviour under irradiation. Importance and needs for in-pile irradiation of samples and rodlets in instrumented rigs were also discussed. This TECDOC contains 20 individual papers presented at the meeting; each of the papers has been indexed separately

  1. Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

    International Nuclear Information System (INIS)

    Edwards, G.W.R.; Priest, N.D.; Richardson, R.B.

    2013-01-01

    The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

  2. Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, G.W.R.; Priest, N.D.; Richardson, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

  3. 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)

  4. Determination of burnup in irradiated nuclear fuels by the method of stable 148Nd fission products

    International Nuclear Information System (INIS)

    Souza Sarkis, J.E. de.

    1982-01-01

    A method is described for the isotopic analysis and determination of burnup in irradiated nuclear fuel by mass spectrometric technique. The burnup is calculed from the determination of the concentration of uranium, plutonium and the fission product 148 Nd in the samples of UO 2 irradiated fuel from a P.W.R. type reactor. The method involves the separation of fractions of uranium, plutonium and neodimium from the dissolved irradiated fuel by ion exchange technique. The determination of uranium, plutonium and the fission product 148 Nd is carried out by isotope diluition mass spectrometry technique using as isotope tracers the nuclides 233 U, 242 Pu and 150 Nd. For the chemical processing and handling of the irradiated sample a Hot Chemistry Laboratory was mounted. It consists of glove boxes and equipments for radiation monitoring and protection. The results obtained indicate an atom percent burnup of 2.181 + - 0.035% wich corresponds an 20.937 + - 0.739 Gwd/ton of thermal energy produced. (Author) [pt

  5. 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

  6. 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

  7. 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.

  8. Irradiation test of HAFM and tag gas samples at the standard neutron field of 'YAYOI'

    International Nuclear Information System (INIS)

    Iguchi, Tetsuo

    1997-03-01

    To check the accuracy of helium accumulation neutron fluence monitors (HAFM) as new technique for fast reactor neutron dosimetry and the applicability of tag gas activation analysis to fast reactor failed fuel detection, their samples were irradiated at the standard neutron field of the fast neutron source reactor 'YAYOI' (Nuclear Engineering Research Laboratory, University of Tokyo). Since October in 1996, the HAFM samples such as 93% enriched boron (B) powders of 1 mg and natural B powders of 10 mg contained in vanadium (V) capsule were intermittently irradiated at the reactor core center (Glory hole: Gy) and/or under the leakage neutron field from the reactor core (Fast column: FC). In addition, new V capsules filled with enriched B of 40 mg and Be of 100 mg, respectively, were put into an experimental hole through the blanket surrounding the core. These neutron fields were monitored by the activation foils consisting of Fe, Co, Ni, Au, 235 U, 237 Np etc., mainly to confirm the results obtained from 1995's preliminary works. In particular, neutron flux distributions in the vicinity of irradiated samples were measured in more detail. At the end of March in 1997, the irradiated neutron fluence have reached the goal necessary to produce the detectable number of He atoms more than ∼10 13 in each HAFM sample. Six kinds of tag gas samples, which are the mixed gases of isotopically adjusted Xe and Kr contained in SUS capsules, were separately irradiated three times at Gy under the neutron fluence of ∼10 16 n/cm 2 in average. After irradiation, γ-ray spectra were measured for each sample. Depending on the composition of tag gas mixtures, the different patterns of γ-ray peak spectra from 79 Kr, 125 Xe, etc. produced through tag gas activation were able to be clearly identified. These experimental data will be very useful for the benchmark test of tag gas activation calculation applied to the fast reactor failed fuel detection. (author)

  9. Change of microflora of two starch samples by gamma irradiation

    International Nuclear Information System (INIS)

    Fretton, R.; Fretton, J.; Delattre, J.M.

    1975-01-01

    Starch is the basic component of a larger number of manufactured foods. The disinfection of such a powder by 60 Co is studied here. Gamma irradiation of two starch samples with different degrees of contamination allows the assumption that, in most cases, good radio-pasteurization can be achieved with 300 krad. The radio-pasteurization doses (varying from 300 to 600 krad) are a function of the initial contamination. Irradiation effects are spectacular with moulds. Activation of spores of some Clostridium species leads us to recommend an irradiation level higher than 200 krad. The most resistant organisms to gamma irradiation are the aerobic and anaerobic sporulated bacteria. The thermophilic forms are the most important. Spores of Bacillus, chiefly Bacillus licheniformis and Bacillus brevis, are the most frequent bacteria. Storage of irradiated starch at room temperature has little effect upon the number of revivable survivors. (orig.) [de

  10. 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)

  11. 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

  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. Preliminary decay heat calculations for the fuel loaded irradiation loop device of the RMB multipurpose Brazilian reactor

    Energy Technology Data Exchange (ETDEWEB)

    Campolina, Daniel; Costa, Antonio Carlos L. da; Andrade, Edison P., E-mail: campolina@cdtn.br, E-mail: aclp@cdtn.br, E-mail: epa@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

    2017-07-01

    The structuring project of the Brazilian Multipurpose Reactor (RMB) is responsible for meeting the capacity to develop and test materials and nuclear fuel for the Brazilian Nuclear Program. An irradiation test device (Loop) capable of performing fuel test for power reactor rods is being conceived for RMB reflector. In this work preliminary neutronic calculations have been carried out in order to determine parameters to the cooling system of the Loop basic design. The heat released as a result of radioactive decay of fuel samples was calculated using ORIGEN-ARP and it resulted less than 200 W after 1 hour of irradiation interruption. (author)

  16. Armor: An αβγ assembly for irradiated fuel analysis

    International Nuclear Information System (INIS)

    Beraud, M.

    1967-04-01

    The assembly ARMOR which was built with a view to carrying out research on irradiated fuels consists of an αβγ enclosure made up of 11 cells in line. After a general description of the assembly in its present form, the various functions are reviewed: introduction of the samples, chemical de-canning, dissolution of the irradiated uranium pellets, preparation of solutions for mass spectrometric analyses, disposal of the effluents and of the solid waste. The assembly-has been working since 1961. During the 5 to 6 years operation, various improvements have been made and a certain number of observations have been collected concerning the work. (author) [fr

  17. 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

  18. 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.

  19. 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.

  20. 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

  1. Water sampling device for detecting fuel failure

    International Nuclear Information System (INIS)

    Masubuchi, Yukio.

    1997-01-01

    A notched portion is formed at the lower end of an outer cap, and an extensible air bag is disposed being in contact with the inner side of the notched portion. A compressed air is sent into the outer gap through an air supply pipe to urge coolants thereby lowering the water level. A portion of the compressed air gets out of the outer gap from the notched portion, and if air bubbles are observed on the surface of coolants in a pressure vessel of a reactor, the outer cap is confirmed to be attached to the upper lattice plate. Compressed air is supplied to the air bag to close the notched portion. Then, coolants are sucked from a water level confirmation pipe. The level of coolants is further lowered, and the compressed air is sucked from the water level confirmation pipe instead of the coolants. Then, the level of the coolants at the inner side of the inner cap is confirmed to be made lower than the upper end of the channel box of a reactor fuel assembly. Then, coolants in the channel box are sampled, as a specimen water, through a water sampling pipe. (I.N.)

  2. Oxidative stability examination of irradiated swine for samples

    International Nuclear Information System (INIS)

    Formanek, Z.

    1995-01-01

    DSC (Differential Scanning Calorimetry), OBM (Oxygen Bomb Method) methods and POV (Peroxide Value) determination was applied to determine the oxidative stability of lard samples irradiated by a Co 60 gamma source at the dose of 0; 0.75; 2.5kGy. Both of the methods seem to be suitable to determine oxidative stability of fats, because in every case lowering oxidative stability was received with increasing irradiation dose. (R.P.). 8 refs., 7 figs., 1 tab

  3. UMo nuclear fuels behaviour under heavy ion irradiation: a μ-XAS study

    International Nuclear Information System (INIS)

    Palancher, H.; Martin, P.; Dubois, S.; Valot, C.; Sabathier, C.; Palancher, H.; Nassif, V.; Proux, O.; Hazemann, J.L.; Wieschalla, N.; Petry, W.; Jarousse, C.

    2007-01-01

    Full text of publication follows. A worldwide program encourages the use of low enriched uranium (LEU, 235 U 235 U concentration up to 93 wt. %). Due to the decrease in 235 U enrichment for the conversion to LEU, the total density of uranium atoms in the fuel must be increased accordingly. To preserve the neutron flux, metallic uranium alloys could be the best fuel material. The fuel, which consists of UMo alloy spherical particles surrounded by an Al matrix (cf. Figure 1-A), is rolled between two aluminium claddings. Post-irradiation examinations of U-7 wt%Mo demonstrated its strong potentialities as fuel but they also pointed out its interaction with aluminium (cf. Figure 1-B). In certain cases this interaction can cause a break-away swelling of the plate. The aim of this project is the understanding of: - the phenomena driving the growth of the interaction layer. - the influence on interaction layer composition of limited adjunction of elements (silicon...) to the Al matrix. To overcome the difficulties inherent to the in-pile irradiated samples, an out-of-pile methodology (collaboration between CEA, FRM II and CERCA) has been developed based on heavy ion irradiation. This methodology enables to simulate the fission fragment damages using a 80 MeV iodine beam at the Maier Leibnitz laboratory (Garching, Germany). After irradiation, samples are characterised at micrometer scale by microscopy (SEM coupled with EDX) and X-Ray techniques (XRD and XAS). The irradiation (final dose: 2 x 10 17 at/cm 2 ) of undoped U-7 wt%Mo fuel plates leads to the formation of an interaction layer surrounding each fuel particles (cf. Figure 1-C). μ-XRD analysis performed at the ESRF (ID18f) showed only the presence of UAl 3 phase in the interaction layer. Same results have been obtained on in-pile irradiated fuel by Sears et al using neutron diffraction confirming the interest of the developed methodology. However the behaviour of the Mo atoms in the interaction layer could not be

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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

  11. Sampling by electro-erosion on irradiated materials

    International Nuclear Information System (INIS)

    Riviere, M.; Pizzanelli, J.P.

    1986-05-01

    Sampling on irradiated materials, in particular for mechanical property study of steels in the FAST NEUTRON program needed the set in a hot cell of a machining device by electroerosion. This device allows sampling of tenacity, traction, resilience test pieces [fr

  12. Scanning electron microscopy analysis of fuel/matrix interaction layers in highly-irradiated U-Mo dispersion fuel plates with Al and Al-Si alloy matrices

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, Dennis D. Jr; Jue, Jan Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adom B.; Medvedev, Pavel; Madden, James; Wachs, Dan; Meyer, Mitch [Nuclear Fuels and Materials Division, Idaho National Laboratory (United States)

    2014-04-15

    In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U-7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifically, samples from irradiated U-7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (-4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U-7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U-7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.

  13. 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)

  14. 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

  15. An estimate of radiation fields in a gamma irradiation facility using fuel elements from a swimming pool reactor

    International Nuclear Information System (INIS)

    Narain, Rajendra

    2002-01-01

    A simple gamma irradiation facility set up using a few irradiated or partially irradiated swimming pool elements can be assembled to provide a convenient facility for irradiation of small and medium sized samples for research. The paper presents results of radiation levels with an arrangement using four elements from a reactor core operating at a power of 20 MW. A maximum gamma field of higher than 1 KGy/h at locations adjacent to fuel elements with negligible neutron contamination can be achieved. (author)

  16. Facilities for post-irradiation examination of experimental fuel elements at Chalk River Nuclear Laboratories

    International Nuclear Information System (INIS)

    Mizzan, E.; Chenier, R.J.

    1979-10-01

    Expansion of post-irradiation facilities at the Chalk River Nuclear Laboratories and steady improvement in hot-cell techniques and equipment are providing more support to Canada's reactor fuel development program. The hot-cell facility primarily used for examination of experimental fuels averages a quarterly throughput of 40 elements and 110 metallographic specimens. New developments in ultrasonic testing, metallographic sample preparation, active storage, active waste filtration, and fissile accountability are coming into use to increase the efficiency and safety of hot-cell operations. (author)

  17. 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.

  18. 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.

  19. 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

  20. 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

  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. 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

  3. Behaviour in air at 175-400 degrees C of irradiated UO2 fuel

    International Nuclear Information System (INIS)

    Hastings, I.J.; McCracken, D.

    1984-09-01

    The authors extended their study of irradiated, defected UO 2 fuel elements to 200 and 400 degrees C. At 200 degrees C there was no diametral change, but at 400 degrees C we observed swelling and severe sheath splitting. Neither short-lived fission products, nor Cs-134, Cs-137 or Ru-106 above background, were detected. Maximum Kr-85 release was 4 Bq ( -6 Ci). Discharge time was 2.5 years. UO 2 fragment studies were extended to 400 degrees C. The oxidation process for unirradiated and irradiated fuel up to 300 degrees C was characterized by activation energies of 140 +- 10 and 120 +- 10 kJ/mol, respectively; enhancement of oxidation rate was confirmed in the irradiated samples. There is an apparent reduction of activation energy above about 300 degrees C. Fuel elements with artificial and natural defects showed similar oxidation and dimensional response at 250 degrees C. Behaviour of fuel fragments from the defect area of a naturally-defected element is consistent with that for fragments from intact elements when prior oxidation during the defect period is considered

  4. 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)

  5. 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)

  6. System of leak inspection of irradiated fuel; Sistema de inspeccion de fuga de combustible irradiado

    Energy Technology Data Exchange (ETDEWEB)

    Delfin L, A.; Castaneda J, G.; Mazon R, R.; Aguilar H, F. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: rmr@nuclear.inin.mx

    2007-07-01

    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

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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.

  12. Inert materials for the GFR fuel. Characterizations, chemical interactions and irradiation damage

    International Nuclear Information System (INIS)

    Audubert, Fabienne; Carlot, Gaoelle; Lechelle, Jacques; David, Laurent; Gomes, Severine

    2005-01-01

    In the framework of an extensive R and D Program on GFR fuel, studies on inert materials have been performed at the French Atomic Energy Commission (CEA). The inert materials would be associated with the fuel with the aim of featuring an efficient barrier to radiotoxic species with regard to the cooling circuit of the reactor. Potential matrices identified for dispersion fuels or particles fuels are SiC, TiN, ZrN, ZrC, TiC. Physical microstructural and thermal properties have been determined in order to evaluate elaboration process effects. The evolution under irradiation of thermal properties (such as conductivity, diffusivity) of the materials has been studied using heavy ions to simulate fission product irradiation. After irradiation, scanning thermal microscopy is used to investigate the thermal degradation of the materials. Thermal conductivity variations were obtained on TiC irradiated with krypton ion at an energy of 86 MeV and a fluence of 5.10 15 ions.cm -2 . They are quantified at 19 W.m -1 .K -1 . On other materials such as SiC, ZrC, TiN, no thermal conductivity contrast was shown. Reactivity between the inert matrix (SiC or TiN) and the fuel (U, Pu)N have been evaluated on powders and on ceramic samples in contact by a thermal treatment under several atmospheres. It was shown that SiC reacts with (U, Pu)N in various atmospheres making secondary phases as PuSi 2 , USi 2 , U 20 Si 16 C 3 . TiN behaviour seems to be better: the only reactivity which may take place would be a variation of the nitrogen stoichiometry in TiN and (U, Pu)N at the interface. (author)

  13. Neutron Focusing Mirrors for Neutron Radiography of Irradiated Nuclear Fuel at Idaho National Laboratory

    Science.gov (United States)

    Rai, Durgesh K.; Wu, Huarui; Abir, Muhammad; Giglio, Jeffrey; Khaykovich, Boris

    Post irradiation examination (PIE) of samples irradiated in nuclear reactors is a challenging but necessary task for the development on novel nuclear power reactors. Idaho National Laboratory (INL) has neutron radiography capabilities, which are especially useful for the PIE of irradiated nuclear fuel. These capabilities are limited due to the extremely high gamma-ray radiation from the irradiated fuel, which precludes the use of standard digital detectors, in turn limiting the ability to do tomography and driving the cost of the measurements. In addition, the small 250 kW Neutron Radiography Reactor (NRAD) provides a relatively weak neutron flux, which leads to low signal-to-noise ratio. In this work, we develop neutron focusing optics suitable for the installation at NRAD. The optics would separate the sample and the detector, potentially allowing for the use of digital radiography detectors, and would provide significant intensity enhancement as well. The optics consist of several coaxial nested Wolter mirrors and is suited for polychromatic thermal neutron radiation. Laboratory Directed Research and Development program of Idaho National Laboratory.

  14. 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)

  15. 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)

  16. 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)

  17. 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.

  18. 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.

  19. 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)

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. TEM characterization of irradiated U-7Mo/Mg dispersion fuel

    Science.gov (United States)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J. F.; Robinson, A. B.; Madden, J.

    2017-10-01

    This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 1021 f/cm3, 7.4 × 1014 f/cm3/s and 123 °C, and 5.5 × 1021 f/cm3, 11.0 × 1014 f/cm3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al3Mg2 and Al12Mg17 along with precipitates of MgO, Mg2Si and FeAl5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.

  5. Hall effect measurements on proton-irradiated ROSE samples

    International Nuclear Information System (INIS)

    Biggeri, U.; Bruzzi, M.; Borchi, E.

    1997-01-01

    Bulk samples obtained from two wafers of a silicon monocrystal material produced by Float-Zone refinement have been analyzed using the four-point probe method. One of the wafers comes from an oxygenated ingot; two sets of pure and oxygenated samples have been irradiated with 24 GeV/c protons in the fluence range from 10 13 p/cm 2 to 2x10 14 p/cm 2 . Van der Pauw resistivity and Hall coefficient have been measured before and after irradiation as a function of the temperature. A thermal treatment (30 minutes at 100C) has been performed to accelerate the reverse annealing effect in the irradiated silicon. The irradiated samples show the same exponential dependence of the resistivity and of the Hall coefficient on the temperature from 370K to 100K, corresponding to the presence of radiation-induced deep energy levels around 0.6-0.7eV in the silicon gap. The free carrier concentrations (n, p) have been evaluated in the investigated fluence range. The inversion of the conductivity type from n to p occurred respectively at 7x10 13 p/cm 2 and at 4x10 13 p/cm 2 before and after the annealing treatment, for both the two sets. Only slight differences have been detected between the pure and oxygenated samples

  6. Database of Samples Irradiated at Reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Muhd Husamuddin Abdul Khalil; Mohd Amin Sharifuldin Salleh; Julia Abdul Karim

    2011-01-01

    Evaluation has been made to data of irradiated samples for the type of sample requested for activation at RTP. Sample types are grouped with percentage of total throughputs to rule out the weight percent of every respective group. The database consists of radionuclide inventory of short and long half-life and high activity radionuclides such as Br and Au have been identified and that database has been constructed using a user-friendly Microsoft Access. Through this, trend of gamma exposure will easily be evaluated at experimental facilities and could ensure radiological effect towards safety and health is limited per Radiation Protection (Basic Safety Standard) Regulation 1988. This database places an important parameter to improve management system in acquiring information of the samples irradiated at RTP and will enhance the safety assurance and reliability of the experimental design basis. (author)

  7. Results of the irradiation of mixed UO2 - PuO2 oxide fuel elements

    International Nuclear Information System (INIS)

    Mikailoff, H.; Mustelier, J.P.; Bloch, J.; Ezran, L.; Hayet, L.

    1966-01-01

    In order to study the behaviour of fuel elements used for the first charge of the reactor Rapsodie, a first batch of eleven needles was irradiated in the reactor EL3 and then examined. These needles (having a shape very similar lo that of the actual needles to be used) were made up of a stack of sintered mixed-oxide pellets: UO 2 containing about 10 per cent of PuO 2 . The density was 85 to 97 per cent of the theoretical, value. The diametral gap between the oxide and the stainless steel can was between 0,06 and 0,27 mm. The specific powers varied from 1230 to 2700 W/cm 3 and the can temperature was between 450 and 630 C. The maximum burn-up attained was 22000 MW days/tonne. Examination of the needles (metrology, radiography and γ-spectrography) revealed certain macroscopic changes, and the evolution of the fuel was shown by micrographic studies. These observations were used, together with flux measurements results, to calculate the temperature distribution inside the fuel. The volume of the fission gas produced was measured in some of the samples; the results are interpreted taking into account the temperature distribution in the oxide and the burn-up attained. Finally a study was made both of the behaviour of a fuel element whose central part was molten during irradiation, and of the effect of sodium which had penetrated into some of the samples following can rupture. (author) [fr

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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)

  14. 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)

  15. 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)

  16. 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.

  17. Effect of alpha self irradiation on helium migration in (U,Pu)O2 samples

    International Nuclear Information System (INIS)

    Pipon, Yves; Roudil, Daniele; Jegou, Christophe; Khodja, Hicham; Raepsaet, Caroline

    2008-01-01

    The helium behavior and its migration mechanisms in nuclear spent fuel (UOX and MOX) significantly impact the possible evolution of the spent fuel matrix in a closed system during interim storage or during a disposal repository. An experimental study has been conducted on (U,Pu)O 2 samples in order to investigate the impact of defects created by alpha decay on helium diffusion. One large part is devoted to thermal atomic diffusion and applied on 3 He implanted samples, annealed at 850 and 1000 deg. C. The He profiles, as implanted and after annealing, were investigated with the 3 He(d,p) 4 He nuclear reaction. Another part deals with the thermal release of 4 He amassed in the samples. The measured thermal diffusion coefficients are compared with previously published values, thereby highlighting the effect of the alpha self-irradiation on helium behavior. (authors)

  18. 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)

  19. Improvement of fuel sampling device for STACY and TRACY

    International Nuclear Information System (INIS)

    Hirose, Hideyuki; Sakuraba, Koichi; Onodera, Seiji

    1998-05-01

    STACY and TRACY, static and transient experiment facilities in NUCEF, use solution fuel. It is important to analyze accurately fuel composition (uranium enrichment, uranium concentration, nitric acid morality, amount of impurities, radioactivity of FP) for their safety operation and improvement of experimental accuracy. Both STACY and TRACY have the sampling devices to sample fuel solution for that purpose. The previous sampling devices of STACY and TRACY had been designed to dilute fuel sample with nitric acid. Its sampling mechanism could pour fuel sample into sampling vessel by a piston drive of nitric acid in the burette. It was, however, sometimes found that sample fuel solution was diluted by mixing with nitric acid in the burette. Therefore, the sampling mechanism was change into a fixed quantity pump drive which didn't use nitric acid. The authors confirmed that the performance of the new sampling device was improved by changing sampling mechanism. It was confirmed through the function test that the uncertainty in uranium concentration measurement using the improved sampling device was 0.14%, and less than the designed value of 0.2% (coefficient of variation). (author)

  20. Measuring a ballooning gap size of irradiated fuels by the indirect method of neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Cheul Muu; Lee, Seung Wook; Lim, In Cheol; Hong, Kwang Pyo; Kim, Young Jin

    2003-11-01

    An indirect method of a neutron radiography is mobilized for inspecting post irradiated nuclear fuel pins, UO{sub 2}Si clad with Al, which swallowing, crack, ballooning, plug gap, thinning and so on, occurred. The system of an indirect method consists of a cask of carrying fuel pins, Dy converter, linear controller of converter, camera of monitoring fuel cassette. The nuclear sample pins of RISO and KAERI are exposed for 25 min. at the first exposure room, 10{sup 7} cm{sup 2}/sec flux. An activation image formed in the Dy foil is subsequently transferred in a dark room for a more than 8 hours to SR film using the decay radiation. Due to L/D ratio an unsharpness of 9.82{approx}14{mu}m and a magnification of 1.0003 are given. After digitizing an image of SR film, the ballooning gap of plug is discernible by H/V filter of image processing.

  1. Measuring a ballooning gap size of irradiated fuels by the indirect method of neutron radiography

    International Nuclear Information System (INIS)

    Sim, Cheul Muu; Lee, Seung Wook; Lim, In Cheol; Hong, Kwang Pyo; Kim, Young Jin

    2003-11-01

    An indirect method of a neutron radiography is mobilized for inspecting post irradiated nuclear fuel pins, UO 2 Si clad with Al, which swallowing, crack, ballooning, plug gap, thinning and so on, occurred. The system of an indirect method consists of a cask of carrying fuel pins, Dy converter, linear controller of converter, camera of monitoring fuel cassette. The nuclear sample pins of RISO and KAERI are exposed for 25 min. at the first exposure room, 10 7 cm 2 /sec flux. An activation image formed in the Dy foil is subsequently transferred in a dark room for a more than 8 hours to SR film using the decay radiation. Due to L/D ratio an unsharpness of 9.82∼14μm and a magnification of 1.0003 are given. After digitizing an image of SR film, the ballooning gap of plug is discernible by H/V filter of image processing

  2. Remotised sliver sample removal from irradiated pressure tube

    Energy Technology Data Exchange (ETDEWEB)

    Rupani, B B; Sharma, B S.V.G.; Shyam, T V; Sinha, R K [Bhabha Atomic Research Centre, Bombay (India). Reactor Engineering Div.

    1994-12-31

    The life of irradiated pressure tubes of Pressurised Heavy Water Reactor (PHWR) can be assessed by analysing hydrogen content in sliver samples of pressure tube material. The sample can be obtained in the form of small slivers by scraping at any specified locations within the bore of the pressure tube of operating reactor. A tool namely Hydride Scraping Tool (HST) has been developed to obtain very thin slivers of hydride bearing samples from irradiated pressure tube. In order to save man-rem consumption during scraping operation, a feeding mechanism has also been designed and developed for axial positioning of scraping tool in the channel remotely. This paper covers general details about constructional features of the scraping tool, feeding mechanism and their control system. It also highlights the operational salient features and capabilities of the system. Possible applications of the feeding mechanism in other fields are also indicated. (author). 4 figs.

  3. Remotised sliver sample removal from irradiated pressure tube

    International Nuclear Information System (INIS)

    Rupani, B.B.; Sharma, B.S.V.G.; Shyam, T.V.; Sinha, R.K.

    1994-01-01

    The life of irradiated pressure tubes of Pressurised Heavy Water Reactor (PHWR) can be assessed by analysing hydrogen content in sliver samples of pressure tube material. The sample can be obtained in the form of small slivers by scraping at any specified locations within the bore of the pressure tube of operating reactor. A tool namely Hydride Scraping Tool (HST) has been developed to obtain very thin slivers of hydride bearing samples from irradiated pressure tube. In order to save man-rem consumption during scraping operation, a feeding mechanism has also been designed and developed for axial positioning of scraping tool in the channel remotely. This paper covers general details about constructional features of the scraping tool, feeding mechanism and their control system. It also highlights the operational salient features and capabilities of the system. Possible applications of the feeding mechanism in other fields are also indicated. (author). 4 figs

  4. 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.

  5. 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)

  6. 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.

  7. 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

  8. 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)

  9. 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)

  10. 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)

  11. Interpretation of measurements made by oscillations of irradiated fuels in natural uranium, graphite-gas piles

    International Nuclear Information System (INIS)

    Laponche, Bernard; Luffin, Jean; Brunet, Max; Guerange, Jacques; Tonolli, Jacky

    1969-06-01

    When considering a pile operation, it is interesting to know the evolution of fuel quality with respect to irradiation, i.e. the variation of its fission rate and of its absorption rate. In order to experimentally obtain these features, a method is to introduce an irradiated cartridge into a critical reactor and to measure the induced effect on its reactivity and on the neutron density at the vicinity of the cartridge. An oscillation method presented in another document and based on a periodic introduction of fuel sample into a critical reactor allows, from the measurement of reactivity variation (global signal), and of the neutron density (local signal), effective macroscopic fission and absorption cross sections of this sample to be obtained. As previous studies revealed that the interpretation of the local signal was notably delicate, this information has been replaced by computed information, the fission rate, which is determined by means of the COREGRAF1 code. Thus, the remaining quantity to be obtained is the fuel absorption rate. The authors report studies performed on several sets of cartridges from different reactors, and with an irradiation range from about 700 to 4000 MWJ/T. In a first part, they describe the characteristics of the studied cartridges, their irradiation and measurement conditions, and the use of the evolution code. In a second part, they try to define the interpretation of oscillation-based measurements by using two methods, a first and fast one which gives an approximation of results, and a more elaborated second one which complies with measurement conditions. The last part presents and discusses the obtained results [fr

  12. Focusing mirrors for enhanced neutron radiography with thermal neutrons and application for irradiated nuclear fuel

    Science.gov (United States)

    Rai, Durgesh K.; Abir, Muhammad; Wu, Huarui; Khaykovich, Boris; Moncton, David E.

    2018-01-01

    Neutron radiography is a powerful method of probing the structure of materials based on attenuation of neutrons. This method is most suitable for materials containing heavy metals, which are not transparent to X-rays, for example irradiated nuclear fuel and other nuclear materials. Neutron radiography is one of the first non-distractive post-irradiated examination methods, which is applied to gain an overview of the integrity of irradiated nuclear fuel and other nuclear materials. However, very powerful gamma radiation emitted by the samples is damaging to the electronics of digital imaging detectors and has so far precluded the use of modern detectors. Here we describe a design of a neutron microscope based on focusing mirrors suitable for thermal neutrons. As in optical microscopes, the sample is separated from the detector, decreasing the effect of gamma radiation. In addition, the application of mirrors would result in a thirty-fold gain in flux and a resolution of better than 40 μm for a field-of-view of about 2.5 cm. Such a thermal neutron microscope can be useful for other applications of neutron radiography, where thermal neutrons are advantageous.

  13. 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

  14. 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)

  15. 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)

  16. 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.)

  17. 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

  18. 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.

  19. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed. - Highlights: •Complementary fission gas release events are reported for U-Mo fuel with and without cladding. •Exothermic reaction between Zr diffusion layer and cladding influences fission gas release. •Mechanisms responsible for fission gas release are similar, but with varying timing and magnitude. •Behavior of samples is similar after 800 °C signaling the onset of superlattice destabilization.

  20. 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

  1. Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing

    International Nuclear Information System (INIS)

    Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

    2003-01-01

    The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer

  2. 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)

  3. 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)

  4. Re-irradiation and limit testing of the fuels PWR type reactors

    International Nuclear Information System (INIS)

    Roche, M.; Molvault, M.

    1978-01-01

    In view of investigating the neutron radiation behavior of PWR fuel pins, the S.P.S. (Services des Piles de Saclay) developed a set of experimental means used at OSIRIS in Saclay Nuclear Research Center. Said devices are shown to be able to meet present problems concerning can failures, power and temperature cycling, remote-control studies. These means can also be used either for statistical studies, they can then receive several samples, or for analytical studies in instrumented devices of large capacity and accelerated irradiation rate [fr

  5. 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)

  6. 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)

  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. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  11. 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.

  12. 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.)

  13. 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)

  14. 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)

  15. A concise design o the irradiation of U-10Zr metallic fuel at a very low burnup

    International Nuclear Information System (INIS)

    Guo, Hai Bing; Zhou, Wei; Sun, Yong; Qian, Dazhi; Ma, Jimin; Leng, Jun; Huo, Hyoung; Wang, Shaohua

    2017-01-01

    In order to investigate the swelling behavior and fuel–cladding interaction mechanism of U–10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel–cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal–hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design

  16. A concise design o the irradiation of U-10Zr metallic fuel at a very low burnup

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Hai Bing; Zhou, Wei; Sun, Yong; Qian, Dazhi; Ma, Jimin; Leng, Jun; Huo, Hyoung; Wang, Shaohua [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang (China)

    2017-06-15

    In order to investigate the swelling behavior and fuel–cladding interaction mechanism of U–10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel–cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal–hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

  17. Swelling behavior detection of irradiated U-10Zr alloy fuel using indirect neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Yong; Huo, He-yong; Wu, Yang [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China); Li, Jiangbo [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Zhou, Wei; Guo, Hai-bing [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China); Li, Hang, E-mail: lihang32@gmail.com [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China); Cao, Chao; Yin, Wei; Wang, Sheng; Liu, Bin; Feng, Qi-jie; Tang, Bin [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China)

    2016-11-21

    It is hopeful that fusion-fission hybrid energy system will become an effective approach to achieve long-term sustainable development of fission energy. U-10Zr alloy (which means the mass ratio of Zr is 10%) fuel is the key material of subcritical blanket for fusion-fission hybrid energy system which the irradiation performance need to be considered. Indirect neutron radiography is used to detect the irradiated U-10Zr alloy because of the high residual dose in this paper. Different burnup samples (0.1%, 0.3%, 0.5% and 0.7%) have been tested with a special indirect neutron radiography device at CMRR (China Mianyang Research Reactor). The resolution of the device is better than 50 µm and the quantitative analysis of swelling behaviors was carried out. The results show that the swelling behaviors relate well to burnup character which can be detected accurately by indirect neutron radiography.

  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. Neutron-Irradiated Samples as Test Materials for MPEX

    International Nuclear Information System (INIS)

    Ellis, Ronald James; Rapp, Juergen

    2015-01-01

    Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility

  20. Defect sizing of post-irradiated nuclear fuels using grayscale thresholding in their radiographic images

    International Nuclear Information System (INIS)

    Chaudhary, Usman Khurshid; Iqbal, Masood; Ahmad, Munir

    2010-01-01

    Quantification of different types of material defects in a number of reference standard post-irradiated nuclear fuel image samples have been carried out by virtue of developing a computer program that takes radiographic images of the fuel as input. The program is based on user adjustable grayscale thresholding in the regime of image segmentation whereby it selects and counts the pixels having graylevel values less than or equal to the computed threshold. It can size the defects due to chipping in nuclear fuel, cracks, voids, melting, deformation, inclusion of foreign materials, heavy isotope accumulation, non-uniformity, etc. The classes of fuel range from those of research and power reactors to fast breeders and from pellets to annular and vibro-compacted fuel. The program has been validated against ground truth realities of some locally fabricated metallic plates having drilled holes of known sizes simulated as defects in them in which the results indicate that it either correctly selects and quantifies at least 94% of the actual required regions of interest in a given image or it gives less than 8.1% false alarm rate. Also, the developed program is independent of image size.

  1. Defect sizing of post-irradiated nuclear fuels using grayscale thresholding in their radiographic images

    Energy Technology Data Exchange (ETDEWEB)

    Chaudhary, Usman Khurshid, E-mail: ukhurshid@hotmail.co [Department of Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan); Iqbal, Masood, E-mail: masiqbal@hotmail.co [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad 45650 (Pakistan); Ahmad, Munir [Nondestructive Testing Group, Directorate of Technology, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad 45650 (Pakistan)

    2010-10-15

    Quantification of different types of material defects in a number of reference standard post-irradiated nuclear fuel image samples have been carried out by virtue of developing a computer program that takes radiographic images of the fuel as input. The program is based on user adjustable grayscale thresholding in the regime of image segmentation whereby it selects and counts the pixels having graylevel values less than or equal to the computed threshold. It can size the defects due to chipping in nuclear fuel, cracks, voids, melting, deformation, inclusion of foreign materials, heavy isotope accumulation, non-uniformity, etc. The classes of fuel range from those of research and power reactors to fast breeders and from pellets to annular and vibro-compacted fuel. The program has been validated against ground truth realities of some locally fabricated metallic plates having drilled holes of known sizes simulated as defects in them in which the results indicate that it either correctly selects and quantifies at least 94% of the actual required regions of interest in a given image or it gives less than 8.1% false alarm rate. Also, the developed program is independent of image size.

  2. 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,

  3. 75 FR 62695 - Physical Protection of Irradiated Reactor Fuel in Transit

    Science.gov (United States)

    2010-10-13

    ... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule. SUMMARY: The... nuclear fuel in transit? H. Why require a telemetric position monitoring system or an alternative tracking... nuclear fuel in transit. The interim final rule added 10 CFR 73.37, ``Requirements for Physical Protection...

  4. Irradiation of pressurized water reactor fuel rods in the Forschungsreaktor Juelich 2

    International Nuclear Information System (INIS)

    Gaertner, M.

    1978-10-01

    Test fuel rods have been irradiated in FRJ-2 to study the interaction between fuel and cladding as well as hydride orientation stability in the prehydrided cladding. The fuel rods achieved burn-ups of 3.500 to 10.000 MWd/tU at surface temperatures of 333 0 C and power levels up to 620 W/cm. (orig.) [de

  5. Characterization of the Microstructure of Irradiated U-Mo Dispersion Fuel with a Matrix that Contains Si

    International Nuclear Information System (INIS)

    Keiser, Jr. D.D.; Robinson, A.B.; Jue, J.F.; Medvedev, P.; Finlay, M.R.

    2009-01-01

    RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. Of particular interest is the irradiation performance of U-Mo dispersion fuels with Si added to the Al matrix. Si is added to improve the performance of U-Mo dispersion fuels. Microstructural examinations have been performed on fuel plates with Al-2Si matrix after irradiation to around 50% LEU burnup. Si-rich layers were observed in many areas around the various U-7Mo fuel particles. In one local area of one of the samples, where the Si-rich layer had developed into a layer devoid of Si, relatively large fission gas bubbles were observed in the interaction phase. There may be a connection between the growth of these bubbles and the amount of Si present in the interaction layer. Overall, it was found that having Si-rich layers around the fuel particles after fuel plate fabrication positively impacted the overall performance of the fuel plate

  6. Stability of zinc stearate under alpha irradiation in the manufacturing process of SFR nuclear fuels

    Science.gov (United States)

    Gracia, J.; Vermeulen, J.; Baux, D.; Sauvage, T.; Venault, L.; Audubert, F.; Colin, X.

    2018-03-01

    The manufacture of new fuels for sodium-cooled fast reactors (SFRs) will involve powders derived from recycling existing fuels in order to keep on producing electricity while saving natural resources and reducing the amount of waste produced by spent MOX fuels. Using recycled plutonium in this way will significantly increase the amount of 238Pu, a high energy alpha emitter, in the powders. The process of shaping powders by pressing requires the use of a solid lubricant, zinc stearate, to produce pellets with no defects compliant with the standards. The purpose of this study is to determine the impact of alpha radiolysis on this additive and its lubrication properties. Experiments were conducted on samples in contact with PuO2, as well as under external helium ion beam irradiation, in order to define the kinetics of radiolytic gas generation. The yield results relating to the formation of these gases (G0) show that the alpha radiation of plutonium can be simulated using external helium ion beam irradiation. The isotopic composition of plutonium has little impact on the yield. However, an increased yield was globally observed with increasing the mean linear energy transfer (LET). A radiolytic degradation process is proposed.

  7. 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)

  8. Research of pneumatic control transmission system for small irradiation samples

    International Nuclear Information System (INIS)

    Bai Zhongxiong; Zhang Haibing; Rong Ru; Zhang Tao

    2008-01-01

    In order to reduce the absorbed dose damage for the operator, pneumatic control has been adopted to realize the rapid transmission of small irradiation samples. On/off of pneumatic circuit and directions for the rapid transmission system are controlled by the electrical control part. The main program initializes the system and detects the location of the manual/automatic change-over switch, and call for the corresponding subprogram to achieve the automatic or manual operation. Automatic subprogram achieves the automatic sample transmission; Manual subprogram completes the deflation, and back and forth movement of the radiation samples. This paper introduces in detail the implementation of the system, in terms of both hardware and software design. (authors)

  9. 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)

  10. Experimental Irradiations of Materials and Fuels in the BR2 Reactor: An Overview of Current Programmes

    International Nuclear Information System (INIS)

    Van Dyck, S.; Koonen, E.; Verwerft, M.; Wéber, M.

    2013-01-01

    The BR2 material test reactor offers a variety of experimental irradiation possibilities for testing of materials, fuels and instruments. The current paper gives an overview of the recent and ongoing programmes in order to illustrate the experimental potential of the reactor. Three domains of applications are reviewed: Irradiation of materials and fuels for pressurised water reactors (PWR); irradiation of materials for accelerator driven systems (ADS), cooled by liquid lead alloys; and irradiation of fuel for Material Test Reactors (MTR). For PWR relevant tests, a dedicated loop is available, providing a full simulation of the thermo hydraulic conditions of a PWR. ADS related tests require particular control of the irradiation environment and the necessary safety precautions in order to avoid 210 Po contamination. In-core mechanical testing of materials is done in comparison and complimentarily to post-irradiation examinations in order to assess flux related effects on the deformation behaviour of materials. (author)

  11. Global Threat Reduction Initiative Fuel Thermo-Physical Characterization Project: Sample Management Plan

    Energy Technology Data Exchange (ETDEWEB)

    Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pereira, Mario M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Steen, Franciska H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-01-01

    This sample management plan provides guidelines for sectioning, preparation, acceptance criteria, analytical path, and end-of-life disposal for the fuel element segments utilized in the Global Threat Reduction Initiative (GTRI), Fuel Thermo-Physical Characterization Project. The Fuel Thermo-Physical Characterization Project is tasked with analysis of irradiated Low Enriched Uranium (LEU) Molybdenum (U-Mo) fuel element samples to support the GTRI conversion program. Sample analysis may include optical microscopy (OM), scanning electron microscopy (SEM) fuel-surface interface analysis, gas pycnometry (density) measurements, laser flash analysis (LFA), differential scanning calorimetry (DSC), thermogravimetry and differential thermal analysis with mass spectroscopy (TG /DTA-MS), Inductively Coupled Plasma Spectrophotometry (ICP), alpha spectroscopy, and Thermal Ionization Mass Spectroscopy (TIMS). The project will utilize existing Radiochemical Processing Laboratory (RPL) operating, technical, and administrative procedures for sample receipt, processing, and analyses. Test instructions (TIs), which are documents used to provide specific details regarding the implementation of an existing RPL approved technical or operational procedure, will also be used to communicate to staff project specific parameters requested by the Principal Investigator (PI). TIs will be developed, reviewed, and issued in accordance with the latest revision of the RPL-PLN-700, RPL Operations Plan. Additionally, the PI must approve all project test instructions and red-line changes to test instructions.

  12. Irradiation tasks within development of fuel elements in Sweden; Rad na ozracivanju u okviru razvoja gorivnih elemenata u Svedskoj

    Energy Technology Data Exchange (ETDEWEB)

    Stevanovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-11-15

    This report contains description of the hot laboratory RMA for irradiation in the R-2 reactor in Studsvik. Activities of the AB Atomenegiyu concerning irradiation and testing of fuel rods and fuel elements are described, as well as methods for testing of irradiated samples in hot cells. Concerning the importance of the problem, determination of burnup level and neutron flux were examined particularly. Dat je opis vruce metalurske laboratorije RMA i sistema za ozracivanje u reaktoru R-2 u Studsviku. Prikazana je aktivnost AB Atomenergy u okviru razvoja gorivnih elemenata na ozracivanju i ispitivanju ozracenog goriva, gorivnih sipki i sklopova gorivnih elemenata. Prikazane su metode ispitivanja ozracenih uzoraka u vrucim celijama. S obzirom na vaznost problema, posebno je razradjeno pitanje odredjivanja stepena izgaranja i fluksa neutrona (author)

  13. Characteristics of fuel CRUD from Ringhals Unit 4. A comparison of CRUD samples from ultrasonic fuel cleaning and fuel scrape

    International Nuclear Information System (INIS)

    Chen, Jiaxin; Eskhult, Jonas; Marks, Chuck; Dingee, John; Bengtsson, Bernt; Wells, Daniel

    2014-01-01

    The characteristics and behaviour of PWR fuel CRUD are closely related to plant radiation field build-up and the risks of CRUD-Induced Power Shifts (CIPS, previously AOA) and CRUD-Induced Localized Corrosion (CILC). At Ringhals NPP fuel scrape technology has frequently been used to collect fuel CRUD samples as a part of plant water chemistry monitoring programs. In 2012, high efficiency ultrasonic fuel cleaning (HE-UFC) technology was first used at Ringhals Unit 4 to reduce the risk of CIPS, which was expected to increase due to steam generator replacement and a subsequent power uprate. In the HE-UFC system a “first-of-a-kind” CRUD sampling system was installed for collecting fuel CRUD. It is of interest to compare the fuel CRUD samples collected by the two different CRUD sampling methods and to understand if HE-UFC crud sampling could be used to replace or supplement the fuel scrape method. This paper presents some preliminary results on isotopic compositions, elemental compositions, and phase compositions of fuel CRUD samples collected from similar fuel rods and assemblies by both methods during the 2012 refueling outage, one cycle after steam generator replacement at Ringhals Unit 4. The results show that the characteristics of fuel CRUD sampled by HE-UFC and fuel scrape, although not always identical, were similar or correlated to some extent in terms of weight ratios of Ni to Fe and Cr to Fe as well as specific activities of Co-58, Co-60 and Cr-51. However, due to the limited experience with the HE-UFC sampling method, more consideration is required if the statistical significance of the obtained data and indications are to be verified. (author)

  14. Welding of metallic fuel elements for the irradiation test in JOYO. Preliminary tests and welding execution tests (Joint research)

    International Nuclear Information System (INIS)

    Kikuchi, Hironobu; Nakamura, Kinya; Iwai, Takashi; Arai, Yasuo

    2009-10-01

    Irradiation tests of metallic fuels elements in fast test reactor JOYO are planned under the joint research of Japan Atomic Energy Agency (JAEA) and Central Research Institute of Electric Power Industry (CRIEPI). Six U-Pu-Zr fuel elements clad with ferritic martensitic steel are fabricated in Plutonium Fuel Research Facility (PFRF) of JAEA-Oarai for the first time in Japan. In PFRF, the procedures of fabrication of the fuel elements were determined and the test runs of the equipments were carried out before the welding execution tests for the fuel elements. Test samples for confirming the welding condition between the cladding tube and top and bottom endplugs were prepared, and various test runs were carried out before the welding execution tests. As a result, the welding conditions were finalized by passing the welding execution tests. (author)

  15. 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)

  16. 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)

  17. On transient irradiation behavior of HTGR fuel particles

    International Nuclear Information System (INIS)

    Mortenson, S.C.; Okrent, D.

    1977-01-01

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

  18. 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.)

  19. Irradiation behavior of U{sub 6}Mn-Al dispersion fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, M.K. E-mail: mitchell.meyer@anl.gov; Wiencek, T.C.; Hayes, S.L.; Hofman, G.L

    2000-04-01

    Irradiation testing of U{sub 6}Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U{sub 6}Mn in an unrestrained plate configuration performs similarly to U{sub 6}Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3x10{sup 27} m{sup -3}. Fuel plate failure occurs by fission gas pressure driven 'pillowing' on continued irradiation.

  20. Determination of fission products in irradiated fuel by X-ray fluorescence

    International Nuclear Information System (INIS)

    Mogensen, M.; Als-Nielsen, J.; Hessel Andersen, N.

    1986-08-01

    X-ray fluorescence is a well established analytical tool for measuring elemental composition of fairly large (approximately 5 cm 2 ) ''cold'' samples. A version of this technique has been developed for analysis of radial distribution of fission products Xe, Cs and Ba in irradiated UO 2 fuel samples. About 0.1 mm thin slices of fuel pellets (full cross sections) are irradiated by 50 keV X-rays. The intensity of the Xe (Cs, Ba) K α fluorescence radiation generated is measured by means of a Ge detector fitted with a collimator. The slit is 0.5 mm wide in the scanning direction and 2 mm long. The measured Xe K α X-ray intensities are converted to absolute concentrations by comparing to the intensity from a Xe gas standard. In the case of Cs and Ba solid standards may be used. The X-ray fluorescence analysis is compared to other techniques used to obtain radial fission product profiles. It is shown how a combination of X-ray fluorescence and electron probe micro analysis is able to reveal the amount of Xe in the grain boundary porosities. (author)

  1. System design description for sampling fuel in K basins

    International Nuclear Information System (INIS)

    Baker, R.B.

    1996-01-01

    This System Design Description provides: (1) statements of the Spent Nuclear Fuel Projects (SNFP) needs requiring sampling of fuel in the K East and K West Basins, (2) the sampling equipment functions and requirements, (3) a general work plan and the design logic being followed to develop the equipment, and (4) a summary description of the design for the sampling equipment. The report summarizes the integrated application of both the subject equipment and the canister sludge sampler in near-term characterization campaigns at K Basins

  2. Gamma scanning of mixed carbide and oxide fuel pins irradiated in FBTR

    International Nuclear Information System (INIS)

    Jayaraj, V.V.; Padalakshmi, M.; Ulaganathan, T.; Venkiteswaran, C.N.; Divakar, R.; Joseph, Jojo; Bhaduri, A.K.

    2016-01-01

    Fission in nuclear fuels results in a number of fission products that are gamma emitters in the energy range of 100 keV to 3 MeV. The gamma emitting fission products are therefore amenable for detection by gamma detectors. Assessment of the fission product distribution and their migration behavior through gamma scanning is important for characterizing the in reactor behavior of the fuel. Gamma scanning is an important non destructive technique used to evaluate the behavior of irradiated fuels. As a part of Post Irradiation Examinations (PIE), axial gamma scanning has been carried out on selected fuel pins of the FBTR Mark I mixed carbide fuel sub-assemblies and PFBR MOX test fuel sub-assembly irradiated in FBTR. This paper covers the results of gamma scanning and correlation of gamma scanning results with other PIE techniques

  3. Irradiated Fuel Management Advisory Programme (IFMAP). An interregional technical co-operation project

    International Nuclear Information System (INIS)

    1993-04-01

    Delays in the implementation of the fuel reprocessing option in some countries, the complete abandonment of this option in other countries, and delays in the availability of final spent fuel disposal in almost all countries, has led to increasingly long periods of 'interim' spent fuel storage. The problem of the management of irradiated fuels has therefore increased in importance for many Member States. For power reactors or research reactors, irradiated fuel management includes all of the procedures involving irradiated fuel from the time that it is discharged from the reactor core until it is either reprocessed or placed in a permanent disposal site. Although the IAEA has had programmes in this area in the past and has ongoing activities at present, there is a clear need to provide support to individual Member States which require advice and/or assistance in the resolution of particular questions and concerns

  4. Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

    International Nuclear Information System (INIS)

    Tachibana, Toshimichi

    1985-01-01

    The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

  5. 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

  6. 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.)

  7. Discrimination of irradiated MOX fuel from UOX fuel by multivariate statistical analysis of simulated activities of gamma-emitting isotopes

    Science.gov (United States)

    Åberg Lindell, M.; Andersson, P.; Grape, S.; Hellesen, C.; Håkansson, A.; Thulin, M.

    2018-03-01

    This paper investigates how concentrations of certain fission products and their related gamma-ray emissions can be used to discriminate between uranium oxide (UOX) and mixed oxide (MOX) type fuel. Discrimination of irradiated MOX fuel from irradiated UOX fuel is important in nuclear facilities and for transport of nuclear fuel, for purposes of both criticality safety and nuclear safeguards. Although facility operators keep records on the identity and properties of each fuel, tools for nuclear safeguards inspectors that enable independent verification of the fuel are critical in the recovery of continuity of knowledge, should it be lost. A discrimination methodology for classification of UOX and MOX fuel, based on passive gamma-ray spectroscopy data and multivariate analysis methods, is presented. Nuclear fuels and their gamma-ray emissions were simulated in the Monte Carlo code Serpent, and the resulting data was used as input to train seven different multivariate classification techniques. The trained classifiers were subsequently implemented and evaluated with respect to their capabilities to correctly predict the classes of unknown fuel items. The best results concerning successful discrimination of UOX and MOX-fuel were acquired when using non-linear classification techniques, such as the k nearest neighbors method and the Gaussian kernel support vector machine. For fuel with cooling times up to 20 years, when it is considered that gamma-rays from the isotope 134Cs can still be efficiently measured, success rates of 100% were obtained. A sensitivity analysis indicated that these methods were also robust.

  8. 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.)

  9. 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

  10. Dissolution of nuclear fuel samples for analytical purposes. I

    International Nuclear Information System (INIS)

    Krtil, J.

    1983-01-01

    Main attention is devoted to procedures for dissolving fuels based on uranium metal and its alloys, uranium oxides and carbides, plutonium metal, plutonium dioxide, plutonium carbides, mixed PuC-UC carbides and mixed oxides (PuU)O 2 . Data from the literature and experience gained with the dissolution of nuclear fuel samples at the Central Control Laboratory of the Nuclear Research Institute at Rez are given. (B.S.)

  11. 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)

  12. Characterization of LWR fuel rod irradiations with power transients in the BR2 reflector

    International Nuclear Information System (INIS)

    Ponsard, B.; Bodart, S.; Meer, K. van der; Raedt, C. de

    1996-01-01

    Fuel rod irradiations in reflector positions of the materials testing reactor BR2 are becoming increasingly important. A typical example is that of irradiation devices containing single LWR fuel rods, to be tested in the framework of a new international fuel investigation and development programme. Some of the irradiations will comprise power transients with central fuel melting (at 2800 deg. C), the power increase being obtained by decreasing the pressure in a He-3 neutron absorbing screen and/or by varying the BR2 reactor operating power. A total power variation by a factor of at least 2.5 in the fuel rod irradiated could thus be achieved. In some of the rods, central temperature measurements (up to 2000 deg. C) will be carried out. Both fresh and pre-irradiated fuel rods are concerned in the programme. For these irradiations, the accurate knowledge of the neutron-induced fission heating and of the gamma heating is required, as one of the purposes of the programme consists in establishing the correlation among the thermal conductivity, the burn-up and the irradiation temperature. Calibration work among various measuring methods and between measurements and one- and two-dimensional calculations is being pursued. (author). 10 refs, 15 figs, 3 tabs

  13. Analysis of gamma irradiator dose rate using spent fuel elements with parallel configuration

    International Nuclear Information System (INIS)

    Setiyanto; Pudjijanto MS; Ardani

    2006-01-01

    To enhance the utilization of the RSG-GAS reactor spent fuel, the gamma irradiator using spent fuel elements as a gamma source is a suitable choice. This irradiator can be used for food sterilization and preservation. The first step before realization, it is necessary to determine the gamma dose rate theoretically. The assessment was realized for parallel configuration fuel elements with the irradiation space can be placed between fuel element series. This analysis of parallel model was choice to compare with the circle model and as long as possible to get more space for irradiation and to do manipulation of irradiation target. Dose rate calculation were done with MCNP, while the estimation of gamma activities of fuel element was realized by OREGEN code with 1 year of average delay time. The calculation result show that the gamma dose rate of parallel model decreased up to 50% relatively compared with the circle model, but the value still enough for sterilization and preservation. Especially for food preservation, this parallel model give more flexible, while the gamma dose rate can be adjusted to the irradiation needed. The conclusion of this assessment showed that the utilization of reactor spent fuels for gamma irradiator with parallel model give more advantage the circle model. (author)

  14. 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.

  15. Main results of post-irradiation examinations of new-generation fuel assemblies VVER-1000

    International Nuclear Information System (INIS)

    Zvir, E.; Markov, D.; Polenok, V.; Zhitelev, V.; Kobylyansky, G.

    2009-01-01

    To increase the competitiveness of Russian nuclear fuel at the foreign market and to improve its technical and economic performance in order to provide a necessary level of safety, it is necessary to solve certain important tasks: Increase of fuel burn-up; Extension of operational lifetime of fuel assemblies and operational reliability of nuclear fuel; Introduction of cost-beneficial and flexible fuel cycles. Alternative fuel assemblies TVSA VVER-1000 and TVS-2 are used as a basis to optimize the nuclear fuel and develop advanced fuel cycles for nuclear power plants with VVER-1000 reactor types. Four fuel assemblies TVSA operated during 1 and up to 6 reactor cycles, reference fuel assembly TVS-2 operated during three reactor cycles and achieved an average fuel burnup of 48MW·day/kgU as well as failed fuel assembly TVS-2 operated during one cycle were examined at RIAR in recent years. The main objectives of these examinations were to obtain experimental data in support of operational integrity of products or to find out reasons of their failure. The performed post-irradiation examinations confirmed the operational integrity of alternative fuel assemblies TVSA including their geometrical stability up to the average fuel burnup of 55 MW·day/kgU over the fuel assembly (FA) (up to the maximal fuel burnup of ∼73 MW·day/kgU in fuel rods) and of TVS-2 up to the average fuel burnup of 48 MW·day/kgU over the fuel assembly. The changes introduced in the design of VVER-1000 fuel assembly during the development of alternative fuel assembly TVSA and TVS-2 did not make any negative effect on fuel rods. It was proved that causes of fuel rod failure were not related to design features of fuel assemblies. The design features and operating conditions of fuel assemblies under examinations are briefly described. Post-irradiation examinations proved the geometrical stability of fuel assemblies TVSA and TVS-2 under operation up to the fuel burnup of ∼50 MW day/kgU, as for the

  16. 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)

  17. 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.)

  18. A shielded measurement system for irradiated nuclear fuel measurements

    International Nuclear Information System (INIS)

    Mosby, W.R.; Aumeier, S.E.; Klann, R.T.

    1999-01-01

    The US Department of Energy (DOE) is driving a transition toward dry storage of irradiated nuclear fuel (INF), toward characterization of INF for final disposition, and toward resumption of measurement-based material control and accountability (MC and A) efforts for INF. For these reasons, the ability to efficiently acquire radiological measurements of INF in a dry environment is important. The DOE has recently developed a guidance document proposing MC and A requirements for INF. The intent of this document is to encourage the direct measurement of INF on inventory within DOE. The guidance document reinforces and clarifies existing material safeguards requirements as they pertain to INF. Validation of nuclear material contents of non-self-protecting INF must be accomplished by direct measurement, application of validated burnup codes using qualified initial fissile content, burnup data, and age or by other valid means. The fuel units must remain intact with readable identification numbers. INF may be subject to periodic inventories with visual item accountability checks. Quantitative measurements may provide greater assurance of the integrity of INF inventories at a lower cost and with less personnel exposure than visual item accountability checks. Currently, several different approaches are used to measure the radiological attributes of INF. Although these systems are useful for a wide variety of applications, there is currently no relatively inexpensive measurement system that is readily deployable for INF measurements for materials located in dry storage. The authors present the conceptual design of a shielded measurement system (SMS) that could be used for this purpose. The SMS consists of a shielded enclosure designed to house a collection of measurement systems to allow measurements on spent fuel outside of a hot cell. The phase 1 SMS will contain 3 He detectors and ionization chambers to allow for gross neutron and gamma-ray measurements. The phase 2 SMS

  19. Integrity, behavior and proposal of CARA fuel irradiation with empty negative coefficient

    International Nuclear Information System (INIS)

    Marino, Armando C.; Brasnarof, Daniel O.; Demarco, Gustavo L.; Agueda, Horacio C.

    2007-01-01

    The main issues of the CARA fuel, CVN version, are its negative void reactivity coefficient and an extraction burnup of ∼20000 MWd/ton U. The analysis of the fuel rod behaviour, under the irradiation conditions of the Embalse, Atucha I and II NPPs, are the key to recognize their demanding under operation, to review the classic issues of the PHWR fuels and to prepare a programme of experimental irradiations in order to demonstrate the CARA concept, to assess the fuel integrity, to improve the performance and the enhancement of the safety margins. (author) [es

  20. Microgamma Scan System for analyzing radial isotopic profiles of irradiated transmutation fuels

    International Nuclear Information System (INIS)

    Hilton, Bruce A.; McGrath, Christopher A.

    2008-01-01

    The U. S. Global Nuclear Energy Partnership / Advanced Fuel Cycle Initiative (GNEP/AFCI) is developing metallic transmutation alloys as a fuel form to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. A micro-gamma scan system is being developed to analyze the radial distribution of fission products, such as Cs-137, Cs-134, Ru-106, and Zr-95, in irradiated fuel cross-sections. The micro-gamma scan system consists of a precision linear stage with integrated sample holder and a tungsten alloy collimator, which interfaces with the Idaho National Laboratory (INL) Analytical Laboratory Hot Cell (ALHC) Gamma Scan System high purity germanium detector, multichannel analyzer, and removable collimators. A simplified model of the micro-gamma scan system was developed in MCNP (Monte-Carlo N-Particle Transport Code) and used to investigate the system performance and to interpret data from the scoping studies. Preliminary measurements of the micro-gamma scan system are discussed. (authors)

  1. Formation of actinides in irradiated HTGR fuel elements

    International Nuclear Information System (INIS)

    Santos, A.M. dos.

    1976-03-01

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

  2. Conceptual design for irradiation device used to irradiate experimental LFR fuel element in TRIGA reactor, ACPR zone

    International Nuclear Information System (INIS)

    Ioan, M.

    2013-01-01

    The paper presents the main steps followed to conceive a small, versatile and rather cheep irradiation device used for irradiation of an experimental fuel element, specific for Lead cooled Fast Reactor (LFR), adapted to TRIGA reactor, ACPR zone. This device must be instrumented with at least 4 thermocouples and a pressure transducer. The fuel element (150 mm fuel pellets column) will be immersed in maximum 0.350 kg pure hot lead (400 deg C). The system has three protection barriers, as follows: first is the fuel tube, second is the lead container (maximum 20 mm inner diameter) and third is the external container (maximum 180 mm outside diameter). Before the reactor pulse, the temperature of the lead is set at the prescribed value using an electrical heater (300 W), coil on the second barrier. Outside the second barrier a very good thermal insulation is provided. (authors)

  3. 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

  4. 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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  6. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Sorescu, Ion [Institute for Nuclear Research, Pitesti (Romania). TRIGA Reactor Loop Facility; Parvan, Marcel [Institute for Nuclear Research, Pitesti (Romania). Hot Cells Lab.

    2012-12-15

    Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO{sub 2} grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)

  7. Status of development and irradiation performance of advanced proliferation resistant MTR fuel at NUKEM

    International Nuclear Information System (INIS)

    Hrovat, M.; Hassel, H.-W.; Wehner, E.

    1985-01-01

    This paper describes the current status of development and irradiation performance of fuel elements for Material Test and Research (MTR) Reactors with Medium Enriched Uranium (MEU, ≤ 45 % 235-U) and Low Enriched Uranium (LEU, ≤ 20 % 235-U). (author)

  8. Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Isabella J van Rooyen

    2012-09-01

    The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

  9. Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

    International Nuclear Information System (INIS)

    Reitsamer, G.; Proksch, E.; Stolba, G.; Strigl, A.; Falta, G.; Zeger, J.

    1985-01-01

    Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e. fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations

  10. Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

    Energy Technology Data Exchange (ETDEWEB)

    Reitsamer, G; Proksch, E; Stolba, G; Strigl, A; Falta, G; Zeger, J [Department of Chemistry, Austrian Research Centre Seibersdorf Ltd., Seibersdorf (Austria)

    1985-07-01

    Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e., fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations.

  11. Measuring method for effective neutron multiplication factor upon containing irradiated fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto; Mitsuhashi, Ishi; Sasaki, Tomoharu.

    1993-01-01

    A portion of irradiated fuel assemblies at a place where a reactivity effect is high, that is, at a place where neutron importance is high is replaced with standard fuel assemblies having a known composition to measure neutron fluxes at each of the places. An effective composition at the periphery of the standard fuel assemblies is determined by utilizing a calibration curve determined separately based on the composition and neutron flux values of the standard assemblies. By using the calibration curve determined separately based on this composition and the known composition of the standard fuel assemblies, an effective neutron multiplication factor for the fuel containing portion containing the irradiated fuel assemblies is recognized. Then, subcriticality is ensured and critical safety upon containing the fuel assemblies can be secured quantitatively. (N.H.)

  12. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

  13. Post-irradiation examination of a failed PHWR fuel bundle of KAPS-2

    International Nuclear Information System (INIS)

    Mishra, Prerna; Unnikrishnan, K.; Viswanathan, U.K.; Shriwastaw, R.S.; Singh, J.L.; Ouseph, P.M.; Alur, V.D.; Singh, H.N.; Anantharaman, S.; Sah, D.N.

    2006-08-01

    Detailed post irradiation examination was carried out on a PHWR fuel bundle irradiated at Kakrapar Atomic Power Station unit 2 (KAPS-2). The fuel bundle had failed early in life at a low burnup of 387 MWd/T. Non destructive and destructive examination was carried out to identify the cause of fuel failure. Visual examination and leak testing indicated failure in two fuel pins of the outer ring of the bundle in the form of axial cracks near the end plug location. Ultrasonic testing of the end cap weld indicated presence of lack of fusion type defect in the two fuel pins. No defect was found in other fuel pins of the bundle. Metallographic examination of fuel sections taken from the crack location in the failed fuel pin showed extensive restructuring of fuel. The centre temperature of the fuel had exceeded 1700 degC at this location in the failed fuel pin, whereas fuel centre temperature in the un-failed fuel pin was only about 1300 degC. Severe fuel clad interaction was observed in the failed fuel pin at and near the location of failure but no such interaction was observed in the un-failed fuel pins. Several incipient cracks originating from the inside surface were found in the cladding near failure location in addition to the main through wall crack. The incipient cracks were filled with interaction products and hydride platelets were present at tip of the cracks. It was concluded from the observations that the primary cause of failure was the presence of a part-wall defect in the end cap weld of the fuel pins. These defects opened up during reactor operation leading to steam ingress into the fuel, which caused high fuel centre temperature and severe fuel-cladding interaction resulting in secondary failures. A more stringent inspection and quality control of end plug weld during fabrication using ultrasonic test has been recommended to avoid such failure. (author)

  14. Microbial Condition of Water Samples from Foreign Fuel Storage Facilities

    International Nuclear Information System (INIS)

    Berry, C.J.

    1998-01-01

    In order to assess the microbial condition of foreign spent nuclear fuel storage facilities and their possible impact on SRS storage basins, twenty-three water samples were analyzed from 12 different countries. Fifteen of the water samples were analyzed and described in an earlier report (WSRC-TR-97-00365 [1]). This report describes nine additional samples received from October 1997 through March 1998. The samples include three from Australia, two from Denmark and Germany and one sample from Italy and Greece. Each water sample was analyzed for microbial content and activity as determined by total bacteria, viable aerobic bacteria, viable anaerobic bacteria, viable sulfate-reducing bacteria, viable acid-producing bacteria and enzyme diversity. The results for each water sample were then compared to all other foreign samples analyzed to date and monthly samples pulled from the receiving basin for off-site fuel (RBOF), at SRS. Of the nine samples analyzed, four samples from Italy, Germany and Greece had considerably higher microbiological activity than that historically found in the RBOF. This microbial activity included high levels of enzyme diversity and the presence of viable organisms that have been associated with microbial influenced corrosion in other environments. The three samples from Australia had microbial activities similar to that in the RBOF while the two samples from Denmark had lower levels of microbial activity. These results suggest that a significant number of the foreign storage facilities have water quality standards that allow microbial proliferation and survival

  15. Post-Irradiation Examination Test of the Parts of X-Gen Nuclear Fuel Assembly

    International Nuclear Information System (INIS)

    Ahn, S. B.; Ryu, W. S.; Choo, Y. S.

    2008-08-01

    The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this report are used to produce the irradiation data of the grid 1x1 cell spring, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300 deg. C during about 100 days From the spring test of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor

  16. Observation on the irradiation behavior of U-Mo alloy dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.; Park, Jong-Man

    2000-01-01

    Initial results from the postirradiation examination of high-density dispersion fuel test RERTR-3 are discussed. The U-Mo alloy fuels in this test were irradiated to 40% U-235 burnup at temperature ranging from 140 0 C to 240 0 C. Temperature has a significant effect on overall swelling of the test plates. The magnitude of the swelling appears acceptable and no unstable irradiation behavior is evident. (author)

  17. LOCA scenario tests of irradiated fuel rod specimens

    International Nuclear Information System (INIS)

    Scott, Harold

    2004-01-01

    Full text: The NRC's cladding performance program at Argonne National Laboratory (ANL) is testing fueled high-burnup segments subjected to LOCA integral phenomena. The data are provided to NRC and the nuclear industry for their independent assessment of the adequacy of licensing criteria for LOCA events. The tests are being conducted with high-burnup 30 cm segments from Limerick (9x9 Zry-2) and H.B. Robinson (15x15 Zry-4) reactors. Prior to testing, sibling samples are characterized with respect to fuel morphology, fuel-cladding bond, cladding oxide layer thickness, hydrogen content and high-temperature steam oxidation kinetics. Specimens that survive quench are subjected to four-point bend tests, followed by local diametral compression tests. The retention of post-quench ductility is a more limiting requirement than surviving thermal stresses during quench. Companion tests are conducted with unirradiated cladding to generate baseline data for comparison with the high-burnup fuel results. LOCA integral tests have the following sequential steps: stabilization of temperature, internal pressure and steam flow at 300 C, ramping of temperature (∼5C/s) through ballooning and burst to 1204 C, hold at 1204 C for 1-5 minutes, slow-cooling (∼3C/s) to 800 C, and water quenching at ∼800C. Two high-burnup tests were completed in 2002 with Limerick BWR rod segments: ramp to burst in argon followed by slow cooling; and the LOCA test with 5-minute hold time at 1204 C, followed by slow cooling. With the exception of burst-opening shape, results for burst temperature, burst pressure, burst length, and ballooning strain profile are more similar to, than different from, results for unirradiated Zry-2 cladding exposed to the same time-temperature history. The 3rd Limerick test with quench was performed in December 2003, and a 4th Limerick test was performed in March 2004. Tests on high-burnup Robinson PWR fuel segments are scheduled to begin in June 2004. The presentation points

  18. Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

    International Nuclear Information System (INIS)

    Ikeuchi, Hirotomo; Sano, Yuichi; Shibata, Atsuhiro; Koizumi, Tsutomu; Washiya, Tadahiro

    2013-01-01

    An efficient dissolution process was established for future reprocessing in which mixed-oxide (MOX) fuels with high plutonium contents and dissolver solution with high heavy-metal (HM) concentrations (more than 500 g dm -3 ) will be treated. This dissolution process involves short stroke shearing of fuels (∼10 mm in length). The dissolution kinetics of irradiated MOX fuels and the effects of the Pu content, HM concentration, and fuel form on the dissolution rate were investigated. Irradiated fuel was found to dissolve as 10 2 -10 3 times fast as non-irradiated fuel, but the rate decreased with increasing Pu content. Kinetic analysis based on the fragmentation model, which considers the penetration and diffusion of nitric acid through fuel matrices prior to chemical reaction, indicated that the dissolution rate of irradiated fuel was affected not only by the volume ratio of liquid to solid (L/S ratio) but also by the exposed surface area per unit mole of nitric acid (A/m ratio). The penetration rate of nitric acid is expected to be decreased at high HM concentrations by a reduction in the L/S ratio, but enhanced by shearing the fuel pieces with short strokes and thus enlarging the A/m ratio. (author)

  19. Calibration of the enigma code for Finnish reactor fuel with support from experimental irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Kelppe, S; Ranta-Puska, K [VTT Energy, Jyvaeskylae (Finland)

    1997-08-01

    Assessment by VTT of the ENIGMA fuel performance code, the original version by Nuclear Electric plc of the UK amended by a set of WWER specific materials correlations, is described. The given examples of results include analyses for BWR 9 x 9 fuel, BWR fuel irradiated in the reinstrumented test of an international Riso project, pre-characterized commercial WWER fuel irradiated in Loviisa reactor in Finland, and instrumented WWER test fuel irradiations in the MR reactor in Russia. The effects of power uncertainty and some model parameters are discussed. Considering the fact that the described cases all mean prototypic application of the code, the results are well encouraging. The importance of the accuracy in temperature calculations is emphasized. (author). 2 refs, 12 figs, 1 tab.

  20. Examinations of the irradiation behaviour of U3Si2 test fuel plates with low enrichment

    International Nuclear Information System (INIS)

    Muellauer, J.

    1989-01-01

    Five low-enriched (19.7% 235 U), high-density (4.7 gU/cm/ 3 ) U 3 Si 2 -test fuel plates (miniplates) with different fine grain contents have been qualified under irradiation. During the course of irradiation up to burnup of 63% 235 U depletion, no released fractions of gaseous or solid fission products from the fuel plate to the rig coolant were detected. The measured swelling rate of the fuel zone (meat) is less than 0.45% ΔV/10 20 fissions/cm 3 the blister-threshold temperature of the fuel plates is above 520 0 C. The favourable irradiation behavior of the U 3 Si 2 fuel plates was not influenced by using higher amounts of fine grained particles (40% [de

  1. The Defect Inspection on the Irradiated Fuel Rod by Eddy Current Test

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, Y. K.; Kim, E. K.

    1996-01-01

    The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase and amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable

  2. 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)

  3. 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)

  4. Irradiation effects and micro-structural changes in large grain uranium dioxide fuel investigated by micro-beam X-ray diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Mieszczynski, C. [NES and SYN, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Kuri, G., E-mail: goutam.kuri@psi.ch [NES and SYN, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Degueldre, C.; Martin, M.; Bertsch, J.; Borca, C.N.; Grolimund, D. [NES and SYN, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Delafoy, Ch. [AREVA NP, 10 Rue Juliette Récamier, 69456 Lyon Cedex 06 (France); Simoni, E. [Institut de Physique Nucléaire, Université Paris-Sud, 91406 Orsay (France)

    2014-01-15

    Microstructural changes in a set of commercial grade UO{sub 2} fuel samples have been investigated using synchrotron based micro-focused X-ray fluorescence (μ-XRF) and X-ray diffraction (μ-XRD) techniques. The results are associated with conventional UO{sub 2} materials and relatively larger grain chromia-doped UO{sub 2} fuels, irradiated in a commercial light water reactor plant (average burn-up: 40 MW d kg{sup −1}). The lattice parameters of UO{sub 2} in fresh and irradiated specimens have been measured and compared with theoretical predictions. In the pristine state, the doped fuel has a somewhat smaller lattice parameter than the standard UO{sub 2} as a result of chromia doping. Increase in micro-strain and lattice parameter in irradiated materials is highlighted. All irradiated samples behave in a similar manner with UO{sub 2} lattice expansion occurring upon irradiation, where any Cr induced effect seems insignificant and accumulated lattice defects prevail. Elastic strain energy densities in the irradiated fuels are also evaluated based on the UO{sub 2} crystal lattice strain and non-uniform strain. The μ-XRD patterns further allow the evaluation of the crystalline domain size and sub-grain formation at different locations of the irradiated UO{sub 2} pellets.

  5. Irradiation-induced growth of zircaloy and its effects on the mechanical design of fuel assemblies

    International Nuclear Information System (INIS)

    Yao Pu

    1991-01-01

    Zircaloy growth could be induced due to irradiation. The ammount of growth is described as a function of texture, irradiation temperature, fast neutron fluence and the reduction of cold work, and it should be given great attention in the mechanical design of fuel assemblies

  6. Pathfinder irradiation of advanced fuel (Th/U mixed oxide) in a power reactor

    International Nuclear Information System (INIS)

    Brant Pinheiro, R.

    1993-01-01

    Within the joint Brazilian-German cooperative R and D Program on Thorium Utilization in Pressurized Water Reactors carried out from 1979 to 1988 by Nuclebras/CDTN, KFA-Juelich, Siemens/KWU and NUKEM, a pathfinder irradiation of Th/U mixed oxide fuel in the Angra 1 nuclear power reactor was planned. The objectives of this irradiation testing, the irradiation strategy, the work performed and the status achieved at the end of the joint Program are presented. (author)

  7. Measurement of fission gases released by HTR irradiation samples in the BR 2 reactor (Mol/Belgium)

    International Nuclear Information System (INIS)

    Koetter, H.; Mueller, H.B.

    1975-04-01

    During the irradiation of HTR-test fuel (coated particles resp. compacts) the release rate of fission gases is measured automatically. For that purpose the γ-spectrum of the sweep gas is analyzed with respect to the isotopes Xe-133, Xe-135, Xe-135sup(m), Xe-137, Xe-138, Kr-85sup(m), Kr-87, Kr-88, Kr-89. Gas sampling, spectral analysis, data handling and evaluation are controlled by computer. (orig.) [de

  8. Effect of gamma irradiation on the storage and vitamin C concentration of allium cepa onion samples

    International Nuclear Information System (INIS)

    Saleh, M. A. M.

    2013-07-01

    This study was done to investigate the effect of gamma irradiation on storage, germination and vitamin C concentration of local onion (Allium cepa). 5 onion samples were irradiated with 5 different radiation doses (0.1, 0.2, 0.3, 0.4, and 0.5 k Gray) using cobalt-60 irradiator (Nor din) compared to non irradiated samples as controls. The irradiated and control onions were stored at room temperature for three months. The results of this study showed that the non irradiated samples were either deteriorated or grown while all the irradiated samples were not. Regarding the concentration of vitamin C it was clear that it decreased with the dose increase from 30.53 to 14.44 mg/100 g. This study concluded that the irradiation is very effective in prevention of spoilage, elongation of germination period and decrease of vitamin C concentration.(Author)

  9. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    Science.gov (United States)

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

  10. Fuel ethanol production from sweet sorghum bagasse using microwave irradiation

    International Nuclear Information System (INIS)

    Marx, Sanette; Ndaba, Busiswa; Chiyanzu, Idan; Schabort, Corneels

    2014-01-01

    Sweet sorghum is a hardy crop that can be grown on marginal land and can provide both food and energy in an integrated food and energy system. Lignocellulose rich sweet sorghum bagasse (solid left over after starch and juice extraction) can be converted to bioethanol using a variety of technologies. The largest barrier to commercial production of fuel ethanol from lignocellulosic material remains the high processing costs associated with enzymatic hydrolysis and the use of acids and bases in the pretreatment step. In this paper, sweet sorghum bagasse was pretreated and hydrolysed in a single step using microwave irradiation. A total sugar yield of 820 g kg −1 was obtained in a 50 g kg −1 sulphuric acid solution in water, with a power input of 43.2 kJ g −1 of dry biomass (i.e. 20 min at 180 W power setting). An ethanol yield based on total sugar of 480 g kg −1 was obtained after 24 h of fermentation using a mixed culture of organisms. These results show the potential for producing as much as 0.252 m 3  tonne −1 or 33 m 3  ha −1 ethanol using only the lignocellulose part of the stalks, which is high enough to make the process economically attractive. - Highlights: • Different sweet sorghum cultivars were harvested at 3 and 6 months. • Sweet sorghum bagasse was converted to ethanol. • Microwave pretreatment and hydrolysis was done in a single step. • Sugar rich hydrolysates were converted to ethanol using co-fermentation

  11. 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

  12. FUJI - a comparative irradiation test with pellet, sphere-pac, and vipac fuel

    International Nuclear Information System (INIS)

    Hellwig, C.; Bakker, K.; Ozawa, T.; Nakamura, M.; Kihara, Y.

    2004-01-01

    Particle fuels such as sphere-pac and vipac fuels have been considered as promising fuel systems for fast reactors, due to their inherent potential in remote operation, cost reduction and incineration of minor actinides or low-decontaminated plutonium. The FUJI test addresses the questions of fabrication of MOX particle fuels with high Pu content (20%) and its irradiation behaviour during the start-up phase. Four kinds of fuel, i.e. MOX sphere-pac, MOX vipac, MOX pellet and Np-MOX sphere-pac fuel, have been and will be simultaneously irradiated under identical conditions in the High Flux Reactor in Petten. First results show that the particle fuel undergoes a dramatic structure change already at the very beginning of the irradiation when the maximum power is reached. The structural changes, i.e. the formation of a central void and the densification of fuel, decrease the fuel central temperature. Thus the fast and strong restructuring helps to prevent central fuel melting at high power levels. (authors)

  13. 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

  14. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Science.gov (United States)

    2010-01-01

    ... fuel in transit. 73.37 Section 73.37 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.37 Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1...

  15. In-pile irradiation test program and safety analysis report of the KAERI fuel for HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Wan; Ryu, Woo Suck; Byun, Taek Sang; Park, Jong Man; Lee, Byung Chul; Kim, Hack No; Park, Hee Tae; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-05-01

    Localization of HANARO fuel has been carried out successfully, and design and fabrication technologies of the fuel are recently arrived the final stage of development. The performance of the fuel which has been fabricated in KAERI is confirmed through out-of-pile characterization, and the quality assurance procedure and assessment criteria are described. In order to verify the KAERI fuel, thus, in-pile irradiation test program of the KAERI fuel is scheduled in HANARO. This report summarizes the in-pile testing schedule, design documents of test rods and assemblies, fabrication history and out-of-pile characteristics of test rods, irradiation test condition and power history, post-irradiation examination scheme, linear power generation distribution, and safety analysis results. The design code for HANARO fuel is used to analyze the centerline temperature and swelling of the KAERI fuels. The results show that at 120 kW/m of linear power the maximum centerline temperature is 267 deg C which is much lower than the limitation temperature of 350 deg C, and that the swelling is 9.3 % at 95 at% lower than criterion of 20 %. Therefore, the KAERI fuels of this in-pile irradiation test is assessed to show good performance of integrity and safety in HANARO. 10 tabs., 7 figs., 3 refs. (Author).

  16. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Science.gov (United States)

    2010-01-01

    ... fuel and nuclear waste. 71.97 Section 71.97 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... advance notification of transportation of nuclear waste was published in the Federal Register on June 30...

  17. Extract of the report of the working party on the handling of irradiated fuel

    International Nuclear Information System (INIS)

    Berest, P.

    1983-01-01

    The French government has requested a working party with Prof. Neel in the chair to submit a report on the handling of irradiated fuel. This part of the report concerns the retreated fuels. It gives important elements for the debate and formulates recommendations for radioactive waste management [fr

  18. 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)

  19. Discussion on the re-irradiated fuel assembly with damaged guide vanes

    International Nuclear Information System (INIS)

    Li Ligang

    2013-01-01

    In January 2011, during the second plant of CNNC Nuclear Power Operations Management Co., Ltd.(hereinafter referred to as the second plant) refueling outage, the visual inspection found the guide vanes of fuel assembly A had felling off. After the National Nuclear Safety Administration (NNSA) estimated and approved, the fuel assembly A was reloaded in the specified location of reactor core. During the refueling outage in March 2012, the fuel assembly A was removed again from the reactor core. Visual inspection confirmed that the fuel assembly A was complete and without abnormal changes. The practice provides reference for re-irradiated of fuel assembly with the same type of damaged guide vanes, and provides case support for standard development for the same type of re-irradiated fuel assembly with damaged guide vanes. (author)

  20. Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

    1999-01-01

    The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

  1. Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin

    International Nuclear Information System (INIS)

    Matos, J.E.

    1994-12-01

    This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy's Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006

  2. Natural draught centralized dry store for irradiated fuel and active waste

    International Nuclear Information System (INIS)

    Bradley, N.; Brown, G.A.

    1981-01-01

    A modular design is described for the long term dry storage of irradiated fuel and vitrified fission products. The specification set by the Central Electricity Generating Board for the AGR fuel store was that the store should be capable of accommodating the lifetime discharge from 10 AGR reactors (7200 tonnes of irradiated fuel) and be cooled by natural convection. The fuel assemblies should be enclosed in individual steel containers. The store has an area for drying the AGR fuel and containering. The single dry cell storage capacities are, 5 years output from 1300 MWe station stored as fuel elements, or 14 year output from 1300 MWe thermal reactors stored as vitrified fission products. (U.K.)

  3. Irradiation of a 19 pin subassembly with mixed carbide fuel in KNK II

    Science.gov (United States)

    Geithoff, D.; Mühling, G.; Richter, K.

    1992-06-01

    The presentation deals with the fabrication, irradiation and nondestructive postirradiation examinations of LMR fuel pins with mixed (U, Pu)-carbide fuels. The mixed carbide fuel was fabricated by the European Institute of Transuranium Elements using various fabrication procedures. Fuel composition varied therefore in a wide range of tolerances with respect to oxygen and phase content and microstructure. The 19 carbide pins were irradiated in the fast neutron flux of the KNK II reactor to a burn-up of about 7 at% without any failure in the centre of a KNK "carrier element" at a maximum linear rating of 800 W/cm. After dismantling in the Hot Cells of KfK nondestructive examinations were carried out comprising dimensional controls, radiography, γ-scanning and eddy-current testing. The results indicate differences in fuel behaviour with respect to composition of the fuel.

  4. 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

  5. Comparison of irradiation behavior of different uranium silicide dispersion fuel element designs

    International Nuclear Information System (INIS)

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

    1995-01-01

    Calculations of fuel swelling of U 3 SiAl-Al and U 3 Si 2 were performed for various dispersion fuel element designs. Breakaway swelling criteria in the form of critical fuel volume fractions were derived with data obtained from U 3 SiAl-Al plate irradiations. The results of the analysis show that rod-type elements remain well below the pillowing threshold. However, tubular fuel elements, which behave essentially like plates, will likely develop pillows or blisters at around 90% 235 U burnup. The U 3 Si 2 -Al compounds demonstrate stable swelling behavior throughout the entire burnup range for all fuel element designs

  6. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    OpenAIRE

    ALEKSEY. L. IZHUTOV; VALERIY. V. IAKOVLEV; ANDREY. E. NOVOSELOV; VLADIMIR. A. STARKOV; ALEKSEY. A. SHELDYAKOV; VALERIY. YU. SHISHIN; VLADIMIR. M. KOSENKOV; ALEKSANDR. V. VATULIN; IRINA. V. DOBRIKOVA; VLADIMIR. B. SUPRUN; GENNADIY. V. KULAKOV

    2013-01-01

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; th...

  7. State of the VVER-1000 spent U-Gd fuel rods based on the results of post-irradiation examinations

    International Nuclear Information System (INIS)

    Shevlyakov, G.; Zvir, E.; Strozhuk, A.; Polenok, V.; Sidorenko, O.; Volkova, I.; Nikitin, O.

    2015-01-01

    The present paper is devoted to post-irradiation examinations (PIE) of U-Gd fuel rods with different geometry of the fuel pellets irradiated as part of the VVER-1000 fuel assembly. As evidenced by their PIE data, they did not exhaust their service life based on the main parameters (geometrical dimensions, corrosion state, and release of fission product gases). (author)

  8. 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.

  9. 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

  10. The analysis of the RA reactor irradiated fuel cooling in the spent fuel pool; Analiza hladjenja ozracenog goriva u bazenu za odlaganje reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Vrhovac, M; Afgan, N; Spasojevic, D; Jovic, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1985-07-01

    According to the RA reactor exploitation plan the great quantity of the irradiated spent fuel will be disposed in the reactor spent fuel pool after each reactor campaign which will including the present spent fuel inventory increase the residual power level in the pool and will soon cause the pool capacity shortage. To enable the analysis of the irradiated fuel cooling the pool and characteristic spent fuel canister temperature distribution at the residual power maximum was done. The results obtained under the various spent fuel cooling conditions in the pit indicate the normal spent fuel thermal load even in the most inconvenient cooling conditions. (author)

  11. 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.

  12. A Study on Cell Size of Irradiated Spacer Grid for PWR Fuel

    International Nuclear Information System (INIS)

    Jin, Y. G.; Kim, G. S.; Ryu, W. S. and others

    2014-01-01

    The spacer grids supporting the fuel rods absorb vibration impacts due to the reactor coolant flow, and grid spring force decreases under irradiation. This reduction of contact force might cause grid-to-rod fretting wear. The fretting failure of the fuel rod is one of the recent significant issues in the nuclear industry from an economical as well as a safety concern. Thus, it is important to understand the characteristics of cell spring behavior and the change in size of grid cells for an irradiated spacer grid. In the present study, the dimensional measurement of a spacer grid was conducted to investigate the cell size of an irradiated spacer grid in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. To evaluate the fretting wear performance of an irradiated spacer grid, hot cell tests were carried out at IMEF of KAERI. Hot cell examinations include dimensional measurements for the irradiated spacer grid. The change of cell sizes was dependent on the direction of the spacer grids, leading to significant gap variations. It was found that the change in size of the cell springs due to irradiation-induced stress relaxation and creep during the fuel residency in the reactor core affect the contact behavior between the fuel rod and the cell spring

  13. Analysis of volatile organic compounds and sensory characteristics of pork loin samples irradiated to high doses

    International Nuclear Information System (INIS)

    Hou Zhengchi; Sun Dakuan; Qin Zongying; Jin Jiang; Zhu Liandi; Yao Side; Sheng Kanglong

    2005-01-01

    Fresh pork loin samples, protein enzyme inactivated at (72 ± 3) degree C and vacuum packaged, were irradiated to up to 45 kGy at -20 degree C by 60 Co γ-rays. The irradiated samples were examined by various kinds of method to study high dose irradiation effects of sensory changes (meat color and off-odor), transverse shearing strength, weight loss in steam cooking, volatile organic compounds, and lipid oxidation. The results showed that the high dose irradiation produced no serious effects to the pork loin samples, and volunteer responses showed fine acceptability to the irradiated meat. (authors)

  14. Defect trap model of gas behaviour in UO2 fuel during irradiation

    International Nuclear Information System (INIS)

    Szuta, A.

    2003-01-01

    Fission gas behaviour is one of the central concern in the fuel design, performance and hypothetical accident analysis. The report 'Defect trap model of gas behaviour in UO 2 fuel during irradiation' is the worldwide literature review of problems studied, experimental results and solutions proposed in related topics. Some of them were described in details in the report chapters. They are: anomalies in the experimental results; fission gas retention in the UO 2 fuel; microstructure of the UO 2 fuel after irradiation; fission gas release models; defect trap model of fission gas behaviour; fission gas release from UO 2 single crystal during low temperature irradiation in terms of a defect trap model; analysis of dynamic release of fission gases from single crystal UO 2 during low temperature irradiation in terms of defect trap model; behaviour of fission gas products in single crystal UO 2 during intermediate temperature irradiation in terms of a defect trap model; modification of re-crystallization temperature of UO 2 in function of burnup and its impact on fission gas release; apparent diffusion coefficient; formation of nanostructures in UO 2 fuel at high burnup; applications of the defect trap model to the gas leaking fuel elements number assessment in the nuclear power station (VVER-PWR)

  15. 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.

  16. Re-irradiation tests of spent fuel at JMTR by means of re-instrumentation technique

    International Nuclear Information System (INIS)

    Nakamura, Jinichi; Shimizu, Michio; Endo, Yasuichi; Nabeya, Hideaki; Ichise, Kenichi; Saito, Junichi; Oshima, Kunio; Uetsuka, Hiroshi

    1999-01-01

    JAERI has developed re-irradiation test procedures of spent fuel irradiated at commercial reactor by means of re-instrumentation technique. Full length rods irradiated at commercial LWRs were re-fabricated to short length rods, and rod inner pressure gauges and fuel center thermocouples were re-instrumented to the rods. Re-irradiation tests to study the fuel behavior during power change were carried out by means of BOCA/OSF-1 facility at the JMTR. In the tests to study the fission gas release during power change, the rod inner pressure increase was observed during power change, especially during power reduction. The fission gas release during power reduction is estimated to be the release from fission gas bubbles on the grain boundary caused by the thermal stress in the pellet during power reduction. Re-irradiation test of gadolinia added fuel was performed by means of dual re-instrumentation technique (fuel center thermocouples and rod inner pressure gauge). A stepwise fission gas release during power change, and the following fuel center temperature change due to gap conductance change were observed. (author)

  17. Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel

    Science.gov (United States)

    Zweifel, T.; Valot, Ch.; Pontillon, Y.; Lamontagne, J.; Vermersch, A.; Barrallier, L.; Blay, T.; Petry, W.; Palancher, H.

    2014-09-01

    U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding.

  18. 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)

  19. Development of high resolution x-ray CT technique for irradiated fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ishimi, Akihiro; Katsuyama, Kozo; Maeda, Koji; Asaga, Takeo [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    High X-ray CT technique was developed to observe the irradiation performance of FBR fuel assembly and MOX fuel. In this technique, the high energy X-ray pulse (12MeV) was used synchronizing detection system with the X-ray pulse to reduce the effect of the gamma ray emissions from the irradiated fuel assembly. In this study, this technique was upgraded to obtain high resolution X-ray CT image. In this upgrading, the collimator which had slit width of 0.1 mm and X-ray detector of a highly sensitive silicon semiconductor detector (100 channels) was introduced in the X-ray CT system. As a result of these developments, high resolution X-ray CT images could be obtained on the transverse cross section of irradiated fuel assembly. (author)

  20. 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)

  1. 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)

  2. 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)

  3. Irradiation testing of LEU fuels in the SILOE Reactor - Progress report

    International Nuclear Information System (INIS)

    Merchie, Francis; Baas, Claude; Martel, Patrick

    1985-01-01

    Irradiation of uranium-silicide fuels has continued in the SILOE reactor during the past year. Thickness vs. fission density data from four U 3 Si plates containing 5.5 and 6.0 g U/cm 3 have been analyzed, and the results are presented. The irradiation of a full 60 g U/cm 3 U 3 Si element has begun. In addition, four U 3 Si 2 plates containing 20 to 54 g U/cm 3 are now being irradiated. These irradiations and future plans are discussed in the paper. (author)

  4. Pre-irradiation testing of experimental fuel elements

    International Nuclear Information System (INIS)

    Basova, B.G.; Davydov, E.F.; Dvoretskij, V.G.; Ivanov, V.B.; Syuzev, V.N.; Timofeev, G.A.; Tsykanov, V.A.

    1979-01-01

    The problems of testing of experimental fuel elements of nuclear reactors on the basis of complex accountancy of the factors defining operating capacity of the fuel elements are considered. The classification of the parameters under control and the methods of initial technological testing, including testing of the fuel product, cladding and fished fuel element, is given. The requirements to the apparatus used for complex testing are formulated. One of the possible variants of representation of the information obtained in the form of the input certificate of a single fuel element under study is proposed. The processing flowsheet of the gathered information using the computer is given. The approach under consideration is a methodological basis of investigation of fuel element operating life at the testing stage of the experimental fuel elements

  5. Management of spent fuel from research and prototype power reactors and residues from post-irradiation examination of fuel

    International Nuclear Information System (INIS)

    1989-09-01

    The safe and economic management of spent fuel is important for all countries which have nuclear research or power reactors. It involves all aspects of the handling, transportation, storage, conditioning and reprocessing or final disposal of the spent fuel. In the case of spent fuel management from power reactors the shortage of available reprocessing capacity and the rising economic interest in the direct disposal of spent fuel have led to an increasing interest in the long term storage and management of spent fuel. The IAEA has played a major role in coordinating the national activities of the Member States in this area. It was against this background that the Technical Committee Meeting on ''Safe Management of Spent Fuel From Research Reactors, Prototype Power Reactors and Fuel From Commercial Power Reactors That Has Been Subjected to PIE (Post Irradiated Examination)'' (28th November - 1st December 1988) was organised. The aims of the current meeting have been to: 1. Review the state-of-the-art in the field of management of spent fuel from research and prototype power reactors, as well as the residues from post irradiation examination of commercial power reactor fuel. The emphasis was to be on the safe handling, conditioning, transportation, storage and/or disposal of the spent fuel during operation and final decommissioning of the reactors. Information was sought on design details, including shielding, criticality and radionuclide release prevention, heat removal, automation and remote control, planning and staff training; licensing and operational practices during each of the phases of spent fuel management. 2. Identify areas where additional research and development are needed. 3. Recommend areas for future international cooperation in this field. Refs, figs and tabs

  6. Laser microsampling method for determination of retained fission gas in irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Graczyk, D.G.; Bandyopadhyay, G.; Gehl, S.M.; Hughes, J.P.; Goodspeed, H.T.

    1979-10-01

    A small ruby laser adapted to fire through a microscope is used to release fission gases from specific sites on a plane surface of an irradiated fuel specimen. Interaction of the focused laser pulse with the specimen surface results in a conical crater from which sampled material has been vaporized; the crater is surrounded by a heat-affected zone in which intergranular fracture and grain separation allow release of grain-boundary gases. Procedures for measuring the amount of krypton-85 released by laser heating and the volume of material from which the release occurred are presented. The data obtained may be used to obtain local krypton fission-gas concentrations and the intragranular/intergranular distribution

  7. Post-Irradiation Examination of Fuel Pin R54-F20A, Irradiated in a NaK Environment. RCN Report

    International Nuclear Information System (INIS)

    Kwast, H.

    1972-12-01

    Fuel pin R54-F20A has been irradiated in a NaK-environment. Temperature measurements in the NaK were carried out at average linear fission powers of 552 and 825 W/cm respectively. A maximum average canning temperature of 920°C was reached. The fuel pin was irradiated for about 50 minutes at the maximum irradiation conditions, while the total irradiation time was two hours. The irradiation had to be broken off before the end condition was reached because of malfunctioning of the fuelfailure detection system. No power peaking did occur at the upper and lower interfaces between the 50%-enriched UO 2 - and the natural UO 2 + 8 w/o UB 4 pellet. About 35% of the fuel has molten, but the fuel pin did not fail. The irradiation has been carried out in the Poolside Facility (PSF) of the High Flux Reactor (HFR) at Petten. (author)

  8. Transfer of test samples and wastes between post-irradiation test facilities (FMF, AGF, MMF)

    International Nuclear Information System (INIS)

    Ishida, Yasukazu; Suzuki, Kazuhisa; Ebihara, Hikoe; Matsushima, Yasuyoshi; Kashiwabara, Hidechiyo

    1975-02-01

    Wide review is given on the problems associated with the transfer of test samples and wastes between post-irradiation test facilities, FMF (Fuel Monitoring Facility), AGF (Alpha Gamma Facility), and MMF (Material Monitoring Facility) at the Oarai Engineering Center, PNC. The test facilities are connected with the JOYO plant, an experimental fast reactor being constructed at Oarai. As introductory remarks, some special features of transferring irradiated materials are described. In the second part, problems on the management of nuclear materials and radio isotopes are described item by item. In the third part, the specific materials that are envisaged to be transported between JOYO and the test facilities are listed together with their geometrical shapes, dimensions, etc. In the fourth part, various routes and methods of transportation are explained with many block charts and figures. Brief explanation with lists and drawings is also given to transportation casks and vessels. Finally, some future problems are discussed, such as the prevention of diffusive contamination, ease of decontamination, and the identification of test samples. (Aoki, K.)

  9. Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

    1990-05-01

    The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs

  10. A method to evaluate fission gas release during irradiation testing of spherical fuel - HTR2008-58184

    International Nuclear Information System (INIS)

    Van Der Merwet, H.; Venter, J.

    2008-01-01

    The evaluation of fission gas release from spherical fuel during irradiation testing is critical to understand expected fuel performance under real reactor conditions. Online measurements of Krypton and Xenon fission products explain coated particle performance and contributions from graphitic matrix materials used in fuel manufacture and irradiation rig materials. Methods that are being developed to accurately evaluate fission gas release are described here together with examples of evaluations performed on irradiation tests HFR-K5, -K6 and EU1bis. (authors)

  11. Water sampling device for fuel rod failure monitoring

    International Nuclear Information System (INIS)

    Oogami, Hideaki; Echigoya, Hironori; Matsuoka, Tesshi.

    1991-01-01

    The device of the present invention accurately samples coolants in a channel box as sampling water even if the upper end of the channel box of a fuel assembly is positioned at the same height or lower than the upper end of an upper lattice plate. An existent device comprises an outer cap, an inner cap, an air supply pipe and a water sampling tube. In addition, the device of the present invention comprises a sealing material disposed at the end of the outer cap for keeping liquid sealing with the upper lattice plate and a water level monitoring pipe extended to lower than the inner cap passing through the liquid sealing of the outer cap for sucking the atmosphere in the outer cap. Pressurized air is sent through the air supply pipe, to lower the water level of the coolants in the outer cap and the water level monitoring pipe sucks the pressurized air, by which the inside and the outside of the channel box are partitioned. Subsequently, if the sample water is sampled by a sampling tube, sampling water which enables accurate evaluation for radioactivity concentration in the fuel assembly can be obtained. (I.S.)

  12. The use of isotopic correlation technique for determination of sup(241)Am and sup(243)Am concentration in nuclear irradiated fuels

    International Nuclear Information System (INIS)

    Souza Sarkis, J.E. de.

    1990-01-01

    In the last years the isotopic correlation technique is emerging as a powerful tool for the determination of concentration and isotopic composition of heavy nuclides in the nuclear fuel cycle. Accordingly, this technique has gained significant importance for the safeguard of the nuclear materials as well as for the accounting and build up of actinides elements in the irradiated nuclear fuels. In this work 42 isotopic correlations between the nuclides sup(241)Am and sup(243)Am and post irradiation isotopic data of 7 samples from fuel element BE-124 and 1 sample from fuel element BE-120 from the Obrigheim pressurized water nuclear power reactor, Federal Republic of Germany, were proposed. These isotopic correlations allowed to estimate the isotopic concentrations of sup(241)Am and sup(243)Am with an average deviation, relative to the experimental data obtained from isotopic dilution mass spectrometry technique, of 10%. These results are more precise than those found using the computer code ORIGEN 2 demonstrating the great potential of this technique for the determination of isotopic concentration and build up of those nuclides in irradiated nuclear fuels. The analytical and other experimental aspects of the post irradiation isotopic analysis of nuclear fuels are also discussed. (author)

  13. High burnup, high power irradiation behavior of helium-bonded mixed carbide fuel pins

    International Nuclear Information System (INIS)

    Levine, P.J.; Nayak, U.P.; Boltax, A.

    1983-01-01

    Large diameter (9.4 mm) helium-bonded mixed carbide fuel pins were successfully irradiated in EBR-II to high burnup (12%) at high power levels (100 kW/m) with peak cladding midwall temperatures of 550 0 C. The wire-wrapped pins were clad with 0.51-mm-thick, 20% cold-worked Type 316 stainless steel and contained hyperstoichiometric (Usub(0.8)Pusub(0.2))C fuel covering the smeared density range from 75-82% TD. Post-irradiation examinations revealed: extensive fuel-cladding mechanical interaction over the entire length of the fuel column, 35% fission gas release at 12% burnup, cladding carburization and fuel restructuring. (orig.)

  14. 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

  15. Finite element method programs to analyze irradiation behavior of fuel pellets

    International Nuclear Information System (INIS)

    Yamada, Rayji; Harayama, Yasuo; Ishibashi, Akihiro; Ono, Masao.

    1979-09-01

    For the safety assessment of reactor fuel, it is important to grasp local changes of fuel pins due to irradiation in a reactor. Such changes of fuel result mostly from irradiation of fuel pellets. Elasto-plastic analysis programs based on the finite element method were developed to analyze these local changes. In the programs, emphasis is placed on the analysis of cracks in pellets; the interaction between cracked-pellets and cladding is not taken into consideration. The two programs developed are FEMF3 based on a two-dimensional axially symmetric model (r-z system) and FREB4 on a two-dimensional plane model (r-theta system). It is discussed in this report how the occurrence and distribution of cracks depend on heat rate of the fuel pin. (author)

  16. A review on the welding technology for the sealing of irradiation test fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. W.; Kang, Y. H.; Kim, B. G.; Joo, K. N.; Oh, J. M.; Park, S. J.; Shin, Y. T

    2000-02-01

    For the irradiation test of nuclear fuel in a research reactor, the fuel manufacturing technology should be developed in advance. Highly radioactive fission products are produced and can be released from the fuel materials during irradiation. Therefore, The sealing of the test is one of the most important procedure among the test fuel manufacturing processes, considering its impacts on the safety of a reactor operation.many welding techniques such as TIG, EBW, LBW, upset butt welding and flash welding are applied in sealing the end of fuel elements. These welding techniques are adopted in conjunction with the weld material, weldability, weld joint design and cost effectiveness. For fuel irradiation test, the centerline temperature of fuel pellets is one of the important item to be measured. For this, a thermocouple is installed into the center of the fuel pellet. The sealing of the penetration hole of the thermocouple sheath should be conducted and the hole should be perfectly sealed using the dissimilar metal joining technique. For this purpose, the dissimilar metal welding between zircaloy-4 and Inconel or stainless steel is needed to be developed. This report describes the techniques sealing the end cap and the penetration of a thermocouple sheath by welding. (author)

  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. Trefoil bundles of NPD 7-element size fuel irradiated to 9100 MWd/tonne U

    Energy Technology Data Exchange (ETDEWEB)

    Bain, A S; Christie, J; Daniel, A R

    1964-01-15

    NPD prototype elements (1 in. OD, 19 in. long) were assembled into trefoil bundles and irradiated in the X-5 pressurized-water loop of NRX. The first tests were for only a few weeks but showed that elements made by sheathing UO{sub 2} pellets in Zircaloy-2 behaved well under irradiation; later similar elements were irradiated for 18000 hours to a burn-up of 9100 MWd/tonne U at {integral}kd{theta} = 40 W/cm. The dimensional stability of all the elements was good. Only those subjected to long irradiation showed progressive diametral increases, and these were attributed to relocation of the UO{sub 2} during interim inspections. Length measurements demonstrated that pellet end-dishing is effective in controlling axial expansion, but that for a given depth of dishing the amount of expansion depends on the shoulder width. The extent of grain growth in the UO{sub 2} was compatible with previously reported results when the duration of irradiation, density of the fuel, and variations in growth characteristics of the different batches of UO{sub 2} are considered. The elements taken to high irradiation released up to 135 ml of fission-product gases, which is 2% of the amount formed. The transverse tensile strength of ring samples from the Zircaloy-2 sheaths increased from 75000 to 95000 lb/in{sup 2} at room temperature, but the ductility dropped. The completely brittle fracture of some rings was due to ZrH{sub 2} precipitation. The failure of one element was caused by increased stress due to a higher heat rating, combined with low ductility of the Zircaloy-2 resulting from radiation damage and with precipitation of ZrH{sub 2} because of a lower coolant temperature. The fission-product release from the split was not excessive, and the element was easily withdrawn from the loop after operating at full power for four days from the time of the failure. (author)

  19. An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Louise G [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, S. J. [Los Alamos National Laboratory; Menlove, H. O. [Los Alamos National Laboratory; Schear, M. A. [Los Alamos National Laboratory; Worrall, Andrew [U.K. NNL

    2011-01-13

    Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

  20. Harmonisation of microbial sampling and testing methods for distillate fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hill, G.C.; Hill, E.C. [ECHA Microbiology Ltd., Cardiff (United Kingdom)

    1995-05-01

    Increased incidence of microbial infection in distillate fuels has led to a demand for organisations such as the Institute of Petroleum to propose standards for microbiological quality, based on numbers of viable microbial colony forming units. Variations in quality requirements, and in the spoilage significance of contaminating microbes plus a tendency for temporal and spatial changes in the distribution of microbes, makes such standards difficult to implement. The problem is compounded by a diversity in the procedures employed for sampling and testing for microbial contamination and in the interpretation of the data obtained. The following paper reviews these problems and describes the efforts of The Institute of Petroleum Microbiology Fuels Group to address these issues and in particular to bring about harmonisation of sampling and testing methods. The benefits and drawbacks of available test methods, both laboratory based and on-site, are discussed.

  1. 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)

  2. 40 CFR Appendix II to Part 600 - Sample Fuel Economy Calculations

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Sample Fuel Economy Calculations II... FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Pt. 600, App. II Appendix II to Part 600—Sample Fuel Economy Calculations (a) This sample fuel economy calculation is applicable to...

  3. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1996-07-01

    This circular provides information on shipment of spent fuel subject to regulation by US NRC. It provides a brief description of spent fuel shipment safety and safeguards requirement of general interest, a summary of data for 1979-1995 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

  4. Determination of the origin of unknown irradiated nuclear fuel.

    Science.gov (United States)

    Nicolaou, G

    2006-01-01

    An isotopic fingerprinting method is presented to determine the origin of unknown nuclear material with forensic importance. Spent nuclear fuel of known origin has been considered as the 'unknown' nuclear material in order to demonstrate the method and verify its prediction capabilities. The method compares, using factor analysis, the measured U, Pu isotopic compositions of the 'unknown' material with U, Pu isotopic compositions simulating well known spent fuels from a range of commercial nuclear power stations. Then, the 'unknown' fuel has the same origin as the commercial fuel with which it exhibits the highest similarity in U, Pu compositions.

  5. Determination of the origin of unknown irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Nicolaou, G.

    2006-01-01

    An isotopic fingerprinting method is presented to determine the origin of unknown nuclear material with forensic importance. Spent nuclear fuel of known origin has been considered as the 'unknown' nuclear material in order to demonstrate the method and verify its prediction capabilities. The method compares, using factor analysis, the measured U, Pu isotopic compositions of the 'unknown' material with U, Pu isotopic compositions simulating well known spent fuels from a range of commercial nuclear power stations. Then, the 'unknown' fuel has the same origin as the commercial fuel with which it exhibits the highest similarity in U, Pu compositions

  6. Safety assessment of U–Mo fuel mini plates irradiated in HANARO reactor

    International Nuclear Information System (INIS)

    Jo, Daeseong; Kim, Haksung

    2015-01-01

    Highlights: • Neutronic and thermal-hydraulic analyses of U–Mo fuel irradiated in HANARO reactor. • A mock-up irradiation target was designed and tested to measure the flow rate. • During normal operation, boiling does not occur. • During limiting accidents, boiling occurs. However, fuel integrity is maintained. - Abstract: Neutronic and thermal hydraulic characteristics of U–Mo fuel mini plates irradiated in the HANARO reactor were analyzed for the safety assessment of these plates. A total of eight fuel plates were double-stacked; each stack contained three 8.0 gU/cc U–7Mo fuel plates and one 6.5 gU/cc U–7Mo fuel plate. The neutronic and thermal hydraulic analyses were carried out using the MCNP code and TMAP code, respectively. The core status used in the study was the equilibrium core, and four Control Absorber Rod (CAR) locations were considered: 350 mm, 450 mm, 550 mm, and 650 mm away from the bottom of the core. For the fuels in the lower stack, the maximum heat flux was found at the CAR located at 450 mm. For the fuels in the upper stack, the maximum heat flux was found at the CAR located at 650 mm. The axial power distributions for the upper and lower stacks were selected on the basis of thermal margin analyses. A mock-up irradiation target assembly was designed and tested at the out-of-pile test facility to measure the flow rate through the irradiation site, given that the maximum flow rate through the irradiation site at the HANARO reactor is limited to 12.7 kg/s. For conservative analyses, measurement and correlation uncertainties and engineering hot channel factors were considered. During normal operation, the minimum ONB temperature margins for the lower and upper stacks are 41.6 °C and 31.8 °C, respectively. This means that boiling does not occur. However, boiling occurs during the limiting accidents. Nevertheless, the fuel integrity is maintained since the minimum DNBR are 1.96 for the Reactivity Insertion Accident (RIA) and 2

  7. EVALUATION OF U10MO FUEL PLATE IRRADIATION BEHAVIOR VIA NUMERICAL AND EXPERIMENTAL BENCHMARKING

    Energy Technology Data Exchange (ETDEWEB)

    Samuel J. Miller; Hakan Ozaltun

    2012-11-01

    This article analyzes dimensional changes due to irradiation of monolithic plate-type nuclear fuel and compares results with finite element analysis of the plates during fabrication and irradiation. Monolithic fuel plates tested in the Advanced Test Reactor (ATR) at Idaho National Lab (INL) are being used to benchmark proposed fuel performance for several high power research reactors. Post-irradiation metallographic images of plates sectioned at the midpoint were analyzed to determine dimensional changes of the fuel and the cladding response. A constitutive model of the fabrication process and irradiation behavior of the tested plates was developed using the general purpose commercial finite element analysis package, Abaqus. Using calculated burn-up profiles of irradiated plates to model the power distribution and including irradiation behaviors such as swelling and irradiation enhanced creep, model simulations allow analysis of plate parameters that are either impossible or infeasible in an experimental setting. The development and progression of fabrication induced stress concentrations at the plate edges was of primary interest, as these locations have a unique stress profile during irradiation. Additionally, comparison between 2D and 3D models was performed to optimize analysis methodology. In particular, the ability of 2D and 3D models account for out of plane stresses which result in 3-dimensional creep behavior that is a product of these components. Results show that assumptions made in 2D models for the out-of-plane stresses and strains cannot capture the 3-dimensional physics accurately and thus 2D approximations are not computationally accurate. Stress-strain fields are dependent on plate geometry and irradiation conditions, thus, if stress based criteria is used to predict plate behavior (as opposed to material impurities, fine micro-structural defects, or sharp power gradients), unique 3D finite element formulation for each plate is required.

  8. Characterization of un-irradiated MIMAS MOX fuel by Raman spectroscopy and EPMA

    Science.gov (United States)

    Talip, Zeynep; Peuget, Sylvain; Magnin, Magali; Tribet, Magaly; Valot, Christophe; Vauchy, Romain; Jégou, Christophe

    2018-02-01

    In this study, Raman spectroscopy technique was implemented to characterize un-irradiated MIMAS (MIcronized - MASter blend) MOX fuel samples with average 7 wt.% Pu content and different damage levels, 13 years after fabrication, one year after thermal recovery and soon after annealing, respectively. The impacts of local Pu content, deviation from stoichiometry and self-radiation damage on Raman spectrum of the studied MIMAS MOX samples were assessed. MIMAS MOX fuel has three different phases Pu-rich agglomerate, coating phase and uranium matrix. In order to distinguish these phases, Raman results were associated with Pu content measurements performed by Electron Microprobe Analysis. Raman results show that T2g frequency significantly shifts from 445 to 453 cm-1 for Pu contents increasing from 0.2 to 25 wt.%. These data are satisfactorily consistent with the calculations obtained with Gruneisen parameters. It was concluded that the position of the T2g band is mainly controlled by Pu content and self-radiation damage. Deviation from stoichiometry does not have a significant influence on T2g band position. Self-radiation damage leads to a shift of T2g band towards lower frequency (∼1-2 cm-1 for the UO2 matrix of damaged sample). However, this shift is difficult to quantify for the coating phase and Pu agglomerates given the dispersion of high Pu concentrations. In addition, 525 cm-1 band, which was attributed to sub-stoichiometric structural defects, is presented for the first time for the self-radiation damaged MOX sample. Thanks to the different oxidation resistance of each phase, it was shown that laser induced oxidation could be alternatively used to identify the phases. It is demonstrated that micro-Raman spectroscopy is an efficient technique for the characterization of heterogeneous MOX samples, due to its low spatial resolution.

  9. Steady-state irradiation testing of U-Pu-Zr fuel to >18% burnup

    International Nuclear Information System (INIS)

    Pahl, R.G.; Wisner, R.S.; Billone, M.C.; Hofman, G.L.

    1990-01-01

    Tests of austenitic stainless steel clad U-xP-10Zr fuel (x=o, 8, 19 wt. %) to peak burnups as high as 18.4 at. % have been completed in the EBR-II. Fuel swelling and fractional fission gas release are slowly increasing functions of burnup beyond 2 at. % burnup. Increasing plutonium content in the fuel reduces swelling and decreases the amount of fission gas which diffuses from fuel to plenum. LIFE-METAL code modelling of cladding strains is consistent with creep by fission gas loading and irradiation-induced swelling mechanisms. Fuel/cladding chemical interaction involves the ingress of rare-earth fission products. Constituent redistribution in the fuel had not limited steady-state performance. Cladding breach behavior at closure welds, in the gas plenum, and in the fuel column region have been benign events. 3 refs., 5 figs

  10. Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

    Science.gov (United States)

    Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

    2018-04-01

    A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

  11. Predicted irradiation behavior of U3O8-Al dispersion fuels for production reactor applications

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Rest, J.

    1990-01-01

    Candidate fuels for the new heavy-water production reactor include uranium/aluminum alloy and U 3 O 8 -Al dispersion fuels. The U 3 O 8 -Al dispersion fuel would make possible higher uranium loadings and would facilitate uranium recycle. Research efforts on U 3 O 8 -Al fuel include in-pile irradiation studies and development of analytical tools to characterize the behavior of dispersion fuels at high-burnup. In this paper the irradiation performance of U 3 O 8 -Al is assessed using the mechanistic Dispersion Analysis Research Tool (DART) code. Predictions of fuel swelling and alteration of thermal conductivity are presented and compared with experimental data. Calculational results indicate good agreement with available data where the effects of as-fabricated porosity and U 3 O 8 -Al oxygen exchange reactions are shown to exert a controlling influence on irradiation behavior. The DART code is judged to be a useful tool for assessing U 3 O 8 -Al performance over a wide range of irradiation conditions

  12. 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.

  13. Post-irradiation examination of fuel elements of Tarapur Atomic Power Station (Report-I)

    International Nuclear Information System (INIS)

    Bahl, J.K.; Sah, D.N.; Chatterjee, S.; Sivaramkrishnan, K.S.

    1979-01-01

    Detailed post-irradiation examination of three initial load fuel elements of the Tarapur Atomic Power Station (TAPS) has been carried out. The causes of the element failures have been analysed. It was observed that almost 90% of the length of the elements exoerienced nodular corrosion. It has been estimated that nodular corrosion would seriously affect the wall thickness and surface temperature of higher rated elements. Lunar shaped fret marks have also been observed at some spacer grid locations in the elements. The depth of the largest fret mark was measured to be 16.9% clad wall thickness. Detailed metallographic examination of the clad and fuel in the three elements has been done. The temperatures at different structural regions of the fuel cross-sections have been estimated. The change in fuel density during irradiation has been evaluated by comparing the irradiated fuel diameter with the mean pellet design diameter. The performance of the end plug welds and spacer grid sites in the elements has been assessed. The burnup distribution along the length of the elements has been evaluated by gamma scanning. The redistribution of fission products in the fuel has been examined by gamma scanning and beta-gamma autoradiography. Mechanical properties of the irradiated cladding have been examined by ring tensile testing. (auth.)

  14. Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

    2018-04-01

    A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with