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Sample records for cadarache fuel element testing reactor

  1. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

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

  2. NEUTRONIC REACTOR FUEL ELEMENT

    Science.gov (United States)

    Picklesimer, M.L.; Thurber, W.C.

    1961-01-01

    A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

  3. Calculation of fuel element temperature TRIGA 2000 reactor in sipping test tubes using CFD

    International Nuclear Information System (INIS)

    It has been calculated the fuel element temperature in the sipping test of Bandung TRIGA 2000 reactor. The calculation needs to be done to ascertain that the fuel element temperatures are below or at the limit of the allowable temperature fuel elements during reactor operation. ensuring that the implementation of the test by using this device, the temperature is still within safety limits. The calculation is done by making a model sipping test tubes containing a fuel element surrounded by 9 fuel elements. according to the position sipping test tubes in the reactor core. by using Gambit. Dimensional model adapted to the dimensions of the tube and the fuel element in the reactor core of Bandung TRIGA 2000 reactor. Sipping test Operation for each fuel element performed for 30 minutes at 300 kW power. Calculations were performed using CFD software and as input adjusted parameters of TRIGA 2000 reactor. Simulations carried out on the operation of the 30, 60, 90, 120, 150, 180 and 210 minutes. The calculation result shows that the temperature of the fuel in tubes sipping test of 236.06 °C, while the temperature of the wall is 87.58 °C. The maximum temperature in the fuel center of TRIGA 2000 reactor in normal operation is 650 °C. and the boiling is not allowed in the reactor. So it can be concluded that the operation of the sipping test device are is very safe because the fuel center temperature is below the temperature limits the allowable fuel under normal operating conditions as well as the fuel element wall temperature is below the boiling temperature of water. (author)

  4. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  5. Power pulse tests on CANDU type fuel elements in TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensor for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (author)

  6. Recent developments in post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting held in Cadarache, France, 17-21 October 1994

    International Nuclear Information System (INIS)

    At the invitation of the Government of France, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) the IAEA convened a Technical Committee meeting from 14 to 21 October 1994 in Cadarache to discuss recent technical advances and improvements in the field of post-irradiation examination (PIE) of fuel used in nuclear power plants. Fifty participants representing 14 countries attended the meeting and 30 papers were presented and discussed during five technical sessions. Working Groups composed of the session chairmen and authors of papers prepared summaries of each session including conclusions and recommendations for future work. Refs, figs and tabs

  7. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  8. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition as part of a fuel meat thickness optimization effort for reactor performance other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  9. Load following tests on CANDU-type fuel elements in TRIGA research reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Two load following (LF) tests on CANDU-type fuel elements were performed in the TRIGA Research Reactor of INR Pitesti, where the tests were designed to represent fuel in a CANDU reactor operating in a load following regime. In the first LF test the designated '78R' fuel element successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, developed under INR-AECL co-operation, the fuel element designated as 'ME01' withstood 200 power cycles from 27 to 54 kW/m average linear power, as well as additional ramps due to reactor trips and restarts during the test period. This experimental program is ongoing at INR Pitesti. Both LF tests were simulated with finite element computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the cladding arising from expansion and contraction of the pellets. New LF tests are planned to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in the INR TRIGA Research Reactor compared with the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (author)

  10. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    The grid-shaped spacer for PWR fuel elements consists of flat, upright metal bars at right angles to the fuel rods. In one corner of a grid mesh it has a spring with two end parts for the fuel rod. The cut-outs for the end parts start from an end edge of the metal bar parallel to the fuel rods. The transverse metal bar is one of four outer metal bars. Both end parts of the spring have an extension parallel to this outer metal arm, which grips a grid mesh adjacent to this grid mesh at the side in one corner of the spacer and forms an end part of a spring for the fuel rod there on the inside of the outer metal bar. (HP)

  11. RIA and LOCA simulating tests on experimental fuel elements in TRIGA MT reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Full text: One of the main objectives of Institute for Nuclear Research (INR), Pitesti R and D Program is to investigate thermal and mechanical behaviour of fuel elements, thresholds and mechanisms of cladding failure during RIA and LOCA tests. Dual core TRIGA Material Testing Reactor of INR Pitesti (TRIGA SS MTR and TRIGA ACPR) is utilized extensively for studies of fuel behaviour under normal and postulated accident condition. A total of 39 test fuel elements have been irradiated in the TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti under RIA conditions. The ACPR tests program is still in progress and new experiments are foreseen to be performed in the following period. The test fuel elements are instrumented with CrAl thermocouples for cladding surface temperature measurement and every test fuel element has a pressure sensor for the internal pressure measurement. An experimental database of fuel behaviour parameters including fission - gas release, sheath strain, power - burnup history, etc. has been obtained using in-pile measurements and PIE results of test fuel elements irradiated in the TRIGA Steady State Material Testing Reactor (TRIGA SS MTR) of INR Pitesti. More than 100 test fuel elements have been irradiated in TRIGA SS MTR in different power history conditions. LOCA simulating tests are planned to be performed in C2 LOCA tests capsule and in Loop A of TRIGA SS MTR of INR Pitesti. The LOCA tests in capsule C2 are instrumented to measure fuel, sheath and coolant temperature, internal element and coolant pressure during the entire irradiation period. In the second phase of the experiment the C2 capsule will be connected to the sweep gas system with the on-line gamma ray spectrometer included. RIA type tests are planned in C6 capsule of TRIGA ACPR on test fuel elements with pre-hydrided claddings in order to investigate the influence of the precipitated hydride on fuel element cladding failure at high burnups in RIA conditions. This paper

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

    International Nuclear Information System (INIS)

    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 UO2 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.)

  13. Hazards review: N-Reactor 1.25% co-producer fuel element test

    Energy Technology Data Exchange (ETDEWEB)

    Miller, N.R.; Nechodom, W.S.

    1964-07-13

    The N-Reactor Hazard Summary Report examines the hazard from operating the N-Reactor with a uniform fuel loading enriched to 0.947% U{sup 235}. Incentives have been developed for reactor testing of a block of 49 tubes loaded with co-producer elements, i.e. elements capable of producing both weapons grade plutonium and tritium. The element utilizes an outer fuel tube enriched to 1.25% U{sup 235} with an inner target lithium-aluminum rod. Criteria have been developed to guide the evaluation of safety aspects of such tests. It is the purpose of this document to review the hazards associated with the proposed test and to set forth special precautions which will be necessary to maintain a high level of safety.

  14. Thermal simulations and tests in the development of a helmet transport spent fuel elements Research Reactor

    International Nuclear Information System (INIS)

    A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half-scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled. (author)

  15. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pope, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morrell, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jamison, R. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nef, E. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nigg, D. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  16. Performance evaluation of two CANDU fuel elements tested in the TRIGA reactor

    International Nuclear Information System (INIS)

    Nuclear Research Institute at Pitesti has a set of facilities, which allow the testing, manipulation and examination of nuclear fuel and structure materials irradiated in CANDU reactors from Cernavoda NPP. These facilities consist of TRIGA materials testing reactor and Post-Irradiation Examination Laboratory (LEPI). The purpose of this work is to describe the post-irradiation examination, of two experimental CANDU fuel elements (EC1 and EC2). The fuel elements were mounted into a pattern port, one in extension of the other in a measuring test for the central temperature evolution. The results of post-irradiation examination are obtained from: Visual inspection and photography of the outer appearance of sheath; Profilometry (diameter, bending, ovalization) and length measuring; Determination of axial and radial distribution of the fission products activity by gamma scanning; Measurement of pressure, volume and isotopic composition of fission gas; Microstructural characterization by metallographic and ceramographic analyzes; Isotopic composition and burn-up determination. The post-irradiation examination results are used, on one hand, to confirm the security, reliability and performance of the irradiated fuel, and on the other hand, for further development of CANDU fuel. (authors)

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

    International Nuclear Information System (INIS)

    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 ∼ 650 deg C. During irradiation the linear heat rating of the rods was in the range 400-535 W/cm. The diameter of the UO2 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 (∼ 750 deg C) and one long time irradiation (∼ 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

  18. Non-Destructive Testing of Reactor-Fuel, Target, and Control Elements

    International Nuclear Information System (INIS)

    At the Savannah River Plant (a production facility of the United States Atomic Energy Commission), fuel, target, and control elements are non-destructively tested before and after irradiation in reactors. Design and performance of unique instruments - used for measuring physical soundness, nuclear properties, and dimensions - are described. A nickel thickness gauge, utilizing the Hall effect in a magnetic field, is used to measure the thickness of nickel layers on uncanned uranium cores. A similar instrument is used after cores have been diffusion bonded to aluminium cladding to determine that each core has a layer of residual nickel, and that the end cap is sufficiently thick. Ultrasonic instruments are used (1) to measure uranium grain size to determine whether the cores were properly heat-treated, and (2) to detect unbonded areas between cladding and core. The Nuclear Test Gauge (NTG), a small subcritical assembly of U235-A1 alloy slugs in an H2O-moderated lattice, is used to determine the fuel (U235) or absorber (Li6) content of reactor elements. These determinations, made from changes in neutron multiplication, have a 1-sigma precision of about ± 0.5% for fuel elements containing up to 250 g U235/ 30.5 cm (1 ft), and about ± 1% for target and control elements containing up to 4 g Li6/30.5 cm (1 ft). Compared to the more commonly used large.critical test pile, the NTG costs about 1/20 as much; measures fuel or absorber content in about one minute vs. ten minutes; and measures the axial distribution of fuel or absorber which the test pile cannot do. Irradiated fuel elements are measured under water with (1) differential transformers that can measure diameter and length to an accuracy of ± 0.05 cm (0.002 in), and (2) simple mechanical linkages with dial indicators above water that can measure inside diameter and warp. By keeping the elements submerged in water, personnel are shielded from radiation, and the elements do not undergo the dimensional changes that

  19. Nondestructive examination of TRIGA reactor fuel elements

    International Nuclear Information System (INIS)

    Neutron radiography has proved to be a very useful method for nondestructive examination of used and nonused reactor elements. The method can be used for determination of homogenity and burn-up of fuel and burnable poisons, for detection of fuel and full clad damage and taking into account the capability to perform accurate geometrical measurements it is also possible to assess mechanical deformations of fuel elements. Active fuel elements of TRIGA reactor have been examined for deformations and fuel clad damage. In the course of these investigations the following methods were tested and compared: - transfer neutronradiographic techniques using In and Dy converter screens, - direct neutrongraphic method using solid state track detectors, - X-ray radiography employing lead shielding masks and highly selective photographic material. Considerable information on the burn-up of reactor fuel elements can be obtained from measuring the distribution of radioactive isotopes in the fuel element by gamma ray spectroscopy. For a used TRIGA fuel element the axial distribution of the isotope Cs-137 has been measured and the burn-up determined. We compare the experimental results with a crude estimate of burn-up

  20. Fabrication of MOX Fuel elements for irradiation in Fast Breeder Test Reactor (FBTR)

    International Nuclear Information System (INIS)

    Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur is fabricating Uranium - Plutonium Mixed Oxide Fuel (MOX) for different types of reactors. Recently MOX fuel pins for an experimental fuel subassembly of 37 pins has been fabricated for irradiation in Fast Breeder Test Reactor (FBTR) at Kalpakkam near Chennai. MOX fuel pins containing 44% PuO2 have also been also made for the hybrid core of FBTR. The experimental sub-assembly for irradiation testing in FBTR consisted of 37 short length Prototype Fast Breeder Reactor (PFBR) MOX fuel elements. The composition of the fuel was (0.71 U - 0.29 Pu) O2 with U233 O2 content of 53.5% of total UO2. Uranium enriched with U233 was used to simulate the heat flux of PFBR in FBTR neutron spectrum. MOX fuel pellets were made by powder metallurgy process consisting of pre-compaction, granulation, final compaction and sintering at high temperature. Initially U3233 O8 / U233 O3 powder was subjected to heat treatment. The pellets were sintered at reducing atmosphere at 1650oC for 4 hours to obtain acceptable quality pellets. Over sized pellets were centrelessly ground.without using a liquid coolant. During the fabrication of pins for experimental subassembly, technology was developed and conditions were optimized for making annular pellets, TIG welding of D9 tubes with SS 316 end plugs and wire wrapping. Quality control procedures and process control procedures at different stages of fabrication were developed. The hybrid core of FBTR consists of Mixed Carbide (MC) sub-assemblies containing (0.70 Pu - 0.30 U) C pellets and MOX fuel sub-assemblies containing (0.44 Pu - 0.56 U) O2. Studies were made to fabricate fuel containing higher percentage of Plutonium and the conditions were established. This paper describes the development of flowsheet for making annular MOX fuel pellets containing plutonium and U233, the technology for welding of D-9 clad tubes, wire wrapping and inspection. The paper also

  1. Nuclear reactor fuel element splitter

    International Nuclear Information System (INIS)

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

  2. Fuel element for nuclear reactor

    International Nuclear Information System (INIS)

    In order to avoid a can box or an adjacent fuel element sitting on the spacer of a fuel element in the corner during assembly, the top and bottom edges of the outer bars of the spacers are provided with deflector bars, which have projections projecting beyond the outside of the outer bars. (orig.)

  3. Material test reactor fuel research at the BR2 reactor

    International Nuclear Information System (INIS)

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  4. Reproduction of the RA reactor fuel element fabrication

    International Nuclear Information System (INIS)

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

  5. Nuclear reactor fuel element. Kernreaktorbrennelement

    Energy Technology Data Exchange (ETDEWEB)

    Lippert, H.J.

    1985-03-28

    The fuel element box for a BWR is situated with a corner bolt on the inside in one corner of its top on the top side of the top plate. This corner bolt is screwed down with a bolt with a corner part which is provided with leaf springs outside on two sides, where the bolt has a smaller diameter and an expansion shank. The bolt is held captive to the bolt head on the top and the holder on the bottom of the corner part. The holder is a locknut. If the expansion forces are too great, the bolt can only break at the expansion shank.

  6. Improved fuel element for fast breeder reactor

    International Nuclear Information System (INIS)

    The invention, in which the United States Department of Energy has participated as co-inventor, relates to breeder reactor fuel elements, and specifically to such elements incorporating 'getters', hereafter designated as fission product traps. The main object of the invention is the construction of a fast breeder reactor fuel pin, free from local stresses induced in the cladding by reactions with cesium. According to the invention, the fast breeder fuel element includes a cladding tube, sealed at both ends by a plug, and containing a fissile stack and a fertile stack, characterized by the interposition of a cesium trap between the fissile and fertile stacks. The trap is effective at reactor operating temperatures in retaining and separating the cesium generated in the fissile material and preventing cesium reaction with the fertile stack. Depending on the construction method adopted, the trap may consists of a low density titanium oxide or niobium oxide pellet

  7. Application of gamma scanning and neutron radiography methods to control fuel element state as tested in MR reactor loop channels

    International Nuclear Information System (INIS)

    The gamma-scanning and neutron radiography methods are described used for non-destructive control of fuel elements after their test in the loop channels of MR reactor, and also the equipment utilized. The techniques and the results obtained while studying fuel elements using the above mentioned methods are provided. It is established that gamma-scanning method can only indicate the presence of defect in the continuity of the fuel element core without identifying its type whereas the advantage of neutron radiography method is in obtaining visual results. At the same time gamma-scanning method makes it possible to determine energy release on the length of fuel elements, to find the burn up fraction, to study the phenomenon of fusion products migration which is difficult or impossible with neutron radiography method. A conclusion is drawn that gamma-scanning and neutron radiography methods successfully supplement each other and make it possible to obtain important information on fuel state in the irradiated fuel elements

  8. Structural analysis of reactor fuel elements

    International Nuclear Information System (INIS)

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design

  9. Structural analysis of reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Weeks, R.W.

    1977-01-01

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design.

  10. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

  11. Method for inspecting nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A technique for disassembling a nuclear reactor fuel element without destroying the individual fuel pins and other structural components from which the element is assembled is described. A traveling bridge and trolley span a water-filled spent fuel storage pool and support a strongback. The strongback is under water and provides a working surface on which the spent fuel element is placed for inspection and for the manipulation that is associated with disassembly and assembly. To remove, in a non-destructive manner, the grids that hold the fuel pins in the proper relative positions within the element, bars are inserted through apertures in the grids with the aid of special tools. These bars are rotated to flex the adjacent grid walls and, in this way relax the physical engagement between protruding portions of the grid walls and the associated fuel pins. With the grid structure so flexed to relax the physical grip on the individual fuel pins, these pins can be withdrawn for inspection or replacement as necessary without imposing a need to destroy fuel element components

  12. Compact Fuel Element Environment Test

    Science.gov (United States)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

    2012-01-01

    Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

  13. Evaluation of endcap welds in thin walled fuel elements of pressurised heavy water reactor by ultrasonic testing

    International Nuclear Information System (INIS)

    In the pressurised heavy water reactor systems of India, the fuel is encapsulated in thin-walled tubes (0.342 mm) closed with endcaps by resistance welding. The integrity of these fuel elements should be such that no fission gas leakage takes place during reactor operation. The quality control of the endcap welds needed to satisfy this requirement includes helium leak test and destructive metallographic test (on sample basis). This paper discusses the feasibility study that has been carried out in the author's laboratory to develop an immersion ultrasonic test method for evaluating the integrity of the endcap weld region. Through holes of various sizes (0.15mm, 0.2mm, 0.4mm diameter and 0.185mm and 0.342mm deep) were machined by spark erosion machining at the weld joints to simulate defects of various sizes. Line focussed probe of 10 MHz frequency was used for the testing. It was possible to detect clearly all the machined holes. Based on the above standardised procedure, further testing was done on endcap welds which were rejected during fabrication on account of showing leak rate of 3 x 10-6 std. c.c/sec. or more during helium leak test. Though it was possible to get echoes from the natural defects in the rejected tubes with echo amplitude of 70%, the signal was accompanied by the geometrical reflection (noise) giving an amplitude of 20% from the weld region, giving rise to the problem of resolving the defect indication from the geometric indications. Therefore, signal analysis approach was adopted. The signal obtained from the weld zone were subjected to various analysis procedures like a) autopower spectrum, b) total energy content and c) demodulated auto correlation function. It was possible by all the three methods to differentiate the defect signal from those due to weld geometry or due to noise. Subsequently, metallography was carried out to characterise the type of defects observed during the ultrasonic testing. (author). 4 figs

  14. Fuel element situation and performance data TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Electronic data acquisition of the position and movement of Triga fuel elements (FE) in the TRIGA II Vienna reactor was the objective of this project. Using one month power data and the Fuel element position in core it is possible to calculate their burnup. Fuel element performance data during 1962 to 2003 are provided. (nevyjel)

  15. The use of U/sub 3/Si/sub 2/ dispersed in aluminum in plate-type fuel elements for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

    1987-10-01

    A high-density fuel based on U/sub 3/Si/sub 2/ dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U/sub 3/Si/sub 2/ fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U/sub 3/Si/sub 2/ particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U/sub 3/Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U/sub 3/Si/sub 2/-aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m/sup 3/ is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs.

  16. Fundamental aspects of nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D.R.

    1976-01-01

    The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

  17. Convective parameters in fuel elements for research nuclear reactors

    International Nuclear Information System (INIS)

    The study of a prototype for the simulation of fuel elements for research nuclear reactors by natural convection in water is presented in this paper. This project is carry out in the thermofluids laboratory of National Institute of Nuclear Research. The fuel prototype has already been test for natural convection in air, and the first results in water are presented in this work. In chapter I, a general description of Triga Mark III is made, paying special atention to fuel-moderator components. In chapter II and III an approach to convection subject in its global aspects is made, since the intention is to give a general idea of the events occuring around fuel elements in a nuclear reactor. In chapter II, where an emphasis on forced convection is made, some basic concepts for forced convection as well as for natural convection are included. The subject of flow through cylinders is annotated only as a comparative reference with natural convection in vertical cylinders, noting the difference between used correlations and the involved variables. In chapter III a compilation of correlation found in the bibliography about natural convection in vertical cylinders is presented, since its geometry is the more suitable in the analysis of a fuel rod. Finally, in chapter IV performed experiments in the test bench are detailed, and the results are presented in form of tables and graphs, showing the used equations for the calculations and the restrictions used in each case. For the analysis of the prototypes used in the test bench, a constant and uniform flow of heat in the whole length of the fuel rod is considered. At the end of this chapter, the work conclusions and a brief explanation of the results are presented (Author)

  18. Testing device for fuel element samples

    International Nuclear Information System (INIS)

    The device described is for testing samples for behavior at high temperature in heavy gamma radiation. The whole device is designed to be maintained in the high neutron flux of a nuclear reactor channel. It comprises two co-axial envelopes with cylindrical side walls and with convex truncated bottom and head walls, these truncated walls being maintained in pairs at a small distance and as constant as possible owing to the inner envelope being designed to accept the fuel element or other sample for testing and to be connected to an intake pipe and a return pipe for a sample environmental gas. The truncated head wall of the outer envelope is joined by a sealed thermal expansion bellows to the cylindrical wall of this same envelope. The restricted annular space between the inner envelope and the outer envelope with its bellows is designed to be coupled to an intake pipe and a return pipe for a variable thermal conductivity gas

  19. Failed MTR Fuel Element Detect in a Sipping Tests

    International Nuclear Information System (INIS)

    This work describes sipping tests performed on Material Testing Reactor (MTR) fuel elements of the IEA-R1 research reactor, in order to find out which one failed in the core during a routine operation. Radioactive iodine isotopes 131I and 133I, employed as failure monitors, were detected in samples corresponding to the failed fuel element. The specific activity of each sample, as well as the average leaking rate, were measured for 137Cs. The nuclear fuels U3O8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137Cs

  20. Study of Tower Reactor Fuel Elements Based on Sintered Uranium Dioxide

    International Nuclear Information System (INIS)

    The paper gives the results of loop tests on a large batch of experimental fuel elements based on sintered uranium dioxide. Generalized data on the operation of fuel elements used in the reactors of the icebreaker ''Lenin'' are also included. (author)

  1. Method to fabricate block fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    The fabrication of block fuel elements for gas-cooled high temperature reactors can be improved upon by adding 0.2 to 2 wt.% of a hydrocarbon compound to the lubricating mixture prior to pressing. Hexanol or octanol are named as substances. The dimensional accuracy of the block is thus improved. 2 examples illustrate the method. (RW)

  2. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  3. Thermionic Fuel Element performance: TFE Verification Program. Final test report

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The program objective is to demonstrate the technology readiness of a Thermionic Fuel Element (TFE) suitable for use as the basic element in a thermionic reactor with electric power output in the 0.5 to 5.0 MW(e) range, and a full power life of 7 years. A TFE was designed that met the reliability and lifetime requirements for a 2 MW(e) conceptual reactor design. Analysis showed that this TFE could be used over the range of 0.5 to 5 megawatts. This was used as the basis for designing components for test and evaluation. The demonstration of a 7-year component lifetime capability was through the combined use of analytical models and accelerated, confirmatory tests in a fast test reactor. Iterative testing was performed in which the results of one test series led to evolutionary improvements in the next test specimens. The TFE components underwent screening and initial development testing in ex-reactor tests. Several design and materials options were considered for each component. As screening tests permitted, down selection occurred to very specific designs and materials. In parallel with ex-reactor testing, and fast reactor component testing, components were integrated into a TFE and tested in the TRIGA test reactor at GA. Realtime testing of partial length TFEs was used to test support, alignment and interconnective TFE components, and to verify TFE performance in-reactor with integral cesium reservoirs. Realtime testing was also used to verify the relation between TFE performance and fueled emitter swelling, to test the durability of intercell insulation, to check temperature distributions, and to verify the adequacy over time of the fission gas venting channels. Predictions of TFE lifetime rested primarily on the accelerated component testing results, as correlated and extended to realtime by the use of analytical models.

  4. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: nadiasam@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work consists in simulating the burnup of two different plate type fuel elements, where one is the benchmark MTR of the IAEA, which is made of an alloy of uranium and aluminum, while the other belonging to a typical multipurpose reactor is composed of an alloy of uranium and silicon. The simulation is performed using the WIMSD-5B computer code, which makes use of deterministic methods for solving neutron transport. In developing this task, fuel element equivalent cells were calculated representing each of the reactors to obtain the initial concentrations of each isotope constituent element of the fuel cell and the thicknesses corresponding to each region of the cell, since this information is part of the input data. The compared values of the k∞ showed a similar behavior for the case of the MTR calculated with the WIMSD-5B and EPRI-CELL codes. Relating the graphs of the concentrations in the burnup of both reactors, there are aspects very similar to each isotope selected. The application WIMSD-5B code to calculate isotopic concentrations and burnup of the fuel element, proved to be satisfactory for the fulfillment of the objective of this work. (author)

  5. Method of locating a leaking fuel element in a fast breeder power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Honekamp, John R. (Downers Grove, IL); Fryer, Richard M. (Idaho Falls, ID)

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  6. Method of locating a leaking fuel element in a fast breeder power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Honekamp, J.R.; Fryer, R.M.

    1978-03-21

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of /sup 134/Xe to /sup 133/Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  7. Fuel irradiation test plan at the Japan materials testing reactor

    International Nuclear Information System (INIS)

    Development of high performance fuels, which enables burnup extension and high duty uses of light water reactors (LWRs) by means of power up rates and flexible operating cycles, is one of key technical issues for extending the uses for longer periods. Introduction of new design fuel rods with new cladding alloys and wider utilization of mixed oxide fuels is expected in Japan. Fuel irradiation tests for development and safety demonstration are quite important, in order to realize theses progress. Operational management on water chemistry, minimizing the long term degradation of reactor components, could have unfavorable influence on the integrity of the fuel rods. Japanese government and the Japan Atomic Energy Agency have decided to re new the Japan Materials Testing Reactor (JMTR) and to install new test rigs, in order to play an active role solving the issues on the development and the safety of the fuel and the plant aging. Fuel integrity under abnormal transient conditions will be investigated using a special capsule type test rig, which has its own power control system under simulated LWR cooling conditions. Water loops for simulation of high duty operation, e.g. high power, high burnup and high rod internal pressure conditions, are proposed for the development and safety examination of the high performance fuels. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor and loss of coolant accident tests in hot laboratories would provide a comprehensive data for safety evaluation and design progress of the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients

  8. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, William Jonathan [Idaho National Laboratory; Barrett, Kristine Eloise [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  9. Pellet-clad interaction observations in boiling water reactor fuel elements

    International Nuclear Information System (INIS)

    Under a programme to assess the performance of fuel elements of Tarapur Atomic Power Station, post-irradiation examination has been carried out on 18 fuel elements in the first phase. Pellet-clad mechanical interaction behaviour in 14 elements with varying burnup and irradiation history has been studied using eddy current testing technique. The data has been analysed to evaluate the role of pellet-clad mechanical interaction in PCI/SCC failure in power reactor operating conditions. (author)

  10. Dry store for spent fuel elements from nuclear reactors

    International Nuclear Information System (INIS)

    In the dry store for spent fuel elements from nuclear reactors which are enclosed in storage tubes and cooled with air, the storage tubes being arranged in shafts of a storage building, a loading device is provided underneath the shafts and in a cooling air shaft designed for transporting. The loading device therefore requires only a small lifting height and the chances of storage tubes falling from great heights are excluded. This invention is applicable in particular for intermediate stores. (orig./RW)

  11. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    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

  13. Nondestructive testing in fabrication of zirconium alloy tubes and PHWR fuel elements in India

    International Nuclear Information System (INIS)

    The methods and technical means for nondestructive testing, applied at the Nuclear Fuel Complex (Hyderabad, India) in the process of fabricating channel, colander and shell tubes from zirconium alloy and fuel elements with the UO2 fuel for reactor cores of the PHWR-Candu power reactors are described in the review. The significant works on improving the methodology and equipment for ultrasonic quality control of the contact joint welding of fuel elements are noted

  14. Fission product release from defected nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    The release of gaseous (krypton and xenon) and iodine radioactive fission products from defective fuel elements is described with a semi-empirical model. The model assumes precursor-corrected 'Booth diffusional release' in the UO2 and subsequent holdup in the fuel-to-sheath gap. Transport in the gap is separately modelled with a phenomenological rate constant (assuming release from the gap is a first order rate process), and a diffusivity constant (assuming transport in the gap is dominated by a diffusional process). Measured release data from possessing various states of defection are use in this analysis. One element (irradiated in an earlier experiment by MacDonald) was defected with a small drilled hole. A second element was machined with 23 slits while a third element (fabricated with a porous end plug) displayed through-wall sheath hydriding. Comparison of measured release data with calculated values from the model yields estimates of empirical diffusion coefficients for the radioactive species in the UO2 (1.56 x 10-10 to 7.30 x 10-9 s-1), as well as escape rate constants (7.85 x 10-7 to 3.44 x 10-5 s-1) and diffusion coefficients (3.39 x 10-5 to 4.88 x 10-2 cm2/s) for these in the fuel-to-sheath gap. Analyses also enable identification of the various rate-controlling processes operative in each element. For the noble gas and iodine species, the rate-determining process in the multi-slit element is 'Booth diffusion'; however, for the hydrided element an additional delay results from diffusional transport in the fuel-to-heath gap. Furthermore, the iodine species exhibit an additional holdup in the drilled element because of significant trapping on the fuel and/or sheath surfaces. Using experimental release data and applying the theoretical results of this work, a systematic procedure is proposed to characterize fuel failures in commercial power reactors (i.e., the number of fuel failures and average leak size)

  15. A combined wet/dry sipping cell for TRIGA fuel element tests

    International Nuclear Information System (INIS)

    A combined wet/dry sipping cell for the investigation of research reactor fuel elements was developed and tested. It is capable of detecting temperature-dependent cladding failures through the release of gaseous fission products. Several TRIGA fuel elements were tested both in the wet in the dry sipping mode. Some elements released fission gases only above 75deg C. (orig.)

  16. ITER at Cadarache; ITER a Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-06-15

    This public information document presents the ITER project (International Thermonuclear Experimental Reactor), the definition of the fusion, the international cooperation and the advantages of the project. It presents also the site of Cadarache, an appropriate scientifical and economical environment. The last part of the documentation recalls the historical aspect of the project and the today mobilization of all partners. (A.L.B.)

  17. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    International Nuclear Information System (INIS)

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code

  18. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  19. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Science.gov (United States)

    Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-01

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  20. Inspection of state of spent fuel elements stored in RA reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Bulkin, S.Yu.; Sokolov, A.V. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Matausek, M.V.; Vukadin, Z. [VINCA Institute of Nuclear Science, Belgrade (Yugoslavia)

    1999-07-01

    About five thousand spent fuel elements from RA reactor have been stored for over 30 years in sealed aluminum barrels in the spent fuel storage pool. This way of storage does not provide complete information about the state of spent fuel elements or the medium inside the barrels, like pressure or radioactivity. The technology has recently been developed and the equipment has been manufactured to inspect the state of the spent fuel and to reduce eventual internal pressure inside the aluminum barrels. Based on the results of this inspection, a procedure will be proposed for transferring spent fuel to a more reliable storage facility. (author)

  1. Fuel element reshuffling and fuel follower control rods (FFCR) replacement for PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    The PUSPATI TRIGA Reactor has been utilized for more than 25 years using the same fuel elements and control rods. Generally, there are four control rods being used to control the neutron production inside the reactor core. A maintenance program has been developed to ensure its integrity, capability and safety of the reactor and it has been maintained twice a year since the first operation in 1982. The activities involve during the maintenance period including fuel elements and control rods inspections, electronics and mechanical systems, and others related works. During the maintenance in August 2008, there are some irregularities found on the fuel follower control rods and needed to be replaced. Even though the irregularities was not contributed into any unwanted incident, it were decided to replace with new control rods to avoid any potential hazards and unsafe condition occurred during operation later. Replacing any of the control rods would involved in imbalance of neutron flux and power distribution inside the core. Therefore, a number of fuel elements need to be reshuffled in order to compensate the neutron flux and power distribution as well as to balance the fuel elements burn-up in the core. This paper will described the fuel elements reshuffling and fuel follower control rods (FFCR) replacement for PUSPATI TRIGA Reactor. (Author)

  2. Development, Testing and Validation of a Waste Assay System for the Measurement and Characterisation of Active Spent Fuel Element Debris From UK Magnox Reactors - 12533

    Energy Technology Data Exchange (ETDEWEB)

    Mason, John A.; Burke, Kevin J.; Looman, Marc R.; Towner, Antony C.N. [ANTECH, A. N. Technology Ltd., Unit 6, Thames Park, Wallingford, Oxfordshire, OX10 9TA (United Kingdom); Phillips, Martin E. [Nympsfield Nuclear Ltd, Chapel House, The Cross, Nympsfield, Stonehouse GL10 3TU (United Kingdom)

    2012-07-01

    This paper describes the development, testing and validation of a waste measurement instrument for characterising active remote handled radioactive waste arising from the operation of Magnox reactors in the United Kingdom. Following operation in UK Magnox gas cooled reactors and a subsequent period of cooling, parts of the magnesium-aluminium alloy cladding were removed from spent fuel and the uranium fuel rods with the remaining cladding were removed to Sellafield for treatment. The resultant Magnox based spent fuel element debris (FED), which constitutes active intermediate level waste (ILW) has been stored in concrete vaults at the reactor sites. As part of the decommissioning of the FED vaults the FED must be removed, measured and characterised and placed in intermediate storage containers. The present system was developed for use at the Trawsfynydd nuclear power station (NPS), which is in the decommissioning phase, but the approach is potentially applicable to FED characterisation at all of the Magnox reactors. The measurement system consists of a heavily shielded and collimated high purity Germanium (HPGe) detector with electromechanical cooling and a high count-rate preamplifier and digital multichannel pulse height analyser. The HPGe based detector system is controlled by a software code, which stores the measurement result and allows a comprehensive analysis of the measured FED data. Fuel element debris is removed from the vault and placed on a tray to a uniform depth of typically 10 cm for measurement. The tray is positioned approximately 1.2 meters above the detector which views the FED through a tungsten collimator with an inverted pyramid shape. At other Magnox sites the positions may be reversed with the shielded and collimated HPGe detector located above the tray on which the FED is measured. A comprehensive Monte Carlo modelling and analysis of the measurement process has been performed in order to optimise the measurement geometry and eliminate

  3. Stuck fuel element experience at the Oregon State TRIGA reactor

    International Nuclear Information System (INIS)

    A stuck fuel element was found in June 1975 during the annual fuel element measuring assignment. When an attempt was made to remove the fuel element from position D-6, it was found the element would start to bind after being withdrawn about 10'', and it would not pass through the upper grid plate. A plan was devised to extract the stuck fuel element without having to remove the upper grid plate. An inhouse inquiry is in process to determine the reasons for the fuel element deformation. When the element cools sufficiently, we plan to obtain neutron radiographs that may help determine the answer. (author)

  4. Destructive Examination of Experimental Candu Fuel Elements Irradiated in TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW(th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (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) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Microstructural characterization by metallographic analyses; (iii) Determination of mechanical properties; (iv) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  5. Post Irradiation Examination of Experomental CANDU Fuel Elements Irradiated in TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW (th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (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) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Gamma scanning and tomography; (iii) Measurement of pressure, volume and isotopic composition of fission gas; (iv) Microstructural characterization by metallographic analyses; (v) Determination of mechanical properties; amd (vi) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  6. Non-destructive-Testing of Nuclear Fuel Element by Means of Neutron Imaging Technique

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Nuclear fuel element is the key component of nuclear reactor. People have to make strictly testing of the element to make sure the reactor operating safely. Neutron imaging is one of Non-destructive-Testing (NDT) techniques, which are very important techniques for

  7. Non destructive testing of irradiated fuel assemblies at the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Fuel performance and nuclear fuel qualification require a post-irradiation analysis. Non-destructive methods are utilised both in irradiated fuel storage pools and in hot-cells laboratories. As Brazil does not have hot-cells facilities for post-irradiation analysis, a qualification program for the Material Testing Reactor (MTR) fuel elements made at IPEN/CNEN-SP was adopted, based on non-destructive tests. The IPEN Fuel Engineering Group - CENC developed basic facilities for fuels post-irradiated analysis inside the reactor pool, which gives indications of: general state, by visual inspection; the integrity of the irradiated fuel cladding, by sipping tests; thickness measurements of the fuel miniplates during the irradiation time, for swelling evaluation; and, local burn-up evaluation by gamma spectrometry along the active area of the fuel element. This work describes that facilities, equipment and examples of some irradiated fuels analysis performed. (author)

  8. Use of silicide fuel in the Ford Nuclear Reactor - to lengthen fuel element lifetimes

    Energy Technology Data Exchange (ETDEWEB)

    Bretscher, M.M.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Burn, R.R.; Lee, J.C. [Univ. of Michigan, Ann Arbor, MI (United States). Phoenix Memorial Lab.

    1995-12-31

    Based on economic considerations, it has been proposed to increase the lifetime of LEU fuel elements in the Ford Nuclear Reactor by raising the {sup 235}U plate loading from 9.3 grams in aluminide (UAl{sub x}) fuel to 12.5 grams in silicide (U{sub 3}Si{sub 2}) fuel. For a representative core configuration, preliminary neutronic depletion and steady state thermal hydraulic calculations have been performed to investigate core characteristics during the transition from an all-aluminide to an all-silicide core. This paper discusses motivations for this fuel element upgrade, results from the calculations, and conclusions.

  9. Dissolution of low burnup Fast Flux Test reactor fuel

    International Nuclear Information System (INIS)

    The first Fast-Flux Test Facility reactor fuel [mixed (U,Pu)O2 composition] has been used in dissolution tests for fuel reprocessing. The fuel tested here had a peak burnup of 0.22 at. %, with peak centerline temperatures of 19970C. Linear dissolution rates of 0.99 to 1.57 mm/h were determined for dissolver solution and fresh acid, respectively. Insoluble residues from dissolution at 950C ranged from 0.18 to 0.28% of the original fuel. From 2 to 37 wt % of the residue was recoverable plutonium. Dissolution at 290C yielded residues of 0.56 to 0.64% of the original fuel. The major elements present in the HF leached residue included Ru, Mo, and Rh. The recovered cladding from the 950C dissolution contained the equivalent of 198 mg of 239Pu per 100 g of hulls, while the cladding from the 290c experiments contained only 0.21 mg of 239Pu per 100 g of hulls. 9 references, 5 figures

  10. Application of a quality control program for developing fuel elements for research reactors

    International Nuclear Information System (INIS)

    The development of nuclear fuel elements for the IEAR-1 research reactor is a task that is being pursued by IPEN/CNEN-SP for several years. The studies included the development of U3O8-Al Nuclear cermets, rolling of U3O8-Al brickets using the picture frame technique for the obtension of Nuclear fuel plates as well as the fabrication of components and the final assembling of the fuel elements. The prototypes are made to conform to stringent quality control specifications. These specifications cover various aspects such as the metallurgical, ceramical, and mechanical properties of the materials involved as well as non-destructive tests and dimensional and visual requirements of the various components. In this context, an extensive specification of the materials and components used have been compiled and are periodically reviewed and revised. An extensive quality control program was planned and is being tested in practice simultaneously to the fuel element development. During the elaboration of the procedures for the characterization tests, special attention has been devoted to the storage of data that could be used for the analysis of the irradiation behaviour of the fuel element. The various procedures used during implementation of the system required for the quality control of the nuclear fuel elements for the IEAR-1 Nuclear reactor are presented. (Author)

  11. Accident simulations and post irradiation examinations on spherical fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    An important aspect of the safety of high temperature reactors is the quality of the nuclear fuel and its ability to remain intact even at high temperatures and to safely contain the radioactive fission products. In combination with a suitable reactor an inherent safety against large release of fission products can be achieved. In this work experimental simulations of severe accidents were conducted on spherical fuel elements for high temperature reactors with TRISO-coated particles and fission product release was measured. The fuel elements originated from various irradiation experiments conducted at high temperatures with high burn-up. The experiments were performed using the cold finger apparatus, a test apparatus which was already used in the past in a former version at the Research Center Juelich. The new cold finger apparatus is installed since 2005 in the Hot Cells of the European Institute for Transuranium Elements. The cold finger apparatus at the Institute for Transuranium enabled incident simulations on irradiated high temperature reactor fuel elements in a helium atmosphere at ambient pressure, at temperatures up to 1800 C and for periods of several hundred hours. Here, both the release of fission gases and the release of solid fission products were measured. In addition, in the context of the present study, the mechanical behavior of the fuel particles and the transport mechanisms of the main fission products were analyzed and the expected release was computed. For a better understanding of the processes post irradiation examinations were conducted on the available fuel elements. It was finally made an assessment of the test results which were compared with results in the existing literature. A key objective of the work was the extension of the existing data base for modern HTR-fuel towards higher burn-up and higher fluences of fast neutrons, higher operating temperatures and extended accident temperatures.

  12. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  13. IFPE/EFE-RO, Experimental Fuel Elements RO89 and RO51 in TRIGA 14 MW Reactor (INR-Pitesti)

    International Nuclear Information System (INIS)

    Description of program or function: Romanian irradiation tests concerned with Candu type fuel elements behavior and with the limits of the design parameters. A particular feature of the Candu fuel project is the small plenum (void volume) added for relaxation of the fission gases, which are inherently released during the fuel irradiation. Two irradiation tests in the C2 device from the TRIGA 14 MW reactor were performed between the years 1985-1987. The tests were done to evaluate the effect of the fuel density on the time-evolution of the fission gas pressure. Experimental fuel elements were adequately instrumented with pressure transducers to follow the fission gas pressure changes during fuel irradiation. The first irradiation test was conducted on the fuel element coded No.89 whose main characteristics were the nominal values of the main fuel design parameters. The second one was conducted on the fuel element coded No.51. Because of the axial flux asymmetry inside the TRIGA reactor core, the experimental elements are shorter in length than the Candu fuel design. The irradiation tests consisted in evaluation of the time-evolution of the internal pressure from two experimental fuel elements having the main design characteristics as the Romanian Candu type fuel element design and to follow the dependence of the internal pressure of the fission gas on the fuel density

  14. Fuels for research and test reactors, status review: July 1982

    Energy Technology Data Exchange (ETDEWEB)

    Stahl, D.

    1982-12-01

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO/sub 2/ rod fuels. Among new fuels, those given major emphasis include H/sub 3/Si-Al dispersion and UO/sub 2/ caramel plate fuels.

  15. Fuels for research and test reactors, status review: July 1982

    International Nuclear Information System (INIS)

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO2 rod fuels. Among new fuels, those given major emphasis include H3Si-Al dispersion and UO2 caramel plate fuels

  16. In-Research Reactor Tests for SCWR Fuel Verifications

    International Nuclear Information System (INIS)

    The Supercritical water cooled reactors (SCWRs) are essentially light water reactors (LWRs) operating at higher pressure and temperature. The SCWRs achieve high thermal efficiency (i.e., about 45% vs. about 35% efficiency for advanced LWRs) and are simpler plants as the need for many of the traditional LWR components is eliminated. The SCWRs build upon two proven technologies, the LWR and the supercritical coal-fired boiler. The main mission of the SCWR is production of low-cost electricity. Thus the SCWR is also suited for hydrogen generation with electrolysis, and can support the development of the hydrogen economy in the near term. In this paper, the SCWR fuel performance verification tests are reviewed. Based on this review results, in-research reactor verification tests to be performed in a fuel test loop through the international joint program are proposed. In addition, capsule tests and fuel test loop tests to be performed in HANARO are also proposed

  17. Welding of sule elements for nuclear reactors with solid state YAG laser using instrumentated testing equipments

    International Nuclear Information System (INIS)

    The instrumentation of the equipment for carrying out safety tests on fuel elements for nuclear reactors requires special thermocouples adapted to the prevailing agressive medium. The investigations described deal essentially with the operational and metallurgical weldability tests out on the safety test zircaloy piping in the pressurized water circuit (PHEBUS-programme)

  18. Testing and implementation program for the modified Darlington 37-element fuel bundle

    International Nuclear Information System (INIS)

    To mitigate the effects of reactor ageing, a design modification to the 37-element fuel is proposed in which the diameter of the centre element will be reduced to 11.5 mm from 13.1 mm. The testing and implementation phase for the 37-element fuel bundle modification is discussed in this paper. The initial plan for testing is to perform a set of out-reactor tests to assess the endurance, acoustic response and cross-flow behaviour of the revised fuel bundle design. The initial schedule outlines activities that will enable OPG to implement full core fuelling of the modified bundle within the next three to four years. (author)

  19. Fabrication of spherical fuel element for 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Cold quasi-isostatic molding with a silicon rubber die was used for manufacturing the spherical fuel elements of 10 MW high temperature gas-cooled reactor. 44 batches of fuel elements, about 20540 of the fuel elements, were produced. The cold properties of the graphite matrix materials satisfies the design specifications. The mean free uranium fraction in spherical fuel element from 44 batches is 4.57 x 10-5, certified products is 99%

  20. Neutron spectrum and radial power distribution measurements in a TRIGA reactor fuel element

    International Nuclear Information System (INIS)

    The neutron spectrum in the Illinois Advanced TRIGA Reactor was measured by a crystal spectrometer utilizing an LiF(1, 1, 1) crystal monochromator whose reflectivity was determined experimentally. The fission heat source distribution in a fuel element was also determined as a function of the fuel element temperature. These two measurements were used to investigate the effects of fuel element temperature and the local core loading on the thermal diffusion length in a fuel element. Changes in the thermal diffusion lengths during a reactor pulse underlie the proposed temperature feedback mechanism for the ZrH fuel material. The results of the measurements confirm, in part, this proposed temperature feedback mechanism

  1. Economical analysis to utilize MTR fuel elements using silicides in research reactors

    International Nuclear Information System (INIS)

    According to international programs on reducing enrichment in research reactors and the necessity to maintain their operation, new fuel elements have been developed in order to meet both objectives. Thus, U-Si alloy fuel elements for research reactors are becoming of greater interest for the international markets. It became necessary to make an economic study about the convenience of introducing this type of fuel elements in the RA-3 reactor and to know the potentiality of this fuel. The economical behavior of the reactor operation has been evaluated comparing the actual U3O8 nuclear fuel cycle with U3Si2 nuclear fuels. Results obtained show that the main economical factor to determine the change of fuels is the cost of fabrication, and the change is advisable up to an 80% difference. The other factors related to the cost of nuclear fuel cycle are not relevant or have real minor impacts. (author)

  2. Design of reactor internals in larger high-temperature reactors with spherical fuel elements

    International Nuclear Information System (INIS)

    In his paper, the author analyzes and summarizes the present state of the art with emphasis on the prototype reactor THTR 300 MWe, because in addition to spherical fuel elements, this type includes other features of future HTR design such as the same flow direction of cooland gas through the core. The paper on hand also elaborates design guidelines for reactor internals applicable with large HTR's of up to 1200 MWe. Proved designs will be altered so as to meet the special requirements of larger cores with spherical elements to be reloaded according to the OTTO principle. This paper is furthermore designed as a starting point for selective and swift development of reactor internals for large HTR's to be refuelled according to the OTTO principle. (orig./GL)

  3. Vibration tests on some models of PEC reactor core elements

    International Nuclear Information System (INIS)

    This paper describes the aims of the experimental tests carried out at ISMES, within an agreement with the Department of Fast Reactors of ENEA, on some models of the elements of PEC Fast Nuclear Reactor Core in the frame of the activities for the seismic verification of the PEC core. The seismic verification is briefly described with particular attention to the problems arising from the shocks among the various elements during an earthquake, as well as the computer code used, the purpose and the techniques used to perform tests, some results and the first comparison between the theory and the experimental data

  4. Specifications for high flux isotope reactor fuel elements HFIR-FE-3

    International Nuclear Information System (INIS)

    This specification covers requirements for two types of aluminum-base fuel elements which together will be used as the fuel assembly in the High Flux Isotope Reactor (HFIR). Requirements are included for materials of construction, fabrication, assembly, inspection, and quality control to produce fuel elements in accordance with Company drawings

  5. Eddy current examination of the nuclear fuel elements of IPR-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Roger F.; Frade, Rangel T.; Oliveira, Paulo F.; Silva, Marlucio A.; Silva Junior, Silverio F., E-mail: rfs@cdtn.br, E-mail: rtf@cdtn.br, E-mail: pfo@cdtn.br, E-mail: mas@cdtn.br, E-mail: silvasf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    Tubes of AISI 304 stainless steel as well as tubes of Aluminum 1100-F are used as cladding of the fuel elements of TRIGA MARK 1 nuclear research reactor. Usually, these tubes are periodically inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements in which the cladding has failed, but it is not able to determine the place where the discontinuity is located. In turn, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In this paper, a study about the use of eddy current testing for detection and characterization of discontinuities in the fuel elements cladding is proposed. The study involves the development of probes able to operate in underwater inspections, the design and manufacture of reference standards and the development of a test methodology to perform the evaluations. (author)

  6. Eddy current examination of the nuclear fuel elements of IPR-R1 research reactor

    International Nuclear Information System (INIS)

    Tubes of AISI 304 stainless steel as well as tubes of Aluminum 1100-F are used as cladding of the fuel elements of TRIGA MARK 1 nuclear research reactor. Usually, these tubes are periodically inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements in which the cladding has failed, but it is not able to determine the place where the discontinuity is located. In turn, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In this paper, a study about the use of eddy current testing for detection and characterization of discontinuities in the fuel elements cladding is proposed. The study involves the development of probes able to operate in underwater inspections, the design and manufacture of reference standards and the development of a test methodology to perform the evaluations. (author)

  7. IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

    2010-10-01

    Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

  8. Modelling of pressurized water reactor fuel, rod time dependent radial heat flow with boundary element method

    International Nuclear Information System (INIS)

    The basic principles of the boundary element method numerical treatment of the radial flow heat diffusion equation are presented. The algorithm copes the time dependent Dirichlet and Neumann boundary conditions, temperature dependent material properties and regions from different materials in thermal contact. It is verified on the several analytically obtained test cases. The developed method is used for the modelling of unsteady radial heat flow in pressurized water reactor fuel rod. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-01

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

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

    International Nuclear Information System (INIS)

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U3O8 powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated

  11. Irradiation testing of miniature fuel plates for the RERTR program. [Reduced Enrichment Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R.L.; Martin, M.M.

    1981-07-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates (miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. These tests screen various candidate fuel materials on their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% /sup 235/U in place of highly enriched fuel for these reactors would reduce the potential for /sup 235/U diversion. The irradiation test facility, designated as HFED, is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The miniplates will achieve burnups of up to approx. 2.2 x 10/sup 27/ fissions/m/sup 3/ of fuel.

  12. Thermal-hydraulic transient analysis of a packed particle bed reactor fuel element

    OpenAIRE

    Casey, William Emerson

    1990-01-01

    Title as it appears in the M.I.T. Graduate List, Jun. 4, 1990: Transient thermal-hydraulic analysis of a packed particle bed reactor fuel element A model which describes the thermal-hydraulic behavior of a packed particle bed reactor fuel element is developed and compared to a reference standard. The model represents a step toward a thermal-hydraulic module for a real-time, autonomous reactor powder controller. The general configuration of the fuel element is a bed of small (diameter about...

  13. Design criteria -- Modification of fuel element test facilities. 1706-KER Project CGI-839

    Energy Technology Data Exchange (ETDEWEB)

    Rudock, E.R.

    1959-08-27

    The following criteria outlines the basis, objectives, and fundamental methods that shall govern the preparation of final design for ``Project CGI-839, Modification to Fuel Element Test Facilities -- 1706 KER.`` These modifications will provide the equipment to test NPR size fuel elements in the KER recirculating loops. The 1706-KER Recirculation Test Facility of KE Reactor is operated to obtain experimental data regarding high temperature reactor coolant technology and high temperature fuel element testing. The facility consists of four in-pile recirculating loops. These loops will permit testing of fuel elements with the existing process tubes of 2.1 inches I.D. To provide adequate in-reactor fuel element test facilities to support the development of NPR fuel, two KER loops, {number_sign}3 and {number_sign}4 will be converted to provide a process tube of 2.7 inches ID that will be operated at typical NPR irradiation conditions. The remaining loops No. 1 and 2, will be modified to provide additional flow and heat transfer capacity for greater flexibility in the testing of high temperature fuel elements smaller than the NPR size. New pumps, heat exchangers, and minor piping modifications will be required in all loops.

  14. Fast neutron reactor fuel elements and power grid duty cycling

    International Nuclear Information System (INIS)

    The PHENIX power grid cycling operation in 1982-1983 will allow verification of the models and criteria developed in the interim. It will provide indispensible statistical data and will open the way to power grid duty for Super PHENIX beginning in 1986. Although at the present time it is impossible to resolve the question of weekly or daily load variations, it is felt that fast neutron reactor fuel subassemblies should provide satisfactory performance for primary and secondary frequency adjustments

  15. Bending of fuel fast reactor fuel elements under action of non-uniform temperature gradients and radiation-induced swelling

    International Nuclear Information System (INIS)

    The bending of rod fuel elements in gas-cooled fast reactors under the action of temperature gradients radiation-induced swelling non-uniform over the perimeter of fuel cans is evaluated. It is pointed out that the radiation-induced swelling gives the main contribution to the bending of fuel elements. Calculated data on the bending of the corner fuel element in the assembly of the fast reactor with dissociating gas coolant are given. With the growth of temperature difference over the perimeter, the bending moment and deformation increase, resulting in the increase of axial stresses. The obtained data give the basis for accounting the stresses connected with thermal and radiation bending when estimating serviceability of fuel elements in gas cooled fast reactors. Fuel element bending must be also taken into account when estimating the thermal hydrualic properties

  16. Fuel element design handbook

    Energy Technology Data Exchange (ETDEWEB)

    Merckx, K.R.

    1958-09-01

    The economic development of nuclear reactors depends upon the integrated progress in the fields of reactor design, fuel element design, reactor operation, and fuel production and separation. Broad criteria, which restrict the fuel element design, are determined by the mutual consideration of the problems encountered in all the above fields. Hence, no stage of reactor design or operation is independent of the fuel element problem, nor can the fuel element designer disregard the interest of any one field. As an introduction to the fuel element design problem, this chapter describes how the general criteria for a fuel element are determined.

  17. Burn-Up Calculations for the Brookhaven Graphite Research Reactor Fuel Elements

    International Nuclear Information System (INIS)

    Fuel bum-up calculations for the Brookhaven Graphite Research Reactor involve a distribution of the thermal megawatt days of operations to the fuel elements in proportion to the average thermal neutron flux at their location in the reactor. The megawatt days so assigned can be converted to equivalent uranium-235 consumption when needed. The original fuel loading for the BGRR was neutral uranium and a single calculation was performed on each fuel element upon discharge from the reactor. A subsequent change to a fully enriched uranium-235 fuel element, however, introduced complications. The average loading of enriched uranium involves about 4800 individual elements, each occupying four different reactor positions during its term in the reactor. The total term for a central channel element is about one year as against six to eight years for an element in a peripheral channel. With the large number of individual fuel elements involved and the approximately monthly small changes needed for operation, it was necessary to resort to a computer programme to follow the burn-up of all the elements on the reactor continuously. Both this and other functions of the computer programme are discussed in the paper. To date, uranium has been recovered from two batches of spent fuel. On the first, involving 3674 elements discharged from the reactor over a period of 4.9 years, the recovery figures were 5.5% higher than the calculated total of 32.3 kg uranium-235. On the second batch, involving 1296 elements discharged from the reactor over a period of one year, the recovery figures were 2.3% higher than the calculated figures of 10.8 kg uranium-235. This relatively close agreement seems to indicate that the assumptions made to simplify the programme are acceptable and that the results of the programme are satisfactory for our particular accounting and operating requirements. (author)

  18. The fuel element situation at the TRIGA mark II reactor Vienna

    International Nuclear Information System (INIS)

    The fuel history, spent fuel storage situation and recent problems covering the period from 1962 until 1.6.2001 were reviewed. After almost 40 years of TRIGA MARK II reactor Vienna operation, it must be mentioned that the experience with TRIGA fuel elements was and is excellent. During this period only 9 fuel elements had to be permanently be removed from the core and 57 fuel elements from the initial start-up are still used in the core. A careful fuel management and a frequent fuel inspection is of most importance, fuel elements should be moved at least two-times a year from their core position to check free movement and a 180 deg. rotation of the fuel element is also recommended (nevyjel)

  19. Proceedings of the specialist meeting on the safety of water reactors fuel elements

    International Nuclear Information System (INIS)

    This specialist meeting on the safety of water reactors fuel elements was held in Saclay (France) in October 1973, and was organized by CSNI and CEA. It attracted specialists from 14 countries. Session I was devoted to normal operating conditions (coolant-cladding and fuel-cladding interactions, fission product release, effects of cladding deformation on fuel element performances and reactor operating limits); Session II was devoted to operating reactor accidents and failures, anomalous transients and handling accidents; Session III was devoted to modifications to be applied to fuel elements in order to enhance their safety and reliability; Session IV was devoted to Loss-of-Coolant Accidents (LOCA)(cladding behaviour during the accident, assembly behaviour during the accident, criteria to be considered for the study of fuel element behaviour during a LOCA)

  20. FUEL ASSAY REACTOR

    Science.gov (United States)

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  1. Out-of-pile modelling of nuclear fuel elements for MTR type reactors. Pt. 1

    Energy Technology Data Exchange (ETDEWEB)

    Farhadi, Kazem [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Radiations Application Research School

    2014-03-15

    In the first part of the present paper, for a 5 MW thermal pool-type research reactor, the fuel element is modelled for when undergoing both natural circulation of the coolant and forced convection of the coolant operational conditions. First, the required dimensionless groups were identified and then the pertinent similarity criteria were derived accordingly. The derived similitude laws were modified under the conditions of identical pressure, identical temperature difference and identical coolant and fuel cladding in the model and the prototype. These modifications were done for the system under both natural and forced convections. The effect of varying cladding materials under normal operating conditions of the research reactor were observed via coolant channel thickness. Also the effect of a wider coolant channel on the nature of the coolant fluid was observed. The results obtained indicate that it is not possible to conserve all the dimensionless groups between the model and the prototype and hence achieve an errorless outcome. Among all the liquids available, methanol is the only liquid which nearly satisfies the thermal-hydraulic similitude and must be used in place of ordinary coolant water. This in turn necessitates the coolant channel to be wider and as a consequence the traditional Aluminium cladding in research reactors should be replaced by Iron. The derived scale down criteria can be used for the design of fuel element for the out-of-pile testing. (orig.)

  2. Design and Testing of Prototypic Elements Containing Monolithic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    N.E. Woolstenhulme; M.K. Meyer; D.M. Wachs

    2011-10-01

    The US fuel development team has performed numerous irradiation tests on small to medium sized specimens containing low enriched uranium fuel designs. The team is now focused on qualification and demonstration of the uranium-molybdenum Base Monolithic Design and has entered the next generation of testing with the design and irradiation of prototypic elements which contain this fuel. The designs of fuel elements containing monolithic fuel, such as AFIP-7 (which is currently under irradiation) and RERTR-FE (which is currently under fabrication), are appropriate progressions relative to the technology life cycle. The culmination of this testing program will occur with the design, fabrication, and irradiation of demonstration products to include the base fuel demonstration and design demonstration experiments. Future plans show that design, fabrication, and testing activities will apply the rigor needed for a demonstration campaign.

  3. Block fuel element for gas-cooled high temperature reactors

    International Nuclear Information System (INIS)

    The invention concerns a block fuel element consisting of only one carbon matrix which is almost isotropic of high crystallinity into which the coated particles are incorporated by a pressing process. This block element is produced under isostatic pressure from graphite matrix powder and coated particles in a rubber die and is subsequently subjected to heat treatment. The main component of the graphite matrix powder consists of natural graphite powder to which artificial graphite powder and a small amount of a phenol resin binding agent are added

  4. Irradiation tests in TRIGA MT reactor of INR Pitesti related to safety of nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore; Olteanu, Gheorghe [Institute for Nuclear Research, INR, PO Box 78, 1 Campului Street, RO-115400 Mioveni, Jud. Arges (Romania); Makihara, Yoshiaki [International Atomic Energy Agency Wagramerstr. 5, A-1400 Vienna (Austria)

    2006-07-01

    The design of modern power reactors reflects the close attention paid to improve safety and reliability of nuclear fuel. With the evolution of fuel design and the possibilities for more stringent operational conditions it is of concern to determine if the present safety criteria are adequate as most of them were established 15 to 20 years ago most of the time on un-irradiated materials. One of the main objectives of Institute for Nuclear Research (INR), Pitesti R and D Program is to investigate thermal and mechanical behaviour of fuel elements, thresholds and mechanisms of cladding failure during RIA and LOCA tests. Dual core TRIGA Material Testing Reactor of INR Pitesti (TRIGA SS MTR and TRIGA ACPR) is utilized extensively for studies of fuel behaviour under normal and postulated accident conditions such as reactivity - initiated accident (RIA) and loss-of-coolant accident (LOCA). A total of 40 test fuel elements have been irradiated in the TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti under RIA conditions. The ACPR tests program is still in progress and new experiments are foreseen to be performed in the following period. The test fuel elements are instrumented with CrAl thermocouples for cladding surface temperature measurement and every test fuel element has a pressure sensor for the internal pressure measurement. New RIA type tests are planned in C6 capsule of TRIGA ACPR on test fuel elements with pre-hydrided claddings in order to investigate the influence of the precipitated hydride on fuel element cladding failure at high burnups in RIA conditions. An experimental database of fuel behaviour parameters concerning fission - gas release, sheath strain, power - burnup history, etc. has been obtained using in-pile measurements and PIE results of test fuel elements irradiated in the TRIGA Steady State Material Testing Reactor (TRIGA SS MTR) of INR Pitesti. More than 110 test fuel elements have been irradiated in TRIGA SS MTR in different power

  5. Japanese study on water reactor fuel element materials and method of measurement

    International Nuclear Information System (INIS)

    So many studies have been carried out in Japan on water reactor fuel element materials and the method of measuring their properties. Some topics will be high-lighted in this report to give an idea of what they have been doing in Japan. Studies on the properties of zircaloy, including the development work for modified alloy have been performed since the late 1950s, both in fundamental work at universities and research organizations and development work for zircaloy tube commercial production in metal industry. Among them, the latest work on the creep characteristics of zircaloy tubing is presented. For other material used in fuel element than cladding tube, spacer-spring will be discussed as the second topic. Reduction of spring force was carried out in Japan PWR fuel spacer to reduce rod bow in the early 1970s which could successfully eliminate the problem since 1977. Relaxation of spring force against burn-up has been discussed in the so-called reliability test program of both BWR and PWR on then-standard fuel since 1975 which was reported at Stockholm symposium in September 1986. The data obtained will be presented. As the study for the method of measurement, the author proposed a modified testing procedure for tensile and burst test for zircaloy tubing, at ASTM-B10 Committee in 1976, to emphasize the shape of mandrel in the tensile test which has a significant effect on elongation values. Various measurement techniques in post irradiation examination of water reactor fuel have been developed in Japan, among which the shadow measurement of tube during ballooning to burst will be described. (author)

  6. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  7. Application of neutron radiography for non-destructive testing nuclear fuel elements

    International Nuclear Information System (INIS)

    This paper describes the experimental procedures, testing information and application advantages when neutron radiography is used for non-destructive inspections and quantitative analysis of fuel elements from nuclear power plants. Both the 235U enrichment and the material distribution inside the pellets can be determined by neutron radiography methods for the non-irradiated fuel elements. Both the structural integrity of fuel elements for different reactors such as PWR, BWR, FBTR and the hydrogen accumulation in the cladding material can be inspected for the irradiated samples. (authors)

  8. Latest development on the disposal of Research Reactor Fuel and Triga Fuel elements

    International Nuclear Information System (INIS)

    The MTR spent nuclear fuel reprocessing by the UKAEA, the intermediate storage+direct disposal for research reactors, the research reactor spent nuclear fuel return to the U.S., shipments and ports of entry, management sites, fees, storage technologies, contracts, actual shipments, legal processes, and NUKEM activities are listed. (HSI)

  9. Non-destructive control of cladding thickness of fuel elements for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karlov, Y.; Zhukov, Y.; Chashchin, S

    1997-07-01

    The control method of fuel elements for research reactors by means of measuring beta particles back scattering made it possible to perform complete automatic non-destructive control of internal and external claddings at our plant. This control gives high guarantees of the fuel element correspondence to the requirements. The method can be used to control the three-layer items of different geometry, including plates. (author)

  10. Non-destructive control of cladding thickness of fuel elements for research reactors

    International Nuclear Information System (INIS)

    The control method of fuel elements for research reactors by means of measuring beta particles back scattering made it possible to perform complete automatic non-destructive control of internal and external claddings at our plant. This control gives high guarantees of the fuel element correspondence to the requirements. The method can be used to control the three-layer items of different geometry, including plates. (author)

  11. Performance of low-enriched U3Si2-aluminum dispersion fuel elements in the Oak Ridge Research Reactor

    International Nuclear Information System (INIS)

    Six high-density, low-enriched U3Si2-Al dispersion fuel elements have been tested in the Oak Ridge Research Reactor (ORR). The elements were geometrically identical to standard ORR elements. The uranium density in the fuel meat ranged between 4.6 and 5.2 Mg/m3. The elements were fabricated by B and W, CERCA, and NUKEM using their normal materials and fabrication practices, with minor modifications necessitated by the new fuel. The U3Si2 contained minor amounts of USi, U3Si and/or uranium solid solution. The elements were irradiated to approximately normal ORR burnup, and three elements were irradiated twice as long, to average burnups of ∼80% of the initially contained 235U, well above the burnups normally achieved in research and test reactors. Peak burnups of 98% were achieved. Following suitable cooling periods, the elements were subjected to a series of nondestructive and destructive examinations. The behavior of the fuel was found to be entirely consistent with the known irradiation behavior of the constituent phases. The extremely stable swelling behavior of the U3Si2 phase dominated in all cases. The plates showed small, uniform thickness changes. Blister threshold temperatures were ≥5500C. It is concluded that low-enriched U3Si2-Al dispersion fuel elements will perform at least as well in research and test reactors with power densities up to that of the ORR as the highly enriched UAl/sub x/-Al and U3O8-Al dispersion fuels currently being used. There were no indications that use of this fuel under substantially more stringent conditions might be precluded. 10 refs., 87 figs., 9 tabs

  12. Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection

    International Nuclear Information System (INIS)

    A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

  13. Papers concerning fuel particles, fuel elements and graphitic materials for high temperature reactors presented at the 1980 annual conference Kerntechnik

    International Nuclear Information System (INIS)

    This report is a compilation of the papers presented by staff of the Institute for Reactor Materials, KFA-Juelich, at the 1980 annual conference Rekatortechnik, held in Berlin, 25-27th March 1980. In some cases, there were co-authors from other organisations. Where possible the manuscripts of the presentations have been reproduced, as well as the display cards shown during the poster session on the conference. In the presentations, the questions of the characterization of fuel particles and the retention of fission products are dealt with, and special attention is given to fission product release at very high temperatures. One presentation deals with the disposal of fuel elements from the AVR. Another report presents the results of radiation experiments on the standard matrix material A3-3 of the THTR fuel elements; the changes in dimensions, creep coefficient and thermal conductivity were measured as functions of the fluence and the radiation temperature. Interim results obtained from long-term radiation experiments on reflector graphites for high and very high flux reactors are presented, and models for the calculation of dimensional changes in components subjected to fluctuating temperatures from data obtained in isothermal tests are discussed. (orig.)

  14. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  15. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  16. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  17. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    International Nuclear Information System (INIS)

    Separate abstracts have been prepared for each paper presented in the following areas of interest: (1) fuel development; (2) post-irradiation examinations; (3) reprocessing; (4) thermite reaction; (5) fuel fabrication; (6) element tests; (7) core tests; (8) criticals; (9) shipping; and (10) reactors and methods

  18. A study of parameters on marking of Prototype Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor Fuel (PFBR) elements are identified with a permanent unique marking. Identification of the fuel elements is very much necessary for traceability during initial fabrication as well as for post irradiation examination. Marking on fuel element has to be permanent and capable of being identified after irradiation. Laser marking is a relatively new method as compared to other marking technologies such as ink marking, mechanical engraving and electro chemical methods. It is used for the product identification and traceability during its service life. Laser marking has many advantages compared to other conventional marking. In laser marking process, mark quality is a very important factor, which depends on so many variables like input current, pulse frequency, marking speed and number of passes. The influence of the pulse frequency and the speed of travel of the laser beam on the mark depth and width have been studied in this paper. An optical microscope, scanning electron microscope were used to measure the effects of pulse frequency on the mark depth and width. It has been found that the mark depth and width depend on the interaction process of the laser beam and the material, which was influenced by the pulse frequency. Micro hardness testing is carried out to report Heat Affected Zone (HAZ) variation with parameters. Marking speed and input current selected for suitable depth and width were mentioned in the present study. (author)

  19. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  20. ENEA TRIGA RC-1 reactor spent fuel elements shipment to the USA

    International Nuclear Information System (INIS)

    TRIGA Mark II reactor of ENEA's Casaccia research Center (in Italy named RC-1) reached first criticality in 1960. In more than thirty years of operation, 1 MW reactor core has been modified many times for fuel elements burn-up optimization. Till now, because of achieved maximum burn-up, 146 fuel elements have been definitively removed from reactor core and transferred to the hot storages in reactor pool (5 racks around reactor vessel) and in the reactor room (pits). The activities planning, the organizing aspect study, the analysis and valuations both nuclear safety and radioprotection have been suitable for the TRIGA RC-1 fuel element shipment. Infact, no operative anomaly is appeared respect the approved procedures. Personnel engagement has been as expectations and the personnel absorbed gamma dose resulted negligible. Finally, the NAC disposable narrow time (only one week at the end of July) has not produced heavy organization problems but it has been a strong goad per all operative structures involved in the TRIGA RC-1 elements shipment

  1. Development of eddy current test system for fuel element based on LabVIEW

    International Nuclear Information System (INIS)

    Fuel element plays an important role in high temperature gas-cooled reactor-pebble-bed module. For adjusting the fuel element precisely, an eddy current test system based on LabVIEW was developed to count the plumbago ball precisely and select the defective balls. The system was composed of the hardware circuit, the computer, the data acquisition card and relative software. The design of the excitation source, head amplifier circuit and the phase-sensitive detector was introduced in detail. The plumbago balls were counted and defects were tested by this system , and the results showed that the system is with good test capability. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  3. Testing of reactor fuel materials using nuclear techniques

    International Nuclear Information System (INIS)

    The tests presented here apply to: the quantitative determination of uranium in the core of fuel element plates by the detection of the number of neutrons produced in photo induced reactions in uranium; the determination of 235U proportion in uranium dioxide samples, in the form of uranyl nitrate, by the technique of the detection of tracks produced by fission fragments and in pellet samples by passive gamma spectrometry and the checking of uranium homogenization distribution in fuel plates and uranium dioxide pellets. (Author)

  4. Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAlx-Al) containing 4.40 gU/cm3 of an MTR was replaced with silicide (U3Si-Al and U3Si2-Al) and oxide (U3O8-Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U3Si2-Al followed by 0.03% for U3Si-Al, and 0.01% for U3O8-Al fuel. The U3O8-Al fueled reactor gave the maximum ρexcess at BOL which was 21.67% more than the original fuel followed by U3Si-Al which was 2.55% more, while that of U3Si2-Al was 2.50% more than the original UAlx-Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U3O8-Al followed by U3Si-Al and then U3Si2-Al fuel.

  5. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  6. Research on Measuring Technology for In-pile Fuel Element Testing

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The tested fuel assembly for In-pile test for PWR fuel element with instrumentation consisted of 4instrumented fuel elements and total 12 sets of transducers. Double claddings are adopted to raise fueltemperature. Two fuel elements each have 2 thermocouples for measuring separately the fuel centerlinetemperature and the cladding surface temperature. The other two elements have membrane type oressure

  7. Reactivity measurements of the IPR-R1 TRIGA reactor fuel elements

    International Nuclear Information System (INIS)

    The thermal power of the IPR-R1 TRIGA reactor, belonging to the Centro de Desenvolvimento da Tecnologia Nuclear, will be upgraded from 100 k W to 250 k W. To attain this objective, mew additional fuel elements will be inserted in the reactor core. In order to provide information to the calculations of the new core arrangement, some fuel rods reactivity measurements were carried out as well as the determination of the reactivity increase due to the substitution of the present fuel by a new one. A first estimate indicates that the addition of 5 new fuel elements might be sufficient to reach the desired value of 3$ ρ excess. (author). 5 refs., 1 fig., 2 tabs

  8. Transients and safety testing of LMFBR fuel pins in the reactor BR2

    International Nuclear Information System (INIS)

    Testing of the behaviour of LMFBR fuel pins under operational transients has been performed in the reactor BR2 at S.C.K./C.E.N.-Mol (Belgium) since 1981 in the framework of the DEBENE programme ''SNR-Betriebstransienten-experimente''. A special purpose sodium loop, called ''VIC'', has therefore been developed to allow off-nominal and transient experiments on single fuel pins under realistic fast reactor operating conditions. Two basic types of tests can be run, either separately or simultaneously: fission power alteration, e.g. steady overpower runs, power cycling and fast transient overpower (TOP); mismatch of the sodium cooling, e.g. operation with reduced sodium flow and transient loss of flow (LOF). The loop allows the loading and testing of pre-irradiated fuel pins. In the field of safety oriented tests, the programme ''MOL 7 C'' investigates the LMFBR fuel element behaviour under locally blocked cooling conditions and the possible failure propagation. The work is jointly carried out by the Karlsruhe center KfK (FRG) and S.C.K./C.E.N.-Mol (Belgium). The related in-pile tests in the reactor BR2 have started in 1977 and are performed in a fully integrated sodium loop. The test section contains a 30-rod bundle with fresh or pre-irradiated fuel pins. A local porous blockage within the fuel bundle initiates severe local damage to the central rods. Important informations are obtained with respect to the problems of pin to pin propagation and the long term behaviour of a fuel subassembly with defect pins. The MOL 7 C loop system can also be used to run operational transients on a fuel bundle with representative fuel pins. The paper describes the irradiation devices VIC and MOL 7 C from their technological point of view and depicts their field of testing applications. Also the major experiments already performed and relevant irradiation data are reviewed

  9. Transactions of 2. international seminars on the mathematical/mechanical modelling of reactor fuel elements

    International Nuclear Information System (INIS)

    Fuel element modelling is a wide field of activity that spans decades of research and code development for different reactor systems and very different situations such as normal operation, off-normal situations and severe accidents. Modern computer technology helps to take the full advantage of detailed model development performed over the past for daily design analyses, safety analyses, conception of new experiments and investigation of an improved nuclear fuel utilization and fuel element performance. The basic development of the concepts of fuel element modelling can be considered as finished. The future trends are the development of refined models based on a deeper understanding of the physical and mechanical basis. Areas of interest are transient phenomena especially the fission product behaviour, burnup-enhanced phenomena, PCI and fuel reliability, severe core damage and chemical aspects. The seminar presentations reflect this variety

  10. Welding procedures used in the fabrication of fuel elements for the DON Reactor exponential experiment

    International Nuclear Information System (INIS)

    This exponential experiment required 74 units (37 loaded with UO2 and 37 with UC) to simulate the Reactor fuel channels. Each unit was enclosed in a tube similar to the calandria ones. It contained the pressure tube, the shroud and the 19 rods cluster. Within the pressure tube, in touch with the elements, was the organic liquid. (Author)

  11. Fabrication and Testing of Prototype APM-Clad UO2 Fuel Elements

    International Nuclear Information System (INIS)

    In support of the 50-MW(e) Prototype Organic Power Reactor Programme (POPR), extensive development work has been performed on aluminium powder metallurgy (ARM) products, toward their use as cladding for UO2 fuel. As part of this development work, eutectic bonding, flash butt welding, and cold-pressure welding were investigated as methods for making end closures in die fuel element cladding. Vibratory packing was studied as a means of filling APM tubes with UO2. Out-of-pile tests were conducted to obtain information on APM-UO2 compatibility. This work revealed that, under present conditions, eutectic bonding was the most suitable method for making end closures; vibratory packing produced fuel densities in the range of 80 to 88% of theoretical density; and no APM-UO2 reaction took place in the range of POPR operating temperatures (850oF maximum fuel-cladding interface temperature). As a result o f this development work, five APM-clad UO2 prototype fuel elements have been fabricated for testing in the Organic Moderated Reactor Experiment (OMRE). Each element consisted of 24 or 25 APM-clad fuel rods, arranged in a 5 x 5 array in a nickel-plated steel or an APM fuel box. To increase surface area, the extruded APM cladding had eight fins which were spiralled to a pitch of 45 or 90e/ ft to further improve heat transfer. The fuel rod end closures were made by eutectic bonding of silver-plated aluminium end plugs to the APM tubing. The elements were instrumented to: (1) Measure cladding surface and coolant temperatures, (2) Detect fuel rod failure, (3) Change coolant velocity (means of achieving peak cladding surface temperature of 850oF), (4) Measure coolant velocity, and (5) Measure fission gas build-up. These elements have been installed in the OMRE with target fuel burn-ups of 25000 to 30000 MWd/t of uranium. As of 1 April 1963, they had achieved accumulated burn-ups ranging from 7700 to 12 000 MWd/t of uranium. Two of the elements had been removed from the reactor as a

  12. Fuel and core testing plan for a target fueled isotope production reactor

    International Nuclear Information System (INIS)

    In recent years there has been an unstable supply of the critical diagnostic medical isotope 99Tc. Several concepts and designs have been proposed to produce 99Mo the parent nuclide of 99Tc, at a commercial scale sufficient to stabilize the world supply. This work lays out a testing and experiment plan for a proposed 2 MW open pool reactor fueled by Low Enriched Uranium (LEU) 99Mo targets. The experiments and tests necessary to support licensing of the reactor design are described and how these experiments and tests will help establish the safe operating envelop for a medical isotope production reactor is discussed. The experiments and tests will facilitate a focused and efficient licensing process in order to bring on line a needed production reactor dedicated to supplying medical isotopes. The Target Fuel Isotope Reactor (TFIR) design calls for an active core region that is approximately 40 cm in diameter and 40 cm in fuel height. It contains up to 150 cylindrical, 1-cm diameter, LEU oxide fuel pins clad with Zircaloy (zirconium alloy), in an annular hexagonal array on a ∼2.0 cm pitch surrounded, radially, by a graphite or a Be reflector. The reactor is similar to U.S. university reactors in power, hardware, and safety/control systems. Fuel/target pin fabrication is based on existing light water reactor fuel fabrication processes. However, as part of licensing process, experiments must be conducted to confirm analytical predictions of steady-state power and accident conditions. The experiment and test plan will be conducted in phases and will utilize existing facilities at the U.S. Department of Energy's Sandia National Laboratories. The first phase is to validate the predicted reactor core neutronics at delayed critical, zero power and very low power. This will be accomplished by using the Sandia Critical Experiment (CX) platform. A full scale TFIR core will be built in the CX and delayed critical measurements will be taken. For low power experiments, fuel

  13. Design and production process of bushing-type fuel elements for channel research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Afanasiev, V.L.; Aleksandrov, A.B.; Enin, A.A. [NZHK, Novosibirsk (Russian Federation)

    1998-07-01

    The design of bushing-type fuel elements (FEs) based on the dioxide fuel composition UO{sub 2}+Al for channel research reactors is described. Commercial technological process for bushing-type FEs with up to 0.8 g/cm{sup 3} uranium concentration in the fuel core is presented. This technology is based on fuel core production using powder metallurgy with subsequent chemical treatment of its surface and enclosing into the finished cladding. Commercial technological process for bushing-type FEs with 0.8-3.8 g/cm{sup 3} uranium concentration in the fuel composition is considered. This process is based on fuel core production by means of extrusion technology followed by fuel core enclosing into the cladding. (author)

  14. Identification of a leaking TRIGA fuel element at the reactor facility of Pavia

    International Nuclear Information System (INIS)

    On January 28th 2004, during a periodical activity of characterization of the ionic-exchange resins of the demineralizer of the primary cooling circuit of the TRIGA Mark II reactor of the University of Pavia a small but detectable amount of 137Cs contamination was measured. Since the reactor has been running for several hundreds of hours at full power without showing any anomaly in the radiometric and thermo-hydraulic parameters, the reactor was brought at the nominal power of 250 kW for one hour and a sample of water was collected from the reactor tank and analyzed in a low background gamma-ray detector. As a result a small amount of fission products were detected in the reactor pool water (few Bq/g) suggesting the existence of a possible clad defect in one ore more fuel elements. As a consequence of this situation a campaign of gamma-ray spectrometry was implemented in order to evaluate the importance of the release. Analyzes using a HGe detector (1.72 keV FWHM - 31.3 % efficiency - 58.5 Photo Peak/Compton) were performed and the most significant results are presented as well as the identification of the leaking fuel element. The fission products leakage was due to a micro-fissure of a fuel element that released only noble gas when it was heated up to a temperature around 90oC , i.e. at the reactor power of about 100 kW. The oldest SST clad instrumented fuel element in the core was identified as the origin of the release. It was removed from its position and stored in a rack of the reactor pool under 4 m of water shield. The reactor came back in regular operation on March 22nd 2004 and no other fission products leakages were detected. After this situation the reactor pool water is sampled and measured with a low-background gamma-ray detector every month before the reactor start-up and after one hour of operation of the reactor at full nominal power. (nevyjel)

  15. Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAlx-Al) containing 4.40 gU/cm3 of an MTR was replaced with silicide (U3Si-Al and U3Si2-Al) and oxide (U3O8-Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U3O8-Al was about 2% more than the original UAlx-Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

  16. The effects of neutron irradiation on the type 316 stainless steel for homemade fast reactor fuel element cladding material

    International Nuclear Information System (INIS)

    The irradiation experiments on the homemade 316 stainless steel of six kinds of chemical composition with different treatment technology used for fast reactor fuel element cladding material are introduced. The materials have been irradiated in the High Flux Engineering Test Reactor (HFETR) to a fluence of 3.1 x 1021 neutron/cm2 (>0.1 MeV) at 650 degree C and subsequently tension has been tested at the same irradiation temperature and room temperature. Microstructure of the some tensile specimens were examined. The experiment results are analyzed and assessed. (authors)

  17. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    Energy Technology Data Exchange (ETDEWEB)

    Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  18. MAW and HTR fuel element test disposal in boreholes

    International Nuclear Information System (INIS)

    The Kernforschungsanlage Juelich, KFA, (Nuclear Research Center Juelich) has been handling a project since 1983 on 'Further Development of the Borehole Technology for the Disposal of Radioactive Wastes in Salt, with the Examples of Dissolver Sludge, Fuel Element Claddings, Fuel Hardware und HTR Fuel Elements'. The project is sponsored by the Bundesminister fuer Forschung und Technologie, BMFT, (Federal Ministry of Research and Technology) under the identification number KWA 5302 3 and bears the short title 'MAW and HTR Fuel Element Test Disposal in Boreholes'. The major objective of the project is to develop a technique for the disposal of the above mentioned wastes in unlined boreholes in salt and to test this technique in the Asse salt mine. The Institut fuer Chemische Technologie der Nuklearen Entsorgung, ICT (Institute of Chemical Technology) at the KFA is responsible for the scientific and organizational management of the project. The Institut fuer Tieflagerung, IfT, (Institute for Underground Disposal) of the Gesellschaft fuer Strahlen- und Umweltforschung mbH, GSF, (Society for Radiological and Environmental Research) is responsible for the geomechanical and mining activities in the project. It supervises the in-situ experiments, and as the owner of the Asse salt mine, it submits applications for the experiments to the licensing authorities. Geomechanical calculations are being carried out by the Bundesanstalt fuer Geowissenschaften und Rohstoffe, BGR, (Federal Institute for Geological Sciences and Natural Resources). (orig./RB)

  19. Study of neutronic effects of water hole in the control fuel element in the Tehran research reactor

    International Nuclear Information System (INIS)

    Due to the existence of some fission products in the core outlet of the Tehran Research Reactor, fuel leakage tests were done on all fuel elements. These experiments showed that the fission products come out mostly from some of the fuel control elements. So the failures of fuel control elements were taken out of the core and reactor started with the new core configuration. Failure of fuel plate in pool type research reactors resulting in a release of fission products to the pool water fall into the following categories: 1. Blistering or swelling due to excessive fission gas or fission products build-up. 2. Failure due to corrosion (chemical-metallurgical, ...). 3. Manufacturing defect such as non-homogeneous fuel distribution. 4. Coolant channel blockage by foreign materials. From neutronic point of view, existance of water hole in the CFE causes increasing thermal flux on the fuel plate adjusted to the water hole. Therefore, there is a probability of blistering or swelling in some parts of the plates. For more studies of this effect, we did neutronic calculations using computer codes. Calculation also showed that the fission rate in the above mentioned plate in the CFE is more than other plates in the core. Therefore blistering or swelling due to excessive fission gas or fission products build-up in some part of the plate is probable. As it was mentioned before, the probability of chemical and metallurgical corrosion on the fuel clad should not be ignored. However, since the release of fission products from some of the CFE were more than other fuel assemblies, the probability of occuring cladding defects by the neutronic effects of the water hole is stronger than other effects. (Author)

  20. Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Russian Research Centre, Moscow (Russian Federation)

    1996-03-01

    Testing of 43 fuel elements (13 fuel elements with high burn-up fuel, 10 fuel elements with preirradiated cladding and fresh fuel, and 20 non-irradiated fuel elements) was carried out in the IGR pulse reactor with a half width of the reactor power pulse of about 0.7 sec. Tests were conducted in capsules with no coolant flow and with standard initial conditions in the capsule of 20{degrees}C and 0.2 MPa. Two types of coolant were used: water and air. One purpose of the test program was to determine the thresholds and mechanisms of fuel rod failure under RIA conditions for VVER fuel rods over their entire exposure range, from zero to high burn-up. These failure thresholds are often used in safety analyses. The tests and analyses were designed to reveal the influence on fuel rod failure of (1) the mechanical properties of the cladding, (2) the pellet-to-cladding gap, (3) fuel burn-up, (4) fuel-to-coolant heat transfer, and other parameters. The resulting data base can also be used for validation of computer codes used for analyzing fuel rod behavior. Three types of test specimens were used in the tests, and diagrams of these specimens are shown in Fig. 1. {open_quotes}Type-C{close_quotes} specimens were re-fabricated from commercial fuel rods of the VVER-1000 type that had been subjected to many power cycles of operation in the Novovoronezh Nuclear Power Plant (NV NPP). {open_quotes}Type-D{close_quotes} specimens were fabricated from the same commercial fuel rods used above, but the high burn-up oxide fuel was removed from the cladding and was replaced with fresh oxide fuel pellets. {open_quotes}Type-D{close_quotes} specimens thus provided a means of separating the effects of the cladding and the oxide fuel pellets and were used to examine cladding effects only.

  1. Determination of the minimum number of fuel elements of the RP-10 research reactor

    International Nuclear Information System (INIS)

    The peruvian research reactor RP-10 is composed of a compound nucleus of boxes containing fuel plates which are cooled with light water in order to remove heat produced by fission of uranium atoms. However from a certainty viewpoint, it exists certain restrictions to design the cooling system. The most admissible caloric flux of 90.3 watts/cm2 is deflux of 90.3 watts/cm2 is determined on the basis of these thermic restrictions when cooling speed is 409 cm/sec permitting at least 24 fuel elements(boxes) within the nucleus. On the basis of restrictions of load loss in the nucleus, it would be permitted at least 18 fuel elements, but this quantity breaks thermic restrictions for this reason, 24 boxes in the nucleus will be the minimum number of elements

  2. Proceedings of the international meeting on development, fabrication and application of reduced enrichment fuels for research and test reactors

    International Nuclear Information System (INIS)

    Separate abstracts were prepared for each of the papers presented in the following areas: (1) Reduced Enrichment Fuels for Research and Test Reactors (RERTR) Program Status; (2) Fuel Development; (3) Fuel Demonstrations; (4) General Topics; and (5) Specific Reactor Applications

  3. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

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

    2010-02-23

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

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

    Science.gov (United States)

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

    2013-09-03

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

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

    Science.gov (United States)

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

    2011-03-01

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

  6. The Status and Development Potential of Plate-Type Fuels for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stahl, D.

    1979-03-01

    Recent U.S. Department of State action to restrict the shipment and use of highly enriched uranium for research and test reactors has renewed fuel development activity. The objective of these development activities is to increase the total uranium loading in the fuel meat so that enrichment reduction can be accomplished without significant performance penalties. This report characterizes the status and the potential for development of the currently utilized plate-type fuels for research and test reactors. The report also characterizes the newer high-density fuels which could be utilized in these reactors and indicates the impact of the utilization of both the new and current fuels on enrichment reduction.

  7. An Expert System to Analyze Homogeneity in Fuel Element Plates for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tolosa, S.C.; Marajofsky, A.

    2004-10-06

    In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up. This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to x-ray images. These images are generated when the x-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized x-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate.

  8. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO2) mixed with urania (UO2). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified

  9. High-density reduced-enrichment fuels for Research and Test Reactors

    International Nuclear Information System (INIS)

    Development and irradiation testing of high-density fuels have been conducted by the US RERTR Program in order to provide the technical means to reduce the enrichment of fuels for research and test reactors. The traditional aluminum dispersion fuel technology has been extended to include the highest practical loadings of uranium-aluminide (UAl/sub x/, 2.3 MgU/m3), uranium-oxide (U3O8, 3.2 MgU/m3), and uranium-silicide (U3Si2, 5.5 MgU/m3; U3Si, 7.0 MgU/m3) fuels. A third uranium-silicide alloy, U3SiAl (U + 3.5 wt % Si + 1.5 wt % Al) has been found to perform poorly at high burnup. Testing of miniature fuel plates and full-sized fuel elements is at an advanced stage for the highest loadings of the aluminide and oxide fuels and intermediate loadings of the silicide fuels, and good results have been obtained for low-enriched uranium. The data obtained to date are discussed. 1 reference, 3 figures, 1 table

  10. Heat transfer burnout in tube-type fuel elements of nuclear power reactors

    International Nuclear Information System (INIS)

    The conditions are formulated under which the results of the experimental research of the boilino. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented. (F.M.)

  11. Melt-dilute treatment of spent nuclear fuel assemblies from research and test reactors

    International Nuclear Information System (INIS)

    The Savannah River Site is the U.S. Department of Energy's preferred site for return and treatment of all aluminum-base, spent, research and test reactor fuel assemblies. There are over 20,000 spent fuel assemblies now stored in different countries around the world, and by 2035 many will be returned to SRS for treatment and interim storage, in preparation for disposal in a geologic repository. The early fuel assemblies for research and test reactors were made using aluminum clad plates that were fabricated from highly enriched (93%) uranium-aluminum alloy. Later, powder metallurgical fabrication methods were developed to produce plate fuels with higher uranium contents using either uranium aluminide, uranium oxide or uranium silicide powders mixed with aluminum. Silicide fuel elements generally are fabricated with low enriched uranium containing less than 20% 2'35U. Following irradiation, the spent fuel assemblies are discharged from the reactor, and most assemblies have been stored in underwater pools, some since the early 1950's. A number of disposition options including direct/co-disposal and melt-dilute treatment were evaluated recently. The melt-dilute technique was identified as the preferred method for treatment of aluminum-base spent fuel. The technique consists of melting the spent fuel assembly and adding depleted uranium to the melt for isotopic dilution to 2'35U. Aluminum is added, if necessary, to produce a predetermined alloy composition. Additionally, neutron poisons may be added to the melt where they form solid solution phases or compounds with uranium and/or aluminum. Lowering the enrichment reduces both criticality and proliferation concerns for storage. Consolidation by melting also reduces the number of storage canisters. Laboratory and small-scale process demonstration using irradiated fuel is underway. Tests of the off gas absorption system have been initiated using both surrogate and irradiated RERTR mini fuel plates. An experimental L

  12. Process for extracting and preparing fuel elements from a reactor for transport and device for carrying out the process

    International Nuclear Information System (INIS)

    The fuel element is lifted from the reactor using the lifting device of a bridge crane, is taken over the opening of a transport container, and is lowered into this. During transfer the fuel element is situated in a shielding hood at the end of the crane chain. The transport container contains liquid salt or a slat mixture, which solidifies after one or several fuel elements are inserted in it. (DG)

  13. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  14. Integrity assessment of test fuel assemblies of the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Assessment of integrity has been made on the B-type fuel assemblies, which will be loaded in the High Temperature Engineering Test Reactor (HTTR) as test fuel assemblies. Specifications of coated fuel particles for the B-1 type fuel assembly have been slightly changed in the fuel kernel diameter and thickness of coating layers from those for the A-type fuel assembly, which is employed as the driver fuel. These changes have been directed toward safer side in developing this advanced fuel for use up to higher burnups at higher temperatures. The B-2 type fuel assembly uses the zirconium-carbide (ZrC) coating layer with excellent high-temperature chemical stability, instead of the silicon carbide (SiC) layer. This change has lead to demonstration of its better performance than the A-type fuel assembly in the kernel migration, corrosion by fission products including palladium, and coating failure at extremely high temperatures. The B-3 type fuel assembly adopts the (U,Th)O2 kernel - SiC TRISO coated fuel articles. The service condition (1000degC and 22,000 MWd/t) of the B-3 type fuel assembly is decided as the range within which the performance data of the fuel have been sufficiently obtained. Thus, it has been judged that the integrity of these B-type fuel assemblies will be maintained under the normal operating conditions of the HTTR. Moreover, the validity of the permissible design limit of the fuel has been confirmed, which requires that the fuel temperature shall not exceed 1,600degC at anticipated operational transients. (author)

  15. Water reactor fuel element computer modelling in steady state, transient and accident conditions

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency, upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT). This meeting was the fifth in the series of IAEA meetings on the topic of Water Reactor Fuel Element Modelling, previous meetings being held in 1978, 1980, 1982 and 1984. Sixty-seven participants from 21 countries attended the meeting, and 35 papers were presented and discussed. These numbers are almost exactly the same as for the 1984 meeting, which demonstrates a continuing interest in the topic. The papers were presented in five sessions under the following headings: Session I - General Modelling (6 papers); Session II - Thermo-Mechanical Modelling and PCI (7 papers); Session III - Fission Gas Release (7 papers); Session IV - Transient Behaviour (8 papers); Session V - Axial Gas Transport and Thermal Modelling (7 papers). A separate abstract was prepared for each of these 35 papers. Refs, figs and tabs

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

    International Nuclear Information System (INIS)

    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 UO2 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.)

  17. Test reactor technology

    International Nuclear Information System (INIS)

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

  18. Identification of a leaking TRIGA fuel element at the nuclear reactor facility of the University of Pavia

    International Nuclear Information System (INIS)

    During a periodical activity of characterization of the ionic-exchange resins of the demineralizer of the primary cooling circuit of the TRIGA Mark-2 reactor of the University of Pavia a small but detectable amount of Cs137 contamination was measured. Since the reactor has been running for several hundreds of hours at full power without showing any anomaly in the radiometric and thermo-hydraulic parameters, the reactor was brought to the nominal power of 250 kW for one hour and a sample of water was collected from the reactor tank and analysed in a low-background gamma ray detector. As a result a small amount of fission products were detected in the reactor pool water (a few Bq/g) suggesting the existence of a possible clad defect in one or more fuel elements. Since no halogens such as iodine and bromine were detected in the sampled water, the more likely hypothesis, also supported by literature, seemed to be a micro-fissure in the neck of an instrumented fuel element. A dedicated apparatus for reactor pool water sampling and on-line spectroscopy measurements was realized. As expected, the fission product leakage was due to a micro-fissure of a fuel element that released noble gas only when the fuel element was heated up to a temperature around 90 Celsius degrees. The leaking fuel element was identified and removed from its position and the reactor was back in regular operation after 2 months from leakage detection. (authors)

  19. Development of Non-Metallic Fuel Elements for a High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    In connection with fuel element development work for the high-temperature gas-coolcd reactor of the Brown-Boveri/Krupp Reaktorbau G.m.b.H., two different fuel element concepts were considered and developed. In both cases the fuel element consists of a graphite ball of 6 cm in diam. which contains the fuel insert, a cylindrical pellet of about 20 mm in diam. and 16 mm in height. The two concepts differ in the type of the.fuel insert as well as in the preparation of the graphite ball. In the first concept the fuel insert consists of a mixture of UC2 and graphite which is prepared by blending U3O8 and graphite, pressing them into pellets and reacting the two components in a vacuum furnace at 1800oC. The atomic ratio of U : C is 1:45. Since this type of fuel pellet does not retain the fission products completely the surrounding graphite sphere had to be made impervious to fission products by impregnation in order to obtain a fission-product retaining element. Permeabilities of the order of 10-6cm2/s could be achieved. In the second concept the fuel insert consists of a solid solution of UC in ZrC and is coated with a layer of ZrC. The molar ratio of UC to ZrC is 1 : 20. The fuel pellet preparation was accomplished by the following procedure: UO2, ZrO2, and graphite were mixed and pressed into pellets. The pellets were reacted to the carbides. Ball milling of the carbides was followed by hot pressing at temperatures o f 2000oC. Densities of more than 95% of the theoretical density could be achieved. A full description of the preparation and of some physical properties of the fuel pellets is given in the paper. A sufficient fission gas retention behaviour of this type of fuel insert which allows it to be put into unimpregnated graphite balls is expected. Other advantages of this kind of fuel are discussed. (author)

  20. AREVA Back-End Possibilities for the Used Fuel of Research Test Reactors

    International Nuclear Information System (INIS)

    One of the major issues faced by the Research and Test Reactor (RTR) operators is the back end management of the used fuel elements. RTR used fuel for both HEU and LEU types are problematic for storing and disposal as their Aluminium cladding degrades leading to activity release, possible loss of containment and criticality concerns. Thus, direct disposal of RTR used fuel, (without prior treatment and conditioning) is in this respect hardly suitable. In the same manner, long term interim storage of RTR used fuel has to take into account the issue of fuel corrosion. Treating RTR used fuel allows separating the content into recyclable materials and residues. It offers many advantages as compared to direct disposal such as the retrieval of valuable fissile material, the reduction of radio-toxicity and a very significant reduction of the volume of the ultimate waste package (reduction factor between 30 and 50). In addition, the vitrification of the residues provides a package that has been specifically designed to ensure long term durability for long term interim storage as well as final disposal (99% of the activity is encapsulated into a stable matrix). RTR fuel treatment process was developed several decades ago by AREVA with now thirty years of experience at an industrial level. The treatment process consists in dissolving the whole assembly (including the Al cladding) in nitric acid and then diluting it with standard Uranium Oxide fuel dissolution liquor prior to treatment with the nominal Tributylphosphate solvent extraction process. A wide range of RTR spent fuel has already been treated in the AREVA facilities. First, at the Marcoule plant over 18 tons of U-Al type RTR fuel from 21 reactors in 11 countries was processed. The treatment activities are now undertaken at the La Hague plant where 17 tons of RTR used fuel from Australia Belgium, and France aligned for treatment. In June 2005, AREVA started to treat at La Hague ANSTO's Australian RTR used fuel from

  1. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2007-09-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  2. Control of criticality risk in the manufacture of fuel elements for research reactors

    International Nuclear Information System (INIS)

    The control of criticality risk in a chemical plant adopts different forms according to the quantities of fissile material and the type of compounds used. This work presents the treatment of the critical excursion risk adopted in production plants of U3 O8 and manufacturing plants of fuel elements for research reactors, located in Constituyentes Atomic Center. The possible events and accidents related to the fissile material control are analyzed, and the systems of administrative control and intrinsic safety through engineering are described. (Author)

  3. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2006-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  4. Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Mckerley, Bill [Los Alamos National Laboratory; Bustamante, Jacqueline M [Los Alamos National Laboratory; Costa, David A [Los Alamos National Laboratory; Drypolcher, Anthony F [Los Alamos National Laboratory; Hickey, Joseph [Los Alamos National Laboratory

    2010-01-01

    We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is in support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts

  5. Theoretical and experimental studies of non-linear structural dynamics of fast breeder reactor fuel elements

    International Nuclear Information System (INIS)

    Descriptions are presented of theoretical and experimental studies of the deformation behaviour of fast-breeder fuel elements as a consequence of extreme impulsive stresses produced by an incident. The starting point for the studies is the assumption that local disturbances in a fuel element have resulted in a thermal interaction between fuel and sodium and in a corresponding increase in pressure. On the basis of the current state of knowledge, the possibility cannot be ruled out that this pressure build-up may lead to the bursting of the fuel-element wrapper, to the propagation of pressure in the core, and to coherent structural movements and deformations. A physical model is established for the calculation of the dynamic response of elastic-plastic beam systems, and the differential equations of p motion for the discrete equivalent system are derived with the aid of D'Alembert's principle. On this basis and with the aid of a semi-empirical pin-bundle model, an appropriate computer program allows a static and dynamic analysis to be obtained for a complete fuel element. In the experimental part of the study, a description is given of static and impulsive loading tests on 1:1 SNR-like fuel-element models. Making use of measured impact forces and of known material characteristics, it was possible to a large extent for the experiments to be reproduced by calculations. In agreement with existing experience from explosion experiments on 1:1 core models, the results (of relevance for fast-breeder safety and in particular the SNR-300) show that only local limited deformations occur and that the compact fuel-element and core structure constitutes an effective inherent barrier in the presence of extreme incident stresses. (author)

  6. Post irradiation examination of HANARO nucler mini-element fuel (metallographic and density test)

    Energy Technology Data Exchange (ETDEWEB)

    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 ''The 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,.

  7. Post irradiation examination of HANARO nucler mini-element fuel (metallographic and density test)

    International Nuclear Information System (INIS)

    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 The 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,

  8. CFD analysis of the 37-element fuel channel for CANDU6 reactor

    International Nuclear Information System (INIS)

    We analyzed the thermal-hydraulic behavior of coolant flow along fuel bundles with appendages of end support plate, spacer pad, and bearing pad, which are the CANDU6 characteristic design. The computer code used is a commercial CFD code, CFX-12. The present CFD analysis model calculates the conjugate heat transfer between the fuel and coolant. Using the same volumetric heat source as the O6 channel, the CFD predictions of the axial temperature distributions of the fuel element are compared with those by the CATHENA (one-dimensional safety analysis code for CANDU6 reactor). It is shown that CFX-12 predictions are in good agreement with those by the CATHENA code for the single liquid convection region (especially before the axial position of the first half of the channel length). However, the CFD analysis at the second half of the fuel channel, where the two-phase flow is expected to occur, over-predicts the fuel temperature, since the wall boiling model is not considered in the present CFD model. (author)

  9. In-pile modelling of nuclear fuel element for the MTR type reactors. Pt. 2

    Energy Technology Data Exchange (ETDEWEB)

    Farhadi, Kazem [AEOI, Tehran (Iran, Islamic Republic of). Radiations Application Research School

    2014-06-15

    In part two of the present paper, neutronic properties of the pool-type research reactor core are used to assess the similitude laws derived for out-of-pile modelling of the fuel element. The benchmark reactor used for this purpose is an IAEA 5 MW thermal pool-type research reactor currently in operation. The neutronic properties analysis are based on typical 2 200 m/sec and neutrons having 0.025 eV energy. The non-leakage capability of the system is estimated in terms of diffusion length. Also the slowing down power and the moderating ratio of the modelled methanol coolant are calculated in terms of lethargy of the diffusing medium. It is shown that the Iron which is substituted for Aluminium cladding is a relatively low absorber of neutrons but has a high neutron leakage. Methanol which replaced ordinary water as coolant is not a suitable coolant due to high neutrons absorbing substance. It is concluded that although Iron as a cladding material and methanol as a coolant meet the modelling out-of-pile criteria but are not satisfying neutronic properties. Therefore, use of them as a model clad and coolant are not suggested for research reactors. (orig.)

  10. International experience and status of fuel element performance and modelling for water reactors

    International Nuclear Information System (INIS)

    Current knowledge concerning water reactor fuel performance and technology is reviewed (212 references). The emphasis is on aspects of in-reactor performance including behaviour in accidents. Computer models for predicting fuel behaviour during the ordinary running of the reactor and during accidents are described. These codes include COMETHE, HOTROD, SLEUTH-SEER and FRAPCON. Their agreement with experimental data is examined. (U.K.)

  11. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  12. Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

    International Nuclear Information System (INIS)

    The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

  13. Studies of the restructuring of fast breeder test reactor fuel by out-of-pile simulation

    International Nuclear Information System (INIS)

    The fast breeder test reactor (FBTR) at Kalpakkam, India, currently employs a mixed carbide of uranium and plutonium with a Pu/(Pu + U) ratio of 0.70 as fuel. The behavior of this fuel in a thermal gradient is investigated. An out-of-pile simulation facility is designed, set up, and commissioned. Experiments are conducted on FBTR fuel pellets to study the restructuring of the fuel at various levels of linear power and its cracking behavior in a thermal gradient. The results are discussed in terms of their significance for reactor operation

  14. An inspection standard of fuel for the high temperature engineering test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Fumiaki; Shiozawa, Shusaku; Sawa, Kazuhiro; Sato, Sadao (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Hayashi, Kimio; Fukuda, Kosaku; Kaneko, Mitsunobu; Sato, Tsutomu.

    1992-06-01

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author).

  15. An inspection standard of fuel for the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author)

  16. Loop facility for investigating of BRIG-300 reactor fuel elements with dissociating N2O4 coolant on ''Mariya'' reactor (PPR)

    International Nuclear Information System (INIS)

    The construction of a gas-cooled loop facility with a dissociating coolant N2O4 is designed that allows the fuel element to be tested under the simultaneous effect of the reactor irradiation and chemically active N2O4 with fast neutron density of (4-6)x1014 neutr./(cm2xs) and pressure of 16.0 MPa. The schematic drawing of the facility and loop channel construction are presented. The principle of the facility operation is described

  17. U-Si and U-Si-Al dispersion fuel alloy development for research and test reactors

    International Nuclear Information System (INIS)

    As part of the National Reduced Enrichment Research and Test Reactor Program, Argonne National Laboratory (ANL) is engaged in a fuel alloy development project. Fuel alloy powder prepared with low-enrichment uranium (235U) is dispersed in an aluminum matrix, and metallurgically roll-bonded within a clad of 6061 Al alloy. Miniplates with up to 55 vol.% fuel alloy (up to 7.0 grams total U per cm3) have been successfully fabricated. Fifty-five of these plates have been or are being irradiated in the Oak Ridge Research Reactor. Three fuel alloys have been used in the ANL miniplates: U3Si (U + 4 wt.% Si), U3Si2 (U + 7.5 wt.% Si), and 'U3SiAl' (U + 3.5 wt.% Si + 1.5 wt.% Al). All are candidates for permitting higher fuel loadings and thus lower enrichments of 235U than would be possible with either UAlx or U3O8, the current fuels for plate-type elements. As an adjunct to the development effort, ANL is engaged in the early stages of technology transfer with commercial fabricators of fuel elements for research reactors. Continuing effort also involves the development of a technology for full-size plate fabrication, and the irradiation of miniplates to a burnup of ∼90% 235U depletion. (author)

  18. Preparation for shipment of spent TRIGA fuel elements from the research reactor of the Medical University of Hannover

    International Nuclear Information System (INIS)

    In the early seventies a research reactor of type TRIGA Mark I was installed in the Department of Nuclear Medicine at the Medical University of Hannover (MHH) for the production of isotopes with short decay times for medical use. Since new production methods have been developed, the reactor has become obsolete and the MHH decided to decommission it. Probably in the second quarter of 1999 all 76 spent TRIGA fuel elements will be shipped to Idaho National Engineering and Environmental Laboratory (INEEL), USA, in one cask of type GNS 16. Due to technical reasons within the MHH a special Mobile Transfer System, which is being developed by the company Noell-KRC, will be used for reloading the fuel elements and transferring them from the reactor to the cask GNS 16. A description of the main components of this system as well as the process for transferring the fuel elements follows. (author)

  19. Preparation for shipment of spent TRIGA fuel elements from the research reactor of the Medical University of Hannover

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele; Cordes, Harro [Medical University of Hannover, D-30625 Hannover (Germany); Ebbinghaus, Kurt; Haferkamp, Dirk [NOELL-KRC, D-97064 Wuerzburg (Germany)

    1998-07-01

    In the early seventies a research reactor of type TRIGA Mark I was installed in the Department of Nuclear Medicine at the Medical University of Hannover (MHH) for the production of isotopes with short decay times for medical use. Since new production methods have been developed, the reactor has become obsolete and the MHH decided to decommission it. Probably in the second quarter of 1999 all 76 spent TRIGA fuel elements will be shipped to Idaho National Engineering and Environmental Laboratory (INEEL), USA, in one cask of type GNS 16. Due to technical reasons within the MHH a special Mobile Transfer System, which is being developed by the company Noell-KRC, will be used for reloading the fuel elements and transferring them from the reactor to the cask GNS 16. A description of the main components of this system as well as the process for transferring the fuel elements follows. (author)

  20. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  1. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

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

  2. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U3O8 were replaced by U3Si2-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is to

  3. Effect of water-chemical regimes and exploitation parameters on the fuel can corrosion and light-water reactor fuel elements security

    International Nuclear Information System (INIS)

    A physico-chemical model of nodal and uniform corrosion of zirconium fuel cans in reactors is proposed. The model takes account of the following physical parameters: values of neutron and thermal fluxes, burnup, temperature, as well as of the water chemistry parameters. The model developed is based on a correlation between zirconium oxide solubility in the coolant and measured values of fuel can corrosion in reactors. Based on the reliability theory, a mathematical model is proposed, connecting the rate of the can corrosion with the fuel element failure intensity at NPPs with RBMK and WWER type reactors

  4. HTGR fuel element size reduction system

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-06-01

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

  5. HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

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

  6. KUEBEL. A Fortran program for computation of cooling-agent-distribution within reactor fuel-elements

    International Nuclear Information System (INIS)

    KUEBEL is a Fortran-program for computation of cooling-agent-distribution within reactor fuel-elements or -zones of theirs. They may be assembled of max. 40 cooling-channels with laminar up to turbulent type of flow (respecting Reynolds' coefficients up to 2.0E+06) at equal pressure loss. Flow-velocity, dynamic flow-, contraction- and friction-losses will be calculated for each channel and for the total zone. Other computations will present mean heat-up of cooling-agent, mean outlet-temperature of the core, boiling-temperature and absolute pressure at flow-outlet. All characteristic coolant-values, including the factor of safety for flow-instability of the most-loaded cooling gap are computed by 'KUEBEL' too. Absolute pressure at flow-outlet or is-factor may be defined as dependent or independent variables of the program alternatively. In latter case 3 variations of solution will be available: Adapted flow of cooling-agent, inlet-temperature of the core and thermal power. All calculations can be done alternatively with variation of parameters: flow of cooling-agent, inlet-temperature of the core and thermal power, which are managed by the program itself. 'KUEBEL' is able to distinguish light- and heavy-water coolant, flow-direction of coolant and fuel elements with parallel, rectangular, respectively concentric, cylindrical shape of their gaps. Required material specifics are generated by the program. Segments of fuel elements or constructively unconnected gaps can also be computed by means of interposition of S.C. 'phantom channels'. (orig.)

  7. A sipping test simulator for identifying defective fuels in MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Highlights: • This simulator based on windows application of C# programming language. • This simulator could be useful for training of technicians in spent nuclear fuels storage facility. • This simulator is user friendly and easy to learn. - Abstract: Integrity of fuel assemblies is critical to continuous operation of any nuclear reactor. NDT methods and sipping test are practical techniques which are used for this purpose. Assessing the fuel integrity by NDT is a troublesome process which could incur personal overdose due to high radiation, requiring large space, and heavy equipment. Therefore to overcome problems associated with the NDT process, sipping test is widely used. The main purpose of this article is introducing sipping test simulator (STS) which is so important for training. Also, this article describes the procedure and methodology used to perform sipping test on the fuel assemblies either in reactor pool or spent fuel storage pool. A unique ability of this simulator is analyzing direct spectroscopy files from experimental data of a real operating reactor. The sipping test simulator is a full-feature training curriculum in spent nuclear fuels storage technology with a PC-based simulator. This simulator is written in C# programming language for a Windows based computer. The simulator will teach everything needed to know for identifying the fuel defects using sipping test process. As learning the basics of sipping test step wise, a freshman operator will soon be able to accomplish all steps in practice

  8. An organization of the thorium fuel cycle start on the basis of fast reactors with spherical fuel elements of the small size

    International Nuclear Information System (INIS)

    The possibility of the organization of thorium fuel cycle start by means of conversion of high background plutonium into isotopically pure Uranium 233 into the highly stressed breeders with the fuel in the form of spherical fuel elements has been studied. A high efficiency of usage of compact plutonium fuel in the form of spherical fuel elements for its transmutation into low background Uranium 233 has been shown as a result of the revealed temporary regularities in the main characteristic behaviour of the reactors of such a type. (authors). 7 refs., 2 figs., 1 tab

  9. Production test IP-603-I evaluation of accidental fuel flushing potential D-DR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hollifield, P.J.; George, D.K.

    1963-07-30

    This purpose of this test is to acquire data to support further studies aimed at determining whether a normal column of I&E enriched fuel elements or a normal column of natural fuel elements can be flushed from a process tube with the crossheader valved down to shutdown position if the front riser pressure is raised to normal operation pressure.

  10. Research reactor fuel bundle design review by means of hydrodynamic testing

    International Nuclear Information System (INIS)

    During the design steps of a fuel bundle for a nuclear reactor, some vibration tests are usually necessary to verify the prototype dynamical response characteristics and the structural integrity. To perform these tests, the known hydrodynamic loop facilities are used to evaluate the vibrational response of the bundle under the different flow conditions that may appear in the reactor. This paper describes the tests performed on a 19 plate fuel bundle prototype designed for a low power research reactor. The tests were done in order to know the dynamical characteristics of the plates and also of the whole bundle under different flow rate conditions. The paper includes a description of the test facilities and the results obtained during the dynamical characterization tests and some preliminary comments about the tests under flowing water are also presented. (author)

  11. Analytical Solution of Fick's Law of the TRISO-Coated Fuel Particles and Fuel Elements in Pebble-Bed High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Two kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation. In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations. (general)

  12. Analytical solution of Fick's law of the TRISO-coated fuel particles and fuel elements in pebble-bed high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Two kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation. In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations. (authors)

  13. Compaction test of fuel element claddings (hulls) and fuel structures waste. 1

    International Nuclear Information System (INIS)

    The PNC (Power and Nuclear Fuel Development Corporation) plans a hull (fuel element cladding) and endpiece hardware compaction facility in Tokai Works. These tests are carried out to confirm the compaction method to be applied. A Series of tests consist of selection of 'simulated hull material', 'design of capsule', 'correlation between pressure and volume reduction ratio', 'estimation of the disk condition' and 'release of zircalloy-fines from the disk during the compressing process'. The results of these tests are as follows. 1) Non-annealing zircalloy for simulated hull material. 2) Optimization of the capsule design. 3) About 80 wt % of the theoretical zircalloy density at 390 MPa pressure. 4) No large void in the disk without cutting the endpiece. 5) The scattering zircalloy fines volume is about 30ppb by the pressing treatment. This test confirm the compaction to be applied. (author)

  14. Radionuclide Compositions and Total Activity of Spent MTR-HEU Fuel Elements of the IAN-R1 Research Reactor

    Science.gov (United States)

    Sarta, Josè A.; Castiblanco, Luis A.

    2005-05-01

    With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energìas Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer of the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.

  15. Analytical Solution of Fick's Law of the TRISO-Coated Fuel Particles and Fuel Elements in Pebble-Bed High Temperature Gas-Cooled Reactors

    Institute of Scientific and Technical Information of China (English)

    CAO Jian-Zhu; FANG Chao; SUN Li-Feng

    2011-01-01

    T wo kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytica,solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation.In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations.%@@ Two kinds of approaches are built to solve the fission products diffusion models(Fick's equation) based on sphere fuel particles and sphere fuel elements exactly.Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented,respectively.The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation.In the fuel element system,a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element.Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations.

  16. Review of consequences of uranium hydride formation in N-Reactor fuel elements stored in the K-Basins

    Energy Technology Data Exchange (ETDEWEB)

    Weber, J.W.

    1994-09-28

    The 105-K Basins on the Hanford site are used to store uranium fuel elements and assemblies irradiated in and discharged from N Reactor. The storage cylinders in KW Basin are known to have some broken N reactor fuel elements in which the exposed uranium is slowly reacting chemically with water in the cylinder. The products of these reactions are uranium oxide, hydrogen, and potentially some uranium hydride. The purpose of this report is to document the results f the latest review of potential, but highly unlikely accidents postulated to occur as closed cylinders containing N reactor fuel assemblies are opened under water in the KW basin and as a fuel assembly is raised from the basin in a shipping cask for transportation to the 327 Building for examination as part of the SNF Characterization Program. The postulated accidents reviews in this report are considered to bound all potential releases of radioactivity and hydrogen. These postulated accidents are: (1) opening and refill of a cylinder containing significant amounts of hydrogen and uranium hydride; and (2) draining of the single element can be used to keep the fuel element submerged in water after the cask containing the can and element is lifted from the KW Basin. Analysis shows the release of radioactivity to the site boundary is significantly less than that allowed by the K Basin Safety Evaluation. Analysis further shows there would be no damage to the K Basin structure nor would there be injury to personnel for credible events.

  17. The OECD Halden reactor project fuels testing programme: methods, selected results and plans

    International Nuclear Information System (INIS)

    The fuels testing programme conducted in the Halden reactor (heavy boiling water reactor (HBWR)) is aimed at providing data for a mechanistic understanding of phenomena, which may affect fuel performance and safety parameters. The investigations focus on implications of high burnup and address thermal property changes, fission gas release as influenced by power level and operation mode, fuel swelling, and pellet-clad interaction. Relevant burnup levels (>50 MWd kg-1 U) are provided through long-term irradiation in the HBWR and through utilisation of re-instrumented fuel segments from commercial light water reactors (LWR). Both urania and MOX fuels are being studied regarding thermal behaviour, conductivity degradation, and aspects of fission gas release. Experiments are also conducted to assess the cladding creep behaviour at different stress levels and to establish the overpressure below which the combination of fuel swelling and cladding creep does not cause increasing fuel temperatures. Clad elongation measurements provide information on the strain during a power increase, the relaxation behaviour and the extent of a possible ratcheting effect during consecutive start-ups. Investigations foreseen in the programme period 2000-2002 include the behaviour of MOX and Gd-bearing fuel and other variants developed in conjunction with burnup extension programmes. Some LWR-irradiated fuel segments will undergo a burnup increase in the HBWR to exposures not yet achieved in LWRs, while others will be re-instrumented and tested for shorter durations

  18. Design of Production Test IP-262-A-11-FP -- Evaluation of projection fuel elements for use in ribbed process tubes -- Demonstration loading

    Energy Technology Data Exchange (ETDEWEB)

    Hodgson, W.H.; Hall, R.E.

    1959-06-29

    For several years, a major category of fuel element failures has been the side corrosion type, characterized by localized accelerated fuel element jacket corrosion. Since it has been demonstrated {sup 1} that misalignment of fuel elements in a process tube will produce flow patterns and accelerated corrosion, termed ``hot spots``, failure to align the fuel elements in process tubes is considered a contributing factor in the production of side corrosion failures. Preliminary testing of both self-supporting and ``bumper`` fuel elements is underway. Data on the self-supporting fuel elements have demonstrated that the bridge-rail projections have sufficient support strength, do not of themselves create a corrosion problem and in actuality probably eliminate any hot-spot areas. Although one tube of bumper fuel elements in KW Reactor {sup 3} has been discharged, data are not as yet available. Potentially, the most sever corrosion conditions exist during the summer months when reactor inlet temperatures are high. It is desirable then, provided bumper fuel elements limit hot- spot corrosion, to evaluate the bumper concept for large scale use possibly by the summer of 1960. To accomplish this, a demonstration loading of the bumper type fuel elements must be underway by about July, 1959. The purpose of this report is to present the design of a test to evaluate the fabrication process and irradiation performance of fuel elements having projections, which may prevent misalignment in ribbed process tubes and meet the aforementioned goals.

  19. Transmission computer microtomography for nondestructive testing of fuel and reactor control means

    International Nuclear Information System (INIS)

    The method for the X-ray transmission computerized microtomography with the defect resolution level up to some μm is developed relative to solving the problem on controlling the quality of the nuclear reactor fuel elements and regularity means. The solution of the problem on the nondestructive control of such objects conditioned the conduct of studies on characteristics of certain range of detectors of the X-ray radiation, organization of scanning principles with application of the laser interferometry principles and construction of the system for the measurement data procession. The performed studies led to the SAPR realization fro the problem-oriented computerized tomographs and the apparatus realization of the experimental version of such a device. The studies on the specially-produced test-samples with calibrated defects demonstrated the correctness of the approach to designing such facilities for the items nondestructive control in the nuclear engineering

  20. Corrosion product deposition on fuel element surfaces of a boiling water reactor

    International Nuclear Information System (INIS)

    Over the last decade the problem of corrosion products deposition on light water reactor fuel elements has been extensively investigated in relation to the possibility of failures caused by them. The goal of the present study is to understand in a quantitative way the formation of such kind of deposits and to analytically understand the mechanism of formation and deposition with help of the quasi-steady state concentrations of a number of 3d metals in reactor water. Recent investigations on the complex corrosion product deposits on a Boiling Water Reactor (BWR) fuel cladding have shown that the observed layer locally presents unexpected magnetic properties. The buildup of magnetic corrosion product deposits (crud) on the fuel cladding of the BWR, Kernkraftwerk Leibstadt (KKL) Switzerland has hampered the Eddy-current based measurements of ZrO2 layer thickness. The magnetic behavior of this layer and its axial variation on BWR fuel cladding is of interest with respect to non-destructive cladding characterization. Consequently, a cladding from a BWR was cut at elevations of 810 mm, where the layer was observed to be magnetic, and of 1810 mm where it was less magnetic. The samples were subsequently analyzed using electron probe microanalysis (EPMA), magnetic analysis and X-ray techniques (μXRF, μXRD and μXAFS). Both EPMA and μXRF have shown that the observed corrosion deposit layer which is situated on the Zircaloy corrosion layer consists mostly of 3-d elements’ oxides (Fe, Zn, Ni and Mn). The distribution of these elements within the investigated layer is rather complex and not homogeneous. The main components identified by 2D μXRD mapping inside the layer were hematite and spinel phases with the common formula (MxFey)[M(1-x)Fe(2-y)]O4, where M = Zn, Ni, Mn. With μXRD it was clearly shown that the cell parameter of analyzed spinel is different from the one of the pure endmembers (ZnFe2O4, NiFe2O4 and MnFe2O4) proving the existence of solid solutions. These

  1. Seismic hazard analysis for the NTS spent reactor fuel test site

    International Nuclear Information System (INIS)

    An experiment is being directed at the Nevada Test Site to test the feasibility for storage of spent fuel from nuclear reactors in geologic media. As part of this project, an analysis of the earthquake hazard was prepared. This report presents the results of this seismic hazard assessment. Two distinct components of the seismic hazard were addressed: vibratory ground motion and surface displacement

  2. Evaluation of the qualification of SPERT [Special Power Excursion Reactor Test] fuel for use in non-power reactors

    International Nuclear Information System (INIS)

    This report summarizes the US Nuclear Regulatory Commission staff's evaluation of the qualification of the stainless-steel-clad uranium/oxide (UO2) fuel pins for use in non-power reactors. The fuel pins were originally procured in the 1960's as part of the Special Power Excursion Reactor Test (SPERT) program. Argonne National Laboratory (ANL) examined 600 SPERT fuel pins to verify that the pins were produced according to specification and to assess their present condition. The pins were visually inspected under 6X magnification and by X-radiographic, destructive, and metallographic examinations. Spectrographic and chemical analyses were performed on the UO2 fuel. The results of the qualification examinations indicated that the SPERT fuel pins meet the requirements of Phillips Specification No. F-1-SPT and have suffered no physical damage since fabrication. Therefore, the qualification results give reasonable assurance that the SPERT fuel rods are suitable for use in non-power reactors provided that the effects of thin-wall defects in the region of the upper end cap and low-density fuel pellets are evaluated for the intended operating conditions. 1 ref., 4 figs., 11 tabs

  3. Cermet fuel reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  4. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  5. Dose Rate Calculations of Spent MTR-HEU Fuel Elements of the IAN-R1 Research Reactor

    Science.gov (United States)

    Sarta Fuentes, Jose Antonio

    2005-04-01

    With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safetly, a program was established at the Instituto de Ciencias Nucleares y Energìas Alternativas (INEA). This program included training, acquisition of hardware and sofware, design and construction of a decay pool, transfer of the spent HEU fuel elements into the decay pool and his final transport to Savanah River in United States. In this paper are presented external dose rates which were calculated for a standard spent MTR-HEU fuel element of the IAN-R1 Research Reactor. The calculations take in consideration the activity due to contributions of fission, activation and actinides products for each relevant radionuclide present in a standard spent MTR-HEU fuel. The datas obtained were the base for the respective dosimetric evaluations in the transfering operations of fuel elements into the decay pool and for shielding calculations in designing of the decay pool.

  6. Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

    International Nuclear Information System (INIS)

    Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators

  7. Advanced fuel in the Budapest research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hargitai, T.; Vidovsky, I. [KFKI Atomic Energy Research Inst., Budapest (Hungary)

    1997-07-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  8. Instrument for measuring a burnup degree of reactor fuel elements by gamma-scanning method

    Energy Technology Data Exchange (ETDEWEB)

    Jozefowicz, E.T.; Kilim, S.; Lewicki, K.; Rusinowski, Z. (Institute of Nuclear Research, Warsaw (Poland))

    1979-01-01

    The principle of gamma scanning of fuel elements is presented. The measuring instrument built in the Institute of Nuclear Research at Swierk and used for gamma scanning of EK-10, WWR-SM, MR and WWER fuel elements is described. The technical drawing of the instrument is given.

  9. Research and Test Reactor Conversion to Low Enriched Uranium Fuel: Technical and Programmatic Progress

    International Nuclear Information System (INIS)

    The U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of High Enriched Uranium (HEU) fuel in research reactors by converting them to low enriched uranium (LEU) fuel. In 2004, the reactor conversion program became the driving pilar of the Global Threat Reduction Initiative (GTRI), a program established by the U.S. DOE's National Nuclear Security Administration. The overall GTRI objectives are the conversion, removal or protection of vunerable civilian radiological and nuclear material. As part of the GTRI, the Conversion Program has accelerated the schedules and plans for conversion of additional research reactors operating with HEU. This paper provides an update on the progress made since 2007 and describes current technical challenges that the program faces. (author)

  10. Eddy current testing of PWR fuel pencils in the pool of the Osiris reactor

    International Nuclear Information System (INIS)

    A nondestructive testing bench is described. It is devoted to examination of high residual power fuel pencils without stress on the cladding nor interference with cooling. Guiding by fluid bearings decrease the background noise. Scanning speed is limited only by safety criteria and data acquisition configuration. Simultaneous control of various parameters is possible. Associated to an irradiation loop, loaded and unloaded in a reactor swinning pool, this bench can follow fuel pencil degradation after each irradiation cycle

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

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-09-01

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

  12. Authorisation to irradiate two U3Si2 prototype fuel elements in the RA-3 research reactor

    International Nuclear Information System (INIS)

    The Argentine RA-3 research reactor (with U3O8-Al LEU MTR fuel type elements and in the process to increase its power from 5 MW to 10 MW since October 2002) is operated by the National Atomic Energy Commission (CNEA). After evaluating the safety studies submitted by CNEA, the Argentine Nuclear Regulatory Authority (ARN) authorised the irradiation of two U3Si2-Al prototype fuel elements named P-06 and P-07, in the RA-3, in September 2000 and September 2001 respectively. These irradiations are part of the qualification programme as manufacturer (CNEA). This report contains a brief description of the regulatory process the ARN carries out in order to approve major modifications or new experiments in research reactors; it also summarises conditions in force for operable core configurations, considers a definition of prototype fuel element and lists the main characteristics of the RA-3 research reactor. Then, the process involved in the authorisation to irradiate the prototypes P-06 and P-07 is particularly analysed. Finally, and in case an authorisation to irradiate a prototype fuel element of a non-qualified material (such as U-Mo) in the RA-3 is requested, the regulatory aspects that should be taken into consideration are analysed. This would be the first time a question of this nature is submitted to the ARN. (author)

  13. Simulation of CANDU Fuel Behaviour into In-Reactor LOCA Tests

    International Nuclear Information System (INIS)

    The purpose of this work is to simulate the behaviour of an instrumented, unirradiated, zircaloy sheathed UO2 fuel element assembly of CANDU type, subjected to a coolant depressurization transient in the X-2 pressurized water loop of the NRX reactor at the Chalk River Nuclear Laboratories in 1983. The high-temperature transient conditions are such as those associated with the onset of a loss of coolant accident (LOCA). The data and the information related to the experiment are those included in the OECD/NEA-IFPE Database (IFPE/CANDU-FIO-131 NEA-1783/01). As tool for this simulation is used the TRANSURANUS fuel performance code, developed at ITU, Germany, along with the corresponding fabrication and in-reactor operating conditions specific of the CANDU PHWR fuel. The results, analyzed versus the experimental ones, are encouraging and perfectible. (author)

  14. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  15. Properties of materials for water reactor fuel elements and methods of measurement

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency upon the proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology and held in the IAEA Headquarters in Vienna from 13 to 16 October 1986. Nineteen participants from 11 countries attended the meeting. Fifteen papers were presented in three sessions covering the following subjects: cladding materials (6 papers), fuel materials (4 papers) and influence of material properties on fuel rod behaviour (5 papers). Three working group meetings covered the above-mentioned topics which were held in order to discuss, in depth, the presentation of papers and to develop the recommendations for future activities aimed at improving properties of cladding and fuel materials for water cooled power reactors. A separate abstract was prepared for each of these 15 papers

  16. Increased burnup of fuel elements

    International Nuclear Information System (INIS)

    The specialists' group for fuel elements of the Kerntechnische Gesellschaft e.V. held a meeting on ''Increased Burnup of Fuel Elements'' on 9th and 10th of November 1982 at the GKSS Research Center Geesthacht. Most papers dealt with the problems of burnup increase of fuel elements for light water reactors with respect to fuel manufacturing, power plant operation and reprocessing. Review papers were given on the burnup limits for high temperature gas cooled reactors and sodium fast breeder reactors. The meeting ended with a presentation of the technical equipment of the hot laboratory of the GKSS and the programs which are in progress there. (orig.)

  17. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

    1997-12-01

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr-Hf alloy or an alloy of Pu-Zr-Hf or a combination of both.

  18. Procedures and techniques for the management of experimental fuels from research and test reactors. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    Almost all countries that have undertaken fuel development programs for power, research or military reactors have experimental and exotic fuels, either stored at the original research reactors where they have been tested or at some away-from-reactor storage facility. These spent fuel liabilities cannot follow the standard treatment recognized for modern power reactor fuels. They include experimental and exotic fuels ranging from liquids to coated spheres and in configurations ranging from full test assemblies to post irradiation examination specimens set in resin. This document contains an overview of the extent of the problem of managing experimental and exotic fuels from research and test reactors and an expert evaluation of the overall situation in countries which participated in the meeting

  19. Safety research program of LWR fuels and materials using the Japan materials testing reactor

    International Nuclear Information System (INIS)

    Power up-rates, burn-up extension and long term operation enable us to utilize Light Water Reactors efficiently. This will have the fuels and structural materials exposed to severe operational condition for a longer period, which can affect their integrity. Continuous researches for solving irradiation-related issues on the fuels for high-duty uses and the plant aging are essential in order to realize the up-graded uses of LWR safely. Japanese regulator has decided to install new irradiation test facilities in the Japan Material Testing Reactor (JMTR) at the Japan Atomic Energy Agency (JAEA). For the fuels tests, transient tests facility is being constructed for the power transient tests of new design BWR fuels. For the materials tests, the irradiation test loops under well controlled environment simulating BWR water chemistry condition and a large irradiation capsule, which can accommodate 1 inch-thickness compact tension specimens in an inert gas environment, are being prepared for the researches on stress corrosion cracking and irradiation embrittlement, respectively. These fuels and materials irradiation tests will be started in 2011 after refurbishment of JMTR. (author)

  20. MR-6 type fuel elements cooling in natural convection conditions after the reactor shut down

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Bykowski, W.; Moldysz, A. [Institute of Atomic Energy, Otwock Swierk (Poland)

    2002-07-01

    Natural cooling conditions of the nuclear fuel in the channel type reactor after its shut down are commonly determined with relatively high uncertainty. This is not only to he lack of adequate measurements of thermal parameters i.e. the residual power generation, the coolant flow and temperatures, but also due to indeterminate model of convection mechanism. The numerical simulation of natural convection in multitube fuel assembly in the fuel channel leads to various convection modes including evidently chaotic behaviour. To determine the real cooling conditions in the MARIA research reactor a series of experiments has been performed with fuel assembly equipped with a set of thermocouples. After some forced cooling period (the shortest was half an hour after the reactor shut down) the reactor was left with the only natural convection. Two completely different cooling modes have been observed. The MARIA core consists of series of individual fuel channel and so called bypasses, maintaining the hydraulic properties of the fuel channel, connected in parallel. Initially, the convection cells were established trough few so-called bypasses providing a very effective mode of cooling. In this mode the flow charts were identical to those existing in forced cooling mode. After certain period the system switched on the second cooling mode with natural circulation within the individual fuel cells. Higher temperatures and temperature fluctuations were characteristic for this mode approaching 30 deg in amplitude. In almost all the cases the system was switching few times between modes, but eventually remained in the second mode. The switching times were not regular and the process has a chaotic behaviour. (author)

  1. Examination on the safety criteria in safety assessment of fuel and core of the High Engineering Test Reactor (HTTR) under abnormal conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio; Shindo, Masami [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Shiozawa, Shusaku [and others

    1991-09-01

    The safety criteria of the fuel and the core of the High Temperature Engineering Test Reactor (HTTR) was examined in order to assess the safety of the reactor at anticipated operational transients and at accidents. The safety criteria at the transients were decided as follows; the maximum fuel temperature shall not exceed 1600degC for the reactor core fuel, and it shall not exceed 2500degC for the fuel limit irradiation specimen, which is a fuel assembly with a graphite block and fuel elements for studying the fuel failure limit. On the other hand, the criteria for the accidents were decided so that (i) the fuel rod shall be held within the graphite block, and (ii) the support post and the post sheet shall maintain the strength enough to support the reactor core. Validity of the above criteria was ensured by the fuel behavior at representative abnormal events, namely, `abnormal events at the fuel limit irradiation test`, `rapid reactivity insertion events`, `air ingress accidents`, and `fuel assembly fall accidents`. (author).

  2. Examination on the safety criteria in safety assessment of fuel and core of the High Temperature Engineering Test Reactor (HTTR) under abnormal conditions

    International Nuclear Information System (INIS)

    The safety criteria of the fuel and the core of the High Temperature Engineering Test Reactor (HTTR) was examined in order to assess the safety of the reactor at anticipated operational transients and at accidents. The safety criteria at the transients were decided as follows; the maximum fuel temperature shall not exceed 1600degC for the reactor core fuel, and it shall not exceed 2500degC for the fuel limit irradiation specimen, which is a fuel assembly with a graphite block and fuel elements for studying the fuel failure limit. On the other hand, the criteria for the accidents were decided so that (i) the fuel rod shall be held within the graphite block, and (ii) the support post and the post sheet shall maintain the strength enough to support the reactor core. Validity of the above criteria was ensured by the fuel behavior at representative abnormal events, namely, 'abnormal events at the fuel limit irradiation test', 'rapid reactivity insertion events', 'air ingress accidents', and 'fuel assembly fall accidents'. (author)

  3. Apparatus for locating defective nuclear fuel elements

    International Nuclear Information System (INIS)

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

  4. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO2 in ThO2) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO2 (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  5. Experimental investigations for determination of heat-transfer coefficients and temperature fields in simulated fuel assemblies of BREST reactor with fuel elements spaced by transverse grids

    International Nuclear Information System (INIS)

    The consideration is given to heat transfer and temperature fields in fuel pin bundles with transverse spacer grids (s/d =1.33) equally spaced along energy deposition length. Experimental data are obtained on two simulated 37-rod core assemblies: one assembly is with uniform geometry along the cross-section and in the other there is nonheated rod simulating supporting pipe in fuel assembly of reactor with heavy coolant. Eutectic Na-K alloy is used as coolant. Nusselt numbers and temperature nonuniformity along the perimeter of measurement fuel element simulator obtained in these assemblies are compared as well as available data for finned (wire to wire) fuel rods

  6. Spacer for a fuel element

    International Nuclear Information System (INIS)

    Spacers for fuel pins arranged to form congish fuel elements can be shaped as plates with openings in accordance with the fuel pin grid. Such a plate that covers the cross section of a fuel element consists according to the invention of at least two parts that are offset in the fuel element's longitudinal direction and joint hinge-like in at least one grid position. Thus, one has smaller parts that are easier to work on with due accuracy. The invention is designed in particular for breeder reactors and high-conversion reactors. (orig.)

  7. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation

  8. Ultrasonic evaluation of end cap weld joints of fuel elements of pressurized heavy water reactors using signal analysis methods

    International Nuclear Information System (INIS)

    This paper describes the application of ultrasonic digital signal analysis for the detection of fine defects of the order of 10% or lower of wall thickness (WT) of 370 microns in the resistance welded end cap-cladding tube joints of fuel elements used in Pressurised Heavy Water Reactors (PHWR s). The results obtained for the detection of such defects, have confirmed the sensitivity and reliability of this approach, and were further validated by destructive metallography. (author)

  9. 3D laser inspection of fuel assembly grid spacers for nuclear reactors based on diffractive optical elements

    Science.gov (United States)

    Finogenov, L. V.; Lemeshko, Yu A.; Zav'yalov, P. S.; Chugui, Yu V.

    2007-06-01

    Ensuring the safety and high operation reliability of nuclear reactors takes 100% inspection of geometrical parameters of fuel assemblies, which include the grid spacers performed as a cellular structure with fuel elements. The required grid spacer geometry of assembly in the transverse and longitudinal cross sections is extremely important for maintaining the necessary heat regime. A universal method for 3D grid spacer inspection using a diffractive optical element (DOE), which generates as the structural illumination a multiple-ring pattern on the inner surface of a grid spacer cell, is investigated. Using some DOEs one can inspect the nomenclature of all produced grids. A special objective has been developed for forming the inner surface cell image. The problems of diffractive elements synthesis, projecting optics calculation, adjusting methods as well as calibration of the experimental measuring system are considered. The algorithms for image processing for different constructive elements of grids (cell, channel hole, outer grid spacer rim) and the experimental results are presented.

  10. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  11. BOTHER: a steady-state code that predicts margin to burnout heat flux for N-Reactor fuel elements

    International Nuclear Information System (INIS)

    In order to operate a nuclear reactor safely, some method must be available which can adequately describe the thermal-hydraulics of the reactor core. Further, some method must be available which can be used to predict the effects of changes in system operation. For example it is often necessary to know or be able to predict the effects of reduced coolant flow, front or rear peaked power distribution, etc., on the overall safe operation of the reactor. Because of the uniqueness of the N Reactor (horizontal pressure tubes with no crossflow between tubes or annular subchannels) the commonly available thermal-hydraulics codes are generally not directly applicable. For these reasons the BOTHER (BurnOut THErmal Ratio) computer code has been developed at UNI. Using experimental results for N Reactor flow splits and heat splits as well as enthalpy imbalance and critical heat flux data, BOTHER computes the steady state margin to burnout for N Reactor fuel elements. The equations used by BOTHER to perform the burnout calculations are described. A sample problem for MARK-IV fuel with input and output listings is also included

  12. Disposal of irradiated fuel elements from German research reactors. Status and outlook

    Energy Technology Data Exchange (ETDEWEB)

    Thamm, G. [Central Research Reactor and Nuclear Operations Division, Research Centre Juelich, Forschungszentrum Juelich GmbH, Juelich (Germany)

    1999-07-01

    There will be a quantity of highly radioactive spent nuclear fuel (snf) from German research reactors amounting to about 9.1 t by the end of the next decade, which has to be disposed of. About 4.1 t of this quantity are intended to be returned to the USA. The remaining approximately 5 t can be loaded into approximately 30 CASTOR-2 casks and will be stored in a central German dry interim store for about 30 to 50 years (first step of the domestic disposal concept). Of course, snf arising from the operation of research reactors beyond 2010 has to be disposed of in the same way (3 MTR-2 casks every two years for BER-II and FRM-II). It is expected that snf from the zero-power facilities probably will be recycled for reusing the uranium. Due to the amendment of the German Atomic Energy Act intended by the new Federal German Government, the interim dry storage of snf from power reactors in central storage facilities like Ahaus or Gorleben will be stopped and the power reactors have to store snf at their own sites. Although the amendment only concerns nuclear power reactors, it could not be excluded that snf from research reactors, too, cannot be stored at Ahaus or Gorleben at present. (author)

  13. Development of mechanical test techniques on the irradiated grid elements in PWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang Bok; Kim, Do Sik; Baik, Seung Jai; Choo, Yong Sun; Baek, Sang Youl; Ryu, Woo Seok [Korea Atomic Energy Research Institute, Daejon (Korea, Republic of); Ha, Dong Keun; Seo, Jeong Min [Korea Nuclear Fuel Company, Daejon (Korea, Republic of)

    2008-11-15

    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 paper are used to produce the irradiation data of the grid 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 30MW thermal output at 300.deg.C. From the spring tests 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. 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.

  14. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

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

  15. Post irradiation examination of thermal reactor fuels

    Science.gov (United States)

    Sah, D. N.; Viswanathan, U. K.; Ramadasan, E.; Unnikrishnan, K.; Anantharaman, S.

    2008-12-01

    The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (˜15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ˜11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

  16. Conceptual design of a uranyl nitrate fueled reactor for the destructive testing of liquid metal fast breeder reactor fuel subassemblies

    International Nuclear Information System (INIS)

    A preliminary design of a uranyl nitrate test reactor is developed, with emphasis placed on the core neutronics and cross section development. ENDF/B-IV cross section data and the AMPX system were used to develop a 25 group neutron cross section library. A series of one-dimensional transport calculations were made in order to arrive at a reference design. Power densities of 16.5 Kw/1 appear to be attainable in the 217 pin FFTF test subassembly, with a peak neutron flux in the test zone of 2.4 x 1014 n/cm2-sec. Other engineering features pertinent to the overall system design are discussed, including: (1) corrosion, (2) treatment of radiolytic gas, (3) heat removal, and (4) reactor control

  17. A thermal-hydraulic test rig for advanced fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    A new design of fast reactor fuel assemblies has been proposed in which the pins are supported in grids attached to the wrapper by flexible skirts. Coolant mixing is enhanced by the skirts diverting flow into the cluster of pins at each grid. There are insufficient empirical data available for the detailed design of the skirt or for the input to computer calculations of flow and heat transfer. A test rig to provide these data has been designed and built. (author)

  18. Evaluation method and prediction result of fuel behavior during the High Temperature Engineering Test Reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Sawa, Kazuhiro; Yoshimuta, Shigeharu; Sato, Masashi; Saito, Kenji; Tobita, Tsutomu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    Small amounts of additional failure of HTGR (High Temperature Gas-cooled Reactor) fuel will occur during operation. In the safety design requirements for the High Temperature Engineering Test Reactor (HTTR) fuel, the additional failure fraction in the coating layers of the coated fuel particles is limited less than 0.2% through the full service period. The failure fraction should be know during the HTTR operation. Short-lived fission gases are released from the through-coatings-failed particles and contamination uranium in the fuel compact matrix since the coating layers can retain short-lived fission gases. Then fission gas concentration in the primary coolant reflects the failure fraction in the core. Based on fuel fabrication data (exposed uranium fractions and the SiC failure fractions) and the HTTR operating condition, the though-coatings-failure fraction and release fraction of {sup 88}Kr are analytically predicted. The results are as follows. (1) The intact particles will not fail by kernel migration, Pd-SiC corrosion and internal pressure, however, some of the as-fabricated SiC-failed particles will be the through-coatings-failed particles by the pressure vessel failure. (2) The release fraction of {sup 88}Kr, that will be determined mainly by the release from the contamination uranium in the fuel compact matrix, will be less than 10{sup -6} considering the additional through-coatings-failure fraction. (author)

  19. Design criteria, production and total integrity assessment of fuels of the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    This report describes the design criteria, production and total integrity of the HTTR fuels for the safety design of the reactor. The fuels were designed so that they should not lose their integrity even though taking account of various kinds of possible deteriorations during reactor service. Sufficiently low values of initial (as-produced) fuel failure fractions have been achieved, and experience of fuel production is enough for full core loading. Results of the present assessment have shown that total integrity of the fuels will be maintained successfully in terms of coating failure of the fuel particles, thermal and mechanical performance of the fuel compacts, graphite sleeves and fuel assemblies. (author)

  20. The possibility of prediction of the lifetime of metallic nuclear fuel elements in a radiation field of thermal nuclear reactors

    International Nuclear Information System (INIS)

    An attempt is made to clarify the possible causes of failure of irradiated nuclear fuel cartridges, in order to determine the parameters which govern the lifetime of the fuel and a way to predict it. Measurements of mechanical properties of irradiated uranium metal and cladding, can serve as a basis for failure prediction. Testing irradiated fuel elements by bending till fracture enables to evaluate the integral character of the fuel element, along the cross-section, taking into account the difference in brittleness of several zones. It is likely that the bending test, which indicates the behaviour of a stress-strain function, is a faster and more reliable way to determine the mechanical properties of irradiated nuclear fuel. Since the stresses applied to the cladding during irradiation are locally hydrostatic, its postirradiation blow-up provide information on strength and elasticity variations of the irradiated cladding material. (B.G.)

  1. System for characterization of objects, particularly of nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    In order to be able to unmistakably identify fissionable materials of single fuel elements within the framework of monitoring, it is suggested to make impurity inclusions in defined regions consisting of materials which can be easily recognized by an ultrasonic sensor device. These 'sealing samples' enable perfect identification and are safe against falsing. (HP)

  2. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  3. Technical concept for a test of geologic storage of spent reactor fuel in the climax granite, Nevada Test Site

    International Nuclear Information System (INIS)

    We plan to emplace spent fuel assemblies from an operating commercial nuclear reactor in the Climax granite at the US Department of Energy's Nevada Test Site. In this generic test, 11 canisters of spent fuel will be emplaced with 6 electrical simulator canisters in a storage drift 420 m below in surface and their effects compared. Two adjacent drifts will contain electrical heaters, operated to simulate the temperature-stress-displacement fields of a large repository. We describe the test objectives, the technical issues, the site, the preoperational measurement program, thermal and mechanical response calculations, the characteristics of the spent fuel, the field instrumentation and data-acquisition systems, and the system for handling the spent fuel

  4. Reactivity calculations for the fuel elements of I.T.U. TRIGA MARK-II reactor by means of one-group perturbation theory

    International Nuclear Information System (INIS)

    The reactivities of the fuel elements of I.T.U. TRIGA MARK-II reactor has been calculated by using both one-group perturbation theory and a one-dimensional, two-group diffusion computer code TRIGAP. For each fuel element, reactivities calculated by both methods are compared with those measured experimentally. It is seen that the reactivity calculations made by using the one-group perturbation theory give the results with better accuracy in comparison to TRIGAP. One-group perturbation theory can be easily applied to the reactivity calculations of fuel elements of TRIGA type reactors in acceptable range (orig.)

  5. Impact of uranium concentration reduction in side plates of the fuel elements of IEA-R1 reactor on neutronic and thermal hydraulic analyses

    International Nuclear Information System (INIS)

    This master thesis presents a study to verify the impact of the uranium concentration reduction in the side plates of the reactor IEA-R1 fuel elements on the neutronic and thermal-hydraulic analyses. To develop such study, a previous IPEN-CNEN/SP research was reproduced by simulating the fuel elements burn-up, with side plate uranium density reduced to 50, 60 and 70% of the standard fuel element plates. This research begins with the neutronic analysis using the computer code HAMMER and the first step consists in the calculation of the cross section of all materials presented at the reactor core, with their initial concentration; the second step consists in the calculation of the fast and thermal neutron group fluxes and power densities for fuel elements using the computer code CITATION. HAMMER output data is used as input data. Once the neutronic analysis is finished and the most critical fuel elements with highest power density have been defined, the thermal-hydraulics analysis begins. This analysis uses MCTR-IEA-R1 thermal-hydraulics model, which equations are solved by commercial code EES. Thermalhydraulics analysis input is the power density data calculated by CITATION: it is considered the highest power density on each fuel element, where there is a higher energy release and, consequently, higher temperatures. This data is used on energy balance equations to calculate temperatures on critical fuel element regions. Reactor operation comparison for three different uranium densities on fuel side plates is presented. Uranium density reduction contributes to the cladding surface temperature to remain below the established limit, as reactor operation safety requirement and it does not affect significantly fuel element final burn-up nor reactor reactivity. The reduction of uranium in the side plates of the fuel elements of the IEA-R1 showed to be a viable option to avoid corrosion problems due to high temperatures. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

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

  7. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have lead to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

  8. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have led to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

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

    International Nuclear Information System (INIS)

    Knowledge of fuel limits and safety margins in normal and off-normal transients in nuclear power plants remains a constant preoccupation for electricity producers and fuel manufacturers. Accurate determination of such limits, through fuel irradiation testing in the OSIRIS reactor at Saclay is closely linked to the reliability of appropriate instrumentation techniques. Two paths are currently followed to obtain short experimental rods: segmented fuel coming directly from power plants, or re-fabrication of rods in hot cells with our FABRICE process. It can be associated with instrumentation such as fuel centerline thermocouple in annular pellets, pressure transducer or fission gas release measurement by gamma-spectrometry using helium sweeping, in analytic experiments. Our present development, to be implemented in 1993, is the the centerline instrumentation of a fuel column with solid pellets. Inserting the thermocouple requires a cold drilling machine, using CO2 freezing of broken UO2 (with liquid nitrogen). During the fuel rod irradiation itself, we try to lower the uncertainties associated to power determination, using thermal balance or neutronic calibration, or even gamma spectrometry. A description of the new test train designed for the ISABELLE water loop in OSIRIS is given, with special emphasis on instrumentation: a LVDT for measuring fuel rod elongation and eventual clad failure, and increased number and better localization of thermocouples and SPDN. The third part is devoted to the measurements by optical microdensitometry of neutron radiographs of the fuel pellet dish modification after irradiation. Dishes are generally disappearing through thermal and mechanical deformation of the pellet, and this can eventually be modelized to better understand pellet-cladding mechanical interaction. (author). 3 refs, 5 figs

  10. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey Phillips; Charles Barnes; John Hunn

    2010-10-01

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on

  11. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    International Nuclear Information System (INIS)

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called 'AGR-1,' graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on fuel

  12. A review of theoretical and experimental studies underlying the thermal-hydraulic design of fast reactor fuel elements

    International Nuclear Information System (INIS)

    The economic performance of fast reactors is closely linked to the achievable burn-up of heavy atoms, that is to the endurance life of the fuel pins. The safety case must also be concerned with the integrity of the cladding, since this is the primary containment envelope for fission products. It is thus important to ensure that cladding temperatures during reactor operation are limited to levels which incur no serious impairment of mechanical properties. The function of thermal-hydraulic analysis is to provide fuel element designers with the means of achieving this objective. This paper reviews the theoretical approaches which have been developed and applied in the UK in the design of LMFBR fuel and breeder sub-assemblies, control rods and experimental clusters. It also presents results of experimental studies undertaken to develop a better understanding of coolant flow distribution and mixing problems in these components, and to provide essential data for computer codes. Problem areas in this field are highlighted, particularly the difficulties arising due to irradiation induced distortions. Reference is made to the experimental and theoretical developments which are in progress, or may be required, to provide adequate predictions of fuel pin temperatures at high burn-up. (author)

  13. Thermal analysis for a spent reactor fuel storage test in granite

    International Nuclear Information System (INIS)

    A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy's Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m3/s and an ambient rock temperature of 230C, the maximum calculated rock temperature (near the center of the heat source) is about 1000C while the maximum air temperature in the drift is around 400C. This ventilation (1 m3/s through the main drift and 1/2 m3/s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage

  14. Report transparency and nuclear safety 2007 CEA Cadarache; Rapport transparence et securite nucleaire 2007 CEA Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    This report presents the activities of the CEA Center of Cadarache for the year 2007. The actions concerning the safety, the radiation protection, the significant events, the release control and the environmental impacts and the wastes stored on the center are discussed. More especially the report discusses the beginning of the RJH reactor construction, the fourth generation reactors research programs, the implementing of la Rotonde the new radioactive wastes management installation, the renovation of the LECA. (A.L.B.)

  15. Feasibility study of power reactor fuel elements factory development: I. Economical aspects

    International Nuclear Information System (INIS)

    For determining the feasibility study on manufacturing nuclear fuel element from economical aspect point of view, it necessary to fix its capacity which it was found from fuel element reloading requirement for nuclear power plat (PLTN). NEWJEC report which use as a base in this study that is possibly of a complex of NPP as big as 7200 MW in Muria region. If the capacity factor is 80 %, the reload requirement is therefore become from 120 to 142 tons uranium every year. So, its considered to fix the nominal capacity of a fabric for nuclear fuel element manufacturing as much as 200 tons-U per year with economical lifetimes of 20 years. NEWJEC data show, for manufacturing capacity of 200 tons-U per year with, plant have a fixed capital investment of US$ 43.9 million. With working capital as much as 15 % correspond to fixed capital investment (FCI); 10 % of interest rate; US$ 17 million of fixed cost; US$ 106.2/kg-U of variable production cost, its calculated that break even point/BEP is 50 % for price of nuclear fuel is US$ 350/kg-U without uranium cost. On this economic condition, it was found that the return on investment/ROI is 20.2 %; the internal rate of return/IRR is 11.2 % and the benefit cost ration/BCR is 1.22. For all of above, it was assumed that such nuclear fuel element manufacturing service will be operate in the year of 2012. Some of NEWJEC data have been revised, there were the value of FCI; cost of salary; the value in percent of working capital/WC; the cost of non-uranium materials and the price of product service are US$ 68 million; US$ 4.1 million; 30 %; US$ 100/kg-U and US$ 370/kg-U respectively, where the new data appear as higher than old date from NEWJEC, excluding the cost of salary. For all new economical data in the latest, we found that 45 %; 16.73 %; 11.8 % and 1.25 for BEB; IRR and BCR respectively

  16. Development of mechanical test techniques on the irradiated grid elements in PWR fuel assembly

    International Nuclear Information System (INIS)

    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 paper are used to produce the irradiation data of the grid 1x1 cell springs, 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 300degC. From the spring tests 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. (author)

  17. Potential of duplex fuel in prebreeder, breeder, and power reactor designs: tests and analyses (AWBA Development Program)

    International Nuclear Information System (INIS)

    Dual region fuel pellets, called duplex pellets, are comprised of an outer annular region of relatively high uranium fuel enrichment and a center pellet of fertile material with no enrichment. UO2 and ThO2 are the fissile and fertile materials of interest. Both prebreeders and breeders are discussed as are the performance advantages of duplex pellets over solid pellets in these two pressurized water reactor types. Advantages of duplex pellets for commercial reactor fuel rods are also discussed. Both irradiation test data and analytical results are used in comparisons. Manufacturing of duplex fuel is discussed

  18. Hydraulic modelling of the CARA Fuel element

    International Nuclear Information System (INIS)

    The CARA fuel element is been developing by the National Atomic Energy Commission for both Argentinean PHWRs. In order to keep the hydraulic restriction in their fuel channels, one of CARA's goals is to keep its similarity with both present fuel elements. In this paper is presented pressure drop test performed at a low-pressure facility (Reynolds numbers between 5x104 and 1,5x105) and rational base models for their spacer grid and rod assembly. Using these models, we could estimate the CARA hydraulic performance in reactor conditions that have shown to be satisfactory. (author)

  19. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  20. Assessment of fuel integrity of the High Temperature Engineering Test Reactor (HTTR) and its permissible design limit

    International Nuclear Information System (INIS)

    This report describes the results of integrity assessment of the HTTR fuel as a part of safety design of the reactor. Functions of individual coating layer of the coated fuel particles were investigated to clarify the characteristic properties of the HTTR fuel. Integrity of the fuel under normal operating conditions of HTTR was shown to be maintained through analyses on fuel failure and due to kernel migration (amoeba effect) and corrosion of the silicon carbide layer by palladium. The fuel integrity has been demonstrated by successful results of the irradiation performance tests. By considering the failure behavior of coated fuel particles, the permissible design limit of the fuel was determined such that the fuel temperature should not exceed 1600degC under abnormal transient conditions. The validity of the determination of the limit has been confirmed by results of heating experiments of coated fuel particles at extremely high temperature. (author)

  1. Fuel elements of thermionic converters

    Energy Technology Data Exchange (ETDEWEB)

    Hunter, R.L. [ed.] [Sandia National Labs., Albuquerque, NM (United States). Environmental Systems Assessment Dept.; Gontar, A.S.; Nelidov, M.V.; Nikolaev, Yu.V.; Schulepov, L.N. [RI SIA Lutch, Podolsk (Russian Federation)

    1997-01-01

    Work on thermionic nuclear power systems has been performed in Russia within the framework of the TOPAZ reactor program since the early 1960s. In the TOPAZ in-core thermionic convertor reactor design, the fuel element`s cladding is also the thermionic convertor`s emitter. Deformation of the emitter can lead to short-circuiting and is the primary cause of premature TRC failure. Such deformation can be the result of fuel swelling, thermocycling, or increased unilateral pressure on the emitter due to the release of gaseous fission products. Much of the work on TRCs has concentrated on preventing or mitigating emitter deformation by improving the following materials and structures: nuclear fuel; emitter materials; electrical insulators; moderator and reflector materials; and gas-exhaust device. In addition, considerable effort has been directed toward the development of experimental techniques that accurately mimic operational conditions and toward the creation of analytical and numerical models that allow operational conditions and behavior to be predicted without the expense and time demands of in-pile tests. New and modified materials and structures for the cores of thermionic NPSs and new fabrication processes for the materials have ensured the possibility of creating thermionic NPSs for a wide range of powers, from tens to several hundreds of kilowatts, with life spans of 5 to 10 years.

  2. RA-3 reactor core with uranium silicide fuel elements P-07 type

    International Nuclear Information System (INIS)

    Following the studies on the utilization of fuel elements (FE) containing uranium silicide, core of the RA-3 was analyzed with several calculation models. At first, the present situation, i.e. the core charged with normal FE (U3O8), has been analyzed to validate the simulation methodology comparing with experimental results and to establish reference data to 5 and 10 MW able to be compared with future new situations. Also, CITVAP's nuclear data libraries to be used in irradiation experiment planning were completed. The results were satisfactory and were applied to the study of the core containing P-07 FE [U3Si2], in face of a future core change. Comparing with the performance of the U3O8FE, the silicides ones show the following advantages: - average burnup: 45 % greater; -extraction burnup increase 12 %; and, -the residence time [in full power days] could be a 117 % greater. (author)

  3. NEUTRONIC REACTOR FUEL COMPOSITION

    Science.gov (United States)

    Thurber, W.C.

    1961-01-10

    Uranium-aluminum alloys in which boron is homogeneously dispersed by adding it as a nickel boride are described. These compositions have particular utility as fuels for neutronic reactors, boron being present as a burnable poison.

  4. Management and storage of spent nuclear fuel at research and test reactors. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    Irradiated fuel from research and test reactors has been stored at various facilities for several decades. As these facilities age and approach or exceed their original design lifetimes, there is mounting concern about closure of the fuel cycle and about the integrity of ageing fuels from the materials point of view as well as some concern about the loss of self-protection of the fuels as their activity decays. It is clear that an international effort is necessary to give these problems sufficient exposure and to ensure that work continues on appropriate solutions. The future of nuclear research, with its many benefits to mankind, is in jeopardy in some countries, especially countries without nuclear power programmes, because effective solutions for extended interim storage and final disposition of spent research reactor fuels are not yet available. An advisory Group meeting was convened in Vienna to consider a Database on the Management and Storage of Spent Nuclear Fuel from Research and Test Reactors. Sixteen experts from sixteen different countries participated in the Advisory Group meeting and presented country reports, which together represent an overview of the technologies used in spent fuel management and storage at research and test reactors world-wide. The sixteen country reports together with the database summary are presented in this publication. Refs, figs, tabs

  5. Non-Destructive Testing Methods Applied to Multi-Finned SAP Tubing for Nuclear-Fuel Elements

    International Nuclear Information System (INIS)

    The Danish Atomic Energy Commission has undertaken a design study oi an organic-cooled, heavy- water-moderated power reactor. The fuel element for the reactor is a 19-rod bundle; the fuel rods contain sintered uranium-dioxide pellets canned in 2-m long, helically-finned tubes of Sintered Aluminium Product (SAP). A very high quality of the canning tubes is necessary to obtain the optimum heat-transfer conditions and to maintain the integrity of the fuel element during reactor service. Two examples of tube design illustrate the narrow dimensional tolerances. In order to ensure an adequate quality of the canning tubes, a stringent quality control has been established, to a wide extent based upon non-destructive methods. An account is presented of the non-destructive techniques developed for measuring wall thickness and diameters and for detecting defects. The complex 24-finned cross-section prevents the application of ultrasonic or eddy-current methods for wall-thickness measurements. Therefore, a special recording beta-gauge has been developed, based upon the attenuation of beta radiation from a Sr90 source placed inside the tube. An ultrasonic immersion resonance method is used for the continuous recording of the wall thickness of the more simple 12-finned tube design. Inner and outer (across fin tips) diameters are continuously recorded by rapid air-gauge systems. Flaw detection is carried out by the ultrasonic pulse-echo immersion technique and by eddy-current inspection.. Transverse cracks can easily be detected by the ultrasonic method whereas inspection for longitudinal flaws has not appeared feasible with this method. Therefore, eddy-current inspection is applied in addition to the ultrasonic testing. (author)

  6. The Application of Various Nondestructive Testing Methods to Fuel Elements of the Orgel Type

    International Nuclear Information System (INIS)

    The paper describes the various methods employed to detect flaws (dimensional or structural) in fuel-element canning tubes. The authors also describe the final tests on complete fuel elements, in particular radiography of welds and leak-tightness tests. This subject has already been discussed to some extent. The dimensional characteristics of smooth SAP (sintered aluminium powder) canning tubes have been fairly extensively investigated, and in particular: 1. The internal and external diameters have been measured using pneumatic pick-ups and recording the result; 2. The thicknesses have been measured using either ultrasonic resonance methods or y-rays (a Euratom- Istituto Sperimentale Metalli Leggeri contract); 3. Checking the deflection; 4. Tests of finned tubes. Work has also been carried out on detecting flaws in smooth canning tubes, and rejection criteria have been adopted depending on the prospective use of the tubes. (a) The making of artificial flaws corresponding to the harmfulness of actual flaws in the SAP is described. This study revealed high sensitivity to flaws of the longitudinal type generally caused by large inclusions during processing. (b) Ultrasonic tests. Longitudinal flaws: Comparison between the method with two pick-ups and that with one shows the limitations of these two methods. Transverse flaws: The single pick-up method used in investigating these is briefly described. Mechanical drive: A laboratory type mechanical test bench for investigating test criteria and a special semi-industrial bench for the continuous inspection of the tubes and the recording of flaws are mentioned. The difficulties encountered and the steps taken to prevent them are described. (c) Radiographic tests. This method will be discussed in a special paper; here we simply indicate the results obtained on pressure tubes and canning tubes. (d) Various tests. The final tests on complete fuel elements can be summed up in two sections: Helium leak-tests developed by SOGEV

  7. Thermal analysis for a spent reactor fuel storage test in granite

    Energy Technology Data Exchange (ETDEWEB)

    Montan, D.N.

    1980-09-01

    A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy`s Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m{sup 3}/s and an ambient rock temperature of 23{sup 0}C, the maximum calculated rock temperature (near the center of the heat source) is about 100{sup 0}C while the maximum air temperature in the drift is around 40{sup 0}C. This ventilation (1 m{sup 3}/s through the main drift and 1/2 m{sup 3}/s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage.

  8. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  9. Design of high temperature irradiation materials inspection cells. (Spent fuel inspection cells) in the High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ino, Hiroichi; Ueta, Shouhei; Suzuki, Hiroshi; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tobita, Tsutomu [Nuclear Engineering Company, Ltd., Tokai, Ibaraki (Japan)

    2002-01-01

    This report summarizes design requirements and design results for shields, ventilation system and fuel handling devices for the high temperature irradiation materials inspection cells (spent fuel inspection cells). These cells are small cells to carry out few post-irradiation examinations of spent fuels, specimen, etc., which are irradiated in the High Temperature Engineering Test Reactor, since the cells should be built in limited space in the HTTR reactor building, the cells are designed considering relationship between the cells and the reactor building to utilize the limited space effectively. The cells consist of three partitioned hot cells with wall for neutron and gamma-ray shields, ventilation system including filtering units and fuel handling devices. The post-irradiation examinations of the fuels and materials are planed by using the cells and the Hot Laboratory of the Japan Materials Testing Reactor to establish the technology basis on high temperature gas-cooled reactors (HTGRs). In future, irradiation tests and post-irradiation examinations will be carried out with the cells to upgrade present HTGR technologies and to make the innovative basic research on high-temperature engineering. (author)

  10. Design of high temperature irradiation materials inspection cells. (Spent fuel inspection cells) in the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    This report summarizes design requirements and design results for shields, ventilation system and fuel handling devices for the high temperature irradiation materials inspection cells (spent fuel inspection cells). These cells are small cells to carry out few post-irradiation examinations of spent fuels, specimen, etc., which are irradiated in the High Temperature Engineering Test Reactor, since the cells should be built in limited space in the HTTR reactor building, the cells are designed considering relationship between the cells and the reactor building to utilize the limited space effectively. The cells consist of three partitioned hot cells with wall for neutron and gamma-ray shields, ventilation system including filtering units and fuel handling devices. The post-irradiation examinations of the fuels and materials are planed by using the cells and the Hot Laboratory of the Japan Materials Testing Reactor to establish the technology basis on high temperature gas-cooled reactors (HTGRs). In future, irradiation tests and post-irradiation examinations will be carried out with the cells to upgrade present HTGR technologies and to make the innovative basic research on high-temperature engineering. (author)

  11. Proceedings of the international meeting on development, fabrication, and application of Reduced Enrichment fuels for Research and Test Reactors (RERTR). Base technology

    International Nuclear Information System (INIS)

    The international effort to develop new fuel materials and designs which will make it feasible to fuel research and test reactors throughout the world with low-enrichment uranium, instead of high-enrichment uranium, has made significant progress during the past year. This progress has taken place at research centers located in many different countries, and is of crucial interest to reactor operators and licensors whose geographical distribution is even more varied. It is appropriate, therefore, that international meetings be held periodically to foster direct communication among the specialists in this area. To achieve this purpose, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the third of a series which begun in 1978. The papers presented at this meeting were divided into sessions according to relevant subject: status of RERTR program and safety issues; development of new fuel types; testing of new fuel elements; specific reactor applications. These proceedings were edited by various members of the RERTR Program

  12. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. B. Grover

    2007-05-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing.1,2 The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment

  13. A State of Art Report on Technology Development of Irradiation Test and Irradiation Behavior for High Temperature Gas-cooled Reactor Coated Particle Fuel

    International Nuclear Information System (INIS)

    Coated particle fuel, Irradiation test of coated particle fuel in foreign countries are introduced in this report. Also reviewed are status of the study on irradiation performance of coated particle fuel, irradiation facilities for irradiation test of coated particle fuel in foreign countries and current status of irradiation test plan of coated particle fuel in foreign countries' research reactor, and preliminary feasibility is analyzed the irradiation test of coated particle fuel in HANARO

  14. Test facility and instrumentation techniques for the irradiation of nuclear fuel in the INR Pitesti TRIGA Reactor to sustain the Nuclear Safety Program

    International Nuclear Information System (INIS)

    An extended program for irradiation testing of CANDU nuclear fuel in the TRIGA-SSR and ACPR reactors at INR Pitesti were performed from 1981 to 1994. The irradiation devices designed to operate mainly in the 14 MWt SSR core allow the irradiation of nuclear fuel elements and structure materials. By means of these irradiation devices there are simulated the normal operation conditions in a NPP as well as the abnormal ones. The paper describes some representative tests which yielded interesting results due to the nuclear instrumentation of irradiated samples and an outlook on future development of nuclear safety program specific for CANDU fuel testing. An appropriate analysis of the experimental results allow the evaluation of fuel behaviour, its performances and the verification of correct modelling of specific phenomena by computer codes (both in normal and accident conditions). (author). 4 figs., 5 refs

  15. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10{sup 14} ncm{sup -2} s{sup -1} and a fast flux of 6,4.10{sup 14} ncm{sup -2}s{sup -1}, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  16. Theory of elasticity of nuclear fuel element mechanics, caused by rapid change of the reactor-power

    International Nuclear Information System (INIS)

    In this report simple formulas are developed from Hooke's law by an extensive expense of mathematics. They allow to calculate, in a convenient manner, the stresses in a nuclear reactor fuel pin, caused by a rapid change of the reactor power. (orig.)

  17. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

  18. Research on Precaution and Detection Technology for Flow Blockage of Plate-type Fuel Element in Research Reactors

    Institute of Scientific and Technical Information of China (English)

    DING; Li; QIAO; Ya-xin; ZHANG; Nian-peng; LUO; Bei-bei; HUA; Xiao; JIA; Shu-jie; YAN; Hui-yang

    2013-01-01

    The main aim of this study is to offer the technical support for safety operation and management of research reactors using plate-type fuel assemblies in China,which is performed from analysis of precaution measures for flow blockage and detection methods of accidents.Study shows that most accidents were induced by in-core foreign objects and the swelling of fuel

  19. Welding procedures used in the fabrication of fuel elements for the DON Reactor exponential experiment; La soldadura en la fabricacion de elementos combustibles destinados a una experiencia exponencial

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Beltran, A.; Jaraiz Franco, E.; Rivas Diaz, M. de las

    1965-07-01

    This exponential experiment required 74 units (37 loaded with UO{sub 2} and 37 with UC) to simulate the Reactor fuel channels. Each unit was enclosed in a tube similar to the calandria ones. It contained the pressure tube, the shroud and the 19 rods cluster. Within the pressure tube, in touch with the elements, was the organic liquid. (Author)

  20. NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  1. Proceedings of the 1984 international meeting on Reduced Enrichment for Research and Test Reactors. Base technology

    International Nuclear Information System (INIS)

    More than 40 papers were presented at this RERTR Meeting during the following sessions: Status of RERTR programs and licensing procedures; LEU fuel element development; fuel fabrication and testing; economics; mixed reactor cores; and applications, i.e. neutronics and thermal hydraulics design of upgraded reactors, with new LEU fuel, fuel cycle studies, feasibility and safety analyses

  2. Neutronic study regarding transmutation fuel research at Jules Horowitz Reactor

    International Nuclear Information System (INIS)

    In order to estimate the possibilities for transmutation experiments at the Jules Horowitz Reactor several ideas for neutronic and fuel behaviour studies are investigated at CEA Cadarache. Naturally an exact replication of the burning of minor actinides in fast reactors, as expected in most transmutation scenarios, is impossible, but some key transmutation parameters can be investigated in a MTR neutron spectrum. In this paper a parametric study regarding fuel damage by He and fission products in AmUO2 is presented. By varying flux level, uranium enrichment and americium content of the sample in the JHR reflector a He production to fission ratio comparable to reference samples in the core of a SFR can be achieved. The calculations were done with the depletion code DARWIN2.2 using JEF2.2 data and spectra from a TRIPOLI model of JHR and an ERANOS model for the SFR respectively. (author)

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

    International Nuclear Information System (INIS)

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

  4. New concept of designing combined fuel for fast reactors with closing fuel cycle

    International Nuclear Information System (INIS)

    New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO2 powder of 2 powder. Framework fuel element having porous meat is fabricated by capillary impregnation method. Granules of (depleted U-Mo or U-Zr alloys) fuel and matrix (zirconium eutectic alloy) fill up fuel element cladding; fuel element is heated to a temperature 50 deg. C above the temperature of matrix melting. Matrix alloy melts down and under action of capillary forces moves into joints between fuel particles to form metallurgic bond. In this case inside the resultant framework fuel element controllable porosity (20 to 40%) is retained that subsequently accommodates Pu and MA dioxides. Zirconium matrix layer available at inner surface of fuel element cladding protects cladding from interaction with both fuel and fission products as well as from cesium induced corrosion. Properties of framework fuel element have been investigated [1]. New novel components are compatible with themselves and fuel element cladding at operating and higher temperatures. Preliminary in-pile tests of new components were carried out. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio, does not interact with fuel cladding and is more environmentally clean. Its principle advantage is a simple production process that is easily realized remotely, feasibility of involving high background Pu and MA isotopes into closed nuclear fuel cycle at the minimal influence on environment. Thus, new concept of designing fuel for fast reactors - minimization of process operations with Pu and employment of

  5. Corrosion Minimization for Research Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  6. Probabilistic-statistical analysis of WWER fuel element leaking causes and comparative analysis of the fuel reliability indicator on NPPs with WWER and PWR reactors

    International Nuclear Information System (INIS)

    The results of a comparative analysis of the fuel reliability indicator on NPPs with WWER and PWR reactors are presented in this report. As an exponent for a comparative reliability analysis of the WANO fuel reliability indicator is used. The fuel reliability indicator provides a general measure of the extent to which the reactor coolant activity is increased as a result of fuel damage. The analysis of fuel reliability indicator values during 1991-2001 at NPPs with WWER-1000 and WWER-440 reactors (Russia, Ukraine, Bulgaria, Czech Republic, Slovakia, Hungary, Finland) is carried out. The high level of WWER fuel reliability is scored except for cases of fuel failures in separate operating periods of some units. (author)

  7. Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection; Avaliacao de integridade de revestimentos de combustiveis de reatores de pesquisa e teste de materiais utilizando o ensaio de correntes parasitas

    Energy Technology Data Exchange (ETDEWEB)

    Alencar, Donizete Anderson de

    2004-07-01

    A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

  8. Proceedings of a specialist meeting on the behaviour of water reactor fuel elements under accident conditions

    International Nuclear Information System (INIS)

    The contributions of this meeting report experimental, numerical and research investigations on the oxidation behaviour of zircaloy in case of a loss-of-coolant accident (LOCA), analysis of the kinetics of the oxidation rate, very high temperature behaviour of fuel rod claddings (failure mechanics, ballooning), the interaction between cladding and fuel, the mechanical behaviour of zircaloy, etc. Numerous experimental and computer code analysis results are given

  9. PUSPATI Triga reactor fuel worth measurement

    International Nuclear Information System (INIS)

    The reactivity worth of fuel elements in the B, C, D, E and F rings in the PUSPATI TRIGA Reactor core with respect to water as well as that of dummy fuel element (graphite filled) in the G ring were measured. The reactivity worth of 8.5 w/o standard TRIGA fuel element with respect to the dummy element in the B to F rings were also determined. The measured results agreed with the typical values given by the reactor supplier, General Atomatic Company, to within eight percents. (author)

  10. Development of a mathematical model for studying rewetting of reactor fuel elements after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    The ZETHYF model allows studying the flow, heat transmission and temperature conditions in a cooling channel and, thus, recretting of reactor fuel elements. Coolant flow is calculated for a single-phase or dual-phase coolant along coolant channel on the assumption of constant pressure. Within dual-phase flow, a thermodynamically balanced slip flow is assumed for small steam volume proportions and saturation temperature is assumed for water still existing in the form of droplets. Droplet velocity is represented by means of a momentum balance between steam and droplet, evaporation of droplets is established by means of heat supply from steam, cladding tube wall and by thermal radiation. Minimum film boiling temperature is taken as a criterium for secretting and is compared with local cladding tube temperature which is established by way of a 2-D thermal conduction equation check calculation of experiments (Flecht) yielded sufficient agreement. Calculation results yielded a somewhat slower cooling of heating element and, consequently, delayed rewetting. (orig./HP)

  11. Options for the management of spent fuel from RA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Bulkin, S.Yu.; Sokolov, A.V. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Matausek, M.V.; Vukadin, Z. [VINCA Institute of Nuclear Science, Belgrade (Yugoslavia)

    2000-07-01

    Leaking fuel elements can be placed in a transportation cask or stored in a long-term dry storage facility if they are placed within an additional metal can. Two different ways of RA reactor spent fuel elements preparation for transportation or long-term storage are considered: 'all fuel elements canning without leak-tightness testing' and 'all fuel elements leak-tightness testing'. Comparison of these two options is performed according to the following criteria: the radiological influence on workers and environment; the need for non-standard equipment fabrication; the time required for work performance; the possibility for fuel elements deterioration during transportation or storage. It is believed that the first option offers several distinct advantages, which can be summarized as: greater reliability in the course of transportation or dry storage; higher safety for workers; lower expenditures for non-standard equipment manufacturing; shorter duration of work. (author)

  12. Behavior analysis of U3Si-Al fuel in MP type fuel elements under irradiation

    International Nuclear Information System (INIS)

    Uranium silicide U3Si is considered as perspective nuclear fuel for Russian research reactors. In order to resolve the problem of enrichment reduction this nuclear fuel is the most real alternative for the Uranium dioxide which is currently used for these purposes. Within RERTR program two MP type fuel element models with the core consisting of U3Si nuclear fuel dispersed in an aluminium matrix were tested in MP reactor. The tests confirmed that the use of U3Si + Al fuel composition is a perspective solution to reduce fuel element enrichment in research reactors. This report represents analysis of post-irradiation tests of the fuel element models. The goal of the analysis being to establish the value and the appropriateness of swelling for the Uranium silicide. The fuel element represents a cylinder tube with four ribs on the outer surface. The claddings are produced of CAB-6 alloy. The contents of nuclear fuel in the core constitute 34% by volume, technological pores constitute 4.5% and the rest is aluminium matrix. The nuclear fuel was produced in ARSRIIM, the fuel elements was produced by ARSRIIM specialists with equipment of NZKH. (author)

  13. Performance of AGR-1 High-Temperature Reactor Fuel During Post-Irradiation Heating Tests

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Robert Noel [ORNL; Baldwin, Charles A [ORNL; Hunn, John D [ORNL; Demkowicz, Paul [Idaho National Laboratory (INL); Reber, Edward [Idaho National Laboratory (INL)

    2014-01-01

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide TRISO fuel compacts from the AGR-1 experiment has been evaluated at temperatures of 1600 1800 C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4 to 19.1% FIMA have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 10-6 after 300 h at 1600 C or 100 h at 1800 C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 C, and 85Kr release was very low during the tests (particles with breached SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 C in one compact. Post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.

  14. Gamma spectrometry of TRIGA fuel elements

    International Nuclear Information System (INIS)

    The burnupt of 19 TRIGA fuel elements was determined by gamma spectrometry using a special fuel element holder developed and constructed at the Atom Institute, Vienna. The investigated fuel element is kept in a horizontal position about 4 m below the reactor pool water surface. A collimator tube extends to the reactor platform where an intrinsic Ge-detector is located. With this system each fuel element was investigated at eight equidistant points along its active zone and the Cs 137 activity was evaluated. (orig.)

  15. Hydrothermal Testing of K Basin Sludge and N Reactor Fuel at Sludge Treatment Project Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Thornton, Brenda M.

    2007-03-30

    The Sludge Treatment Project (STP), managed for the U. S. DOE by Fluor Hanford (FH), was created to design and operate a process to eliminate uranium metal from K Basin sludge prior to packaging for Waste Isolation Pilot Plant (WIPP). The STP process uses high temperature liquid water to accelerate the reaction, produce uranium dioxide from the uranium metal, and safely discharge the hydrogen. Under nominal process conditions, the sludge will be heated in pressurized water at 185°C for as long as 72 hours to assure the complete reaction (corrosion) of up to 0.25-inch diameter uranium metal pieces. Under contract to FH, the Pacific Northwest National Laboratory (PNNL) conducted bench-scale testing of the STP hydrothermal process in November and December 2006. Five tests (~50 ml each) were conducted in sealed, un-agitated reaction vessels under the hydrothermal conditions (e.g., 7 to 72 h at 185°C) of the STP corrosion process using radioactive sludge samples collected from the K East Basin and particles/coupons of N Reactor fuel also taken from the K Basins. The tests were designed to evaluate and understand the chemical changes that may be occurring and the effects that any changes would have on sludge rheological properties. The tests were not designed to evaluate engineering aspects of the process. The hydrothermal treatment affected the chemical and physical properties of the sludge. In each test, significant uranium compound phase changes were identified, resulting from dehydration and chemical reduction reactions. Physical properties of the sludge were significantly altered from their initial, as-settled sludge values, including, shear strength, settled density, weight percent water, and gas retention.

  16. Structural behaviour of fuel assemblies for water cooled reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    At the invitation of the Government of France and in response to a proposal of the IAEA Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT), the IAEA convened a Technical Meeting on Fuel Assembly Structural Behaviour in Cadarache, France, from 22 to 26 November 2004. The meeting was hosted by the CEA Cadarache Centre, AREVA Framatome-ANP and Electricite de France. The meeting aimed to provide in depth technical exchanges on PWR and WWER operational experience in the field of fuel assembly mechanical behaviour and the potential impact of future high burnup fuel management on fuel reliability. It addressed in-service experience and remedial solutions, loop testing experience, qualification and damage assessment methods (analytic or experimental ones), mechanical behaviour of the fuel assembly including dynamic and fluid structure interaction aspects, modelling and numerical analysis methods, and impact of the in-service evolution of the structural materials. Sixty-seven participants from 17 countries presented 30 papers in the course of four sessions. The topics covered included the impact of hydraulic loadings on fuel assembly (FA)performance, FA bow and control rod (CR) drop kinetics, vibrations and rod-to-grid wear and fretting, and, finally, evaluation and modelling of accident conditions, mainly from seismic causes. FA bow, CR drop kinetics and hydraulics are of great importance under conditions of higher fuel duties including burnup increase, thermal uprates and longer fuel cycles. Vibrations and rod-to-grid wear and fretting have been identified as a key cause of fuel failure at PWRs during the past several years. The meeting demonstrated that full-scale hydraulic tests and modelling provide sufficient information to develop remedies to increase FA skeleton resistance to hydraulic loads, including seismic ones, vibrations and wear. These proceedings are presented as a book with an attached CD-ROM. The first part of the CD

  17. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    International Nuclear Information System (INIS)

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  18. Desludging of N Reactor fuel canisters: Analysis, Test, and data requirements

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.

    1996-01-01

    The N Reactor fuel is currently stored in canisters in the K East (KE) and K West (KW) Basins. In KE, the canisters have open tops; in KW, the cans have sealed lids, but are vented to release gases. Corrosion products have formed on exposed uranium metal fuel, on carbon steel basin component surfaces, and on aluminum alloy canister surfaces. Much of the corrosion product is retained on the corroding surfaces; however, large inventories of particulates have been released. Some of the corrosion product particulates form sludge on the basin floors; some particulates are retained within the canisters. The floor sludge inventories are much greater in the KE Basin than in the KW Basin because KE Basin operated longer and its water chemistry was less controlled. Another important factor is the absence of lids on the KE canisters, allowing uranium corrosion products to escape and water-borne species, principally iron oxides, to settle in the canisters. The inventories of corrosion products, including those released as particulates inside the canisters, are only beginning to be characterized for the closed canisters in KW Basin. The dominant species in the KE floor sludge are oxides of aluminum, iron, and uranium. A large fraction of the aluminum and uranium floor sludge particulates may have been released during a major fuel segregation campaign in the 1980s, when fuel was emptied from 4990 canisters. Handling and jarring of the fuel and aluminum canisters seems likely to have released particulates from the heavily corroded surfaces. Four candidate methods are discussed for dealing with canister sludge emerged in the N Reactor fuel path forward: place fuel in multi-canister overpacks (MCOs) without desludging; drill holes in canisters and drain; drill holes in canisters and flush with water; and remove sludge and repackage the fuel.

  19. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  20. Behavior of BWR-type fuel elements with B{sub 4}C/steel absorber tested under severe fuel damage conditions in the CORA facility

    Energy Technology Data Exchange (ETDEWEB)

    Sepold, L.; Hagen, S.; Hofmann, P.; Schanz, G.

    2009-01-15

    The CORA experiments carried out in an out-of-pile facility at the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, are part of the ''Severe Fuel Damage'' (SFD) program. The experimental program was to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in a few cases up to 2400 C. In the CORA experiments two different bundle configurations were tested: PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles. The BWR-type bundles consisted of 18 fuel rod simulators (heated and unheated rods), an absorber blade of steel containing eleven absorber rods filled with boron carbide powder. The larger bundle CORA-18 contained the same number of absorber rods but was made up of 48 fuel rod simulators. All BWR bundles were surrounded by a zircaloy shroud and the absorber blades by a channel box wall on each side, also made of zircaloy. The test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence was simulated, which may develop from a small-break loss-of-coolant accident of a LWR. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1100 C, leading the bundles to maximum temperatures of approximately 2000 C. In all experiments bundle destruction started in the upper region (axially) with melting of the absorber blade and the absorber rod cladding at about 1250 C by interaction of boron carbide and steel. After destruction of the channel box walls this melt attacked the zircaloy fuel rod cladding and started to interact with the UO{sub 2} pellets. The test bundles also resulted in severe oxidation of the following components made of zircaloy: shroud, cladding, and grid spacers at the central and upper positions. Relocated absorber melt

  1. Longitudinal ridging test for ACR-1000® fuel sheathing

    Energy Technology Data Exchange (ETDEWEB)

    Bates, M.; Abbas, S.A.; Chakraborty, K.; Wang, X.; Xu, Z. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2010-07-01

    CANDU® fuel uses thin wall sheathing to reduce neutron absorption, enhance uranium utilization and provide excellent heat transfer between the sheathing and the UO{sub 2} pellets. A large diametral clearance between the fuel pellets and the sheath can result in severe longitudinal ridging under reactor operating conditions. The Advanced CANDU Reactor(ACR-1000®) fuel will also adopt thin wall sheathing. Since the ACR-1000 reference fuel element dimensions and the primary heat transport operating parameters are different compared to the CANDU-6, out-reactor tests were performed for ACR-1000 fuel covering the applicable range of parameters that have an impact on longitudinal ridge formation. This paper presents the range of parameters tested and the test method developed, which are used in predicting critical collapse pressures for longitudinal ridge formation of fuel elements with dimensions spanning both of the ACR-1000 fuel element types. The results from this testing are discussed. (author)

  2. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  3. Long-term testing of HTR fuel elements in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    The extensive results from irradiation experiments carried out on coated particles, on graphitic matrices of different composition and on integral fuel elements have shown that the spherical fuel elements with high-enriched uranium/thorium mixed-oxide particles and optimized graphitic matrix are available for use in the planned HTR facilities. A concentrated qualification programme is on the way in order to bring the fuel elements with particles from low-enriched uranium dioxide (LEU) and TRISO coating to a comparable level of experience and knowledge, i.e. to make them licensable for the planned HTR facilities. (orig.)

  4. Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (orig.)

  5. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

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

  6. Estimation of the activity and isotopic composition of the fuel elements of the reactor in decaying; Estimacion de la actividad y composicion isotopica de los elementos combustibles del reactor en decaimiento

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-03-15

    At the present time its are had 59 fuel elements, 3 control bars with follower and 2 instrumented irradiated fuels that its are decaying in the pool of the reactor. The burnt one that its have these fuels is not uniform, the quantity of U-235 that contain at the moment it varies between 33.5 g up to 35.2 and its have a decay of at least 12 years. The burnt of the fuels was obtained with the CREMAT code, this burnt was takes like base to estimate the current isotopic inventory and the activity of the same ones using the ORIGEN2 code. (Author)

  7. Report transparency and nuclear safety 2007 CEA Cadarache

    International Nuclear Information System (INIS)

    This report presents the activities of the CEA Center of Cadarache for the year 2007. The actions concerning the safety, the radiation protection, the significant events, the release control and the environmental impacts and the wastes stored on the center are discussed. More especially the report discusses the beginning of the RJH reactor construction, the fourth generation reactors research programs, the implementing of la Rotonde the new radioactive wastes management installation, the renovation of the LECA. (A.L.B.)

  8. Features of temperature control of fuel element cladding for pressurized water nuclear reactor “WWER-1000” while simulating reactor accidents

    International Nuclear Information System (INIS)

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones “cladding-junction” and “junction-coolant”. The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ∼5% of maximum value by the final moment of the stage of linear increase in the temperature

  9. Analysis of axial fuel relocation based on gamma scan data from OECD Halden Reactor Project LOCA tests

    International Nuclear Information System (INIS)

    The on-going LOCA test program IFA-650 at the OECD Halden Reactor Project (HRP) conducts in-house gamma scanning as standard post-irradiation examination (PIE) procedure on the fuel rod. One of the primary objectives of the program is to investigate fuel relocation into the balloon region. A simple model called Gamma Transport Model was formulated for purpose of interpretation of fuel relocation based on the gamma scan data. Fuel relocation may have strong effect on the linear heat generation rate at the balloon due to, firstly, increase in linear fuel density, and secondly due to differences in burn-up and local heat generation rate at the periphery and bulk of the pellet. For this analysis, a pair of isotopes with very different FP yields for U and Pu isotopes is selected from the gamma scan spectrum. The intention is to use the difference in their ratio in the balloon region to qualitatively make conclusion on the fuel relocation. As a separate outcome, the same analysis can be applied to the ejected fuel region and draw conclusion on its origin (pellet rim or bulk). The Gamma Transport Model is validated against a special case from the Halden's LOCA test program and then applied for the analysis of selected tests. (author)

  10. The DOE Advanced Gas Reactor Fuel Development and Qualification Program

    International Nuclear Information System (INIS)

    The high outlet temperatures and high thermal-energy conversion efficiency of modular High Temperature Gas-cooled Reactors (HTGRs) enable an efficient and cost effective integration of the reactor system with non-electricity generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300 C and 900 C. The Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission-product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete, fundamental understanding of the relationship between the fuel fabrication process and key fuel properties, the irradiation and accident safety performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. An overview of the program and recent progress is presented.

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

    International Nuclear Information System (INIS)

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates (mini-plates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly-enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low-uranium enrichment of about 20% 235U in place of highly enriched fuel for these reactors would reduce the potential for 235U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminium-base dispersion-type fuel plates with fuel cores of: (1) high uranium content U3O8-Al being developed by ORNL; (2) high uranium content UAl/sub x/-Al being developed by EG and G Idaho, Inc.; and (3) very high uranium content U3Si-Al being developed by ANL. The irradiation test facility, designated as HFED-1, is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. Ultimately, fuel plate types with suitable characteristics will be manufactured into full-sized plate-type fuel elements suitable for testing in the ORR. Specifications for these elements are described in Appendix A

  12. The Application of Non-Destructive Techniques in the Quality Control Inspection and Testing of Fuel Materials for the Dragon Reactor Experiment

    International Nuclear Information System (INIS)

    The O.E.C.D. high temperature gas-cooled reactor experiment. Dragon, is graphite-moderated and helium-cooled and employs fission product-retaining fuel of the coated-particle type. The development and use of this new type of fuel have required the application of a variety of non-destructive techniques for quality control, inspection and testing. This paper describes certain of the methods which are employed in the inspection of fuel materials both in the development phase and subsequently in production. Physical methods of analysis are used for control of the chemical analysis of fuel cartridges and assessment of fuel contamination of fuel-particle coatings, and particular gamma spectrometric and alpha-scintillation counting techniques are described. The general quality of coated particles is assessed and metrology of coatings performed by means of micro-radiography with the X-ray projection microscope; the procedure is outlined and typical results are presented. Fuel cartridges which consist of fuel particles in a graphite matrix are inspected for homogeneity of fuel distribution and freedom from defects by fluoroscopy and panoramic slit-scanning radiography. Colour radiography is also employed to a limited extent and the various techniques are presented together with illustrative examples of the results obtained. The fuel elements are fabricated almost entirely from various grades of graphite, and eddy-current non-destructive testing techniques are being developed to permit inspection of raw materials and finished machined components for freedom from significant defects. A brief description of these techniques is given. (author)

  13. Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core

    International Nuclear Information System (INIS)

    The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO2 and UO2/metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO2 crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process

  14. Results of recent reactor material tests on dispersal of oxide fuel from a disrupted core

    International Nuclear Information System (INIS)

    The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage tests in which molten fuel drained downward through simulated below-core structure as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO2 and UO2/metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existence of stable UO2 crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process

  15. Behavior of water reactor fuel rod

    International Nuclear Information System (INIS)

    This paper reviewed the fuels used widely in forms of (1) Zircaloy-sheathed UO2 fuel in light water-commercial power reactor, (2) Zircaloy-sheathed PuO2-UO2 fuel in plutonium-thermal reactor and advanced reactor (ATR), (3) aluminide and silicide fuel in Material Testing Reactors. From fundamental view points, physical/chemical properties and irradiation behaviors of both fuels and zircaloy claddings are briefly reviewed in chapters 1 and 2. Change of the fuel rod physical parameters with progress of burn-up are summed up in chapter 3. Some fuel troubles and failures encountered in past usage of worldwide LWR fuels are introduced with counterplans taken. In the last session of this chapter, recent results of R and D works have been carried out by fuel vendors are reviewed. Especially, in-core behaviors of PCI-remedy fuels developed to use for high burn-up extension and for load-follow operation are highlighted. Reactor accidents occurred through past forty years are surveyed and reviewed. Fuel behaviors during the reactivity initiated accident (RIA), the power-coolant mismatch (PCM), and the loss-of-coolant accident (LOCA) are taken into this review by using disclosed literatures. Safety criteria being used in Japanese licensing authorities are introduced relating to the fuel design limit. (author)

  16. Fast Reactor Fuel Development in Japan

    International Nuclear Information System (INIS)

    The future fast reactor and its fuel cycle system under development in Japan uses oxide fuel with simplified pelletizing fuel fabrication technology as a reference concept. Its driver fuel consists of large diameter annular fuel pellets, oxide dispersion strengthened ferritic steel cladding fuel pins with a ferritic-martensitic steel subassembly wrapper tube and minoractinide- bearing oxide fuel. The target burnup of the driver fuel is 150 GW.d/t in discharge average, which corresponds to 250 GW.d/t of peak burnup and 250 dpa of peak neutron dose. Fuel developmental efforts, including out-of-pile studies such as material characteristics experimental evaluation and fuel property measurements, various irradiation tests and fuel fabrication technology developments were planned and are in progress. Future fuels will be realized through Joyo irradiation tests and Monju demonstrations. International collaborative efforts are also an important part of such activities. (author)

  17. Advanced research reactor fuel development

    International Nuclear Information System (INIS)

    The fabrication technology of the U3Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U3Si2 dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U3Si2 fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 ∼ 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The γ-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U3Si2. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49

  18. Investigations of uraniumsilicide-based dispersion fuels for the use of low enrichment uranium (LEU) in research and test reactors

    International Nuclear Information System (INIS)

    The work presents at the outset, a review of the preparation and properties of uranium silicides (U3Si and U3Si2) in so far as these are relevant for their use as dispersants in research reactor fuels. The experimental work deals with the preparation and powder metallurgical processing of Al-clad miniature fuel element plates with U3Si- und U3Si-Al up to U-densities of 6.0 g U/cm3. The compatibility of these silicides with the Al-matrix under equilibrium conditions (873 K) and the influence of the reaction on the dimensional stability of the miniplates is described and discussed. (orig.)

  19. Qualification and interpretation of MR test reactor irradiation data on WWER-440 type fuel rods for fuel thermal model validation

    International Nuclear Information System (INIS)

    Conducted jointly by the Kurchatov Institute of Atomic Energy of the USSR and the Finnish state owned power utility Imatran Voima Oy, an irradiation program, SOFIT, including a series of characterised and instrumented test rod bundles is in progress to examine the thermal and mechanical behaviour of WWER type fuel rods. The main purposes of the work are to promote the flexibility and the economy of the fuel utilisation, and to verify, for the WWER annular fuel and Zr1%Nb cladding, the fuel behaviour computer programs used in the Soviet Union and in Finland. The more specific objectives are to obtain information on fuel temperature, internal gas pressure and cladding elongation during irradiation for different fabrication parameters. Qualification and interpretation of in-pile results from the first irradiation, completed at maximum local exposure of about 16 MWd/kgU, are presented. Quantitative evaluation of affecting phenomena could be deduced from the restored thermocouple readings. The measured temperatures show dependency on the as-fabricated gap size and weakness on the helium fill pressure. The results of the calculations, obtained by the existing steady state fuel behaviour computer codes are compared with the experimental temperatures. The importance of careful qualification of the data is emphasised. The related uncertainties are discussed. Preliminary fission gas data of some of the rods have become available, which show low or moderate gas release. (author). 5 refs, 7 figs, 1 tab

  20. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in unirradiated highly-enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less-than-20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. To first order, this implies the requirement that the 235U loading in the reduced-enrichment fuel elements be the same as the 235U loading in the 90-93% enriched fuel elements. This can be accomplished by substitution of higher uranium density fuel technology for currently-used fuel technology in the fuel meat volume of the current fuel element design and/or by increasing the usable fuel meat volume. For research and test reactors of power greater than 5-10 megawatts, fuel technology does not currently exist that would permit enrichment reductions to below 20% utilizing this criterion. A program is now beginning in the U.S. to develop the necessary fuel technology. Currently-proven fuel technology is capable, however, of accommodating enrichment reductions to the 30-45% range (from 90-93%) for many reactors in the 5-50MW range. Accordingly the U.S. is proposing to convert existing reactors (and new designs) in the 5-50MW range from the use of highly-enriched fuel to the use of 30-45% enriched fuel, and reactors of less that about 5MW to less-than-20% enrichment, wherever this can be done without significant performance

  1. Water reactor fuel element modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The Technical Committee Meeting on Fuel Element Modelling at High Burnup and its Experimental Support was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its subject had been touched on in many of the IAEA's activities; however for the first time modellers and experimentalists were brought together to have an exchange of views on the research under way and to identify areas where new knowledge is necessary to improve the safety, reliability and/or economics of nuclear fuel. The timely organization of this meeting in conjunction with the second meeting of the Co-ordinated Research Programme on Fuel Modelling at Extended Burnup, in short ''FUMEX'', allowed fruitful participation of representatives of developing countries which are only rarely exposed to such a scientific event. The thirty-nine papers presented covered the status of codes and experimental facilities and the main phenomena affecting the fuel during irradiation, namely: thermal fuel performance, clad corrosion and pellet-cladding interaction (PCI) and fission gas release (FGR). Refs, figs, tabs

  2. An analytical assessment of the longitudinal ridging of CANDU type fuel element

    International Nuclear Information System (INIS)

    There are 380 fuel channels in a CANDU-6 reactor, and twelve fuel bundles are loaded into each fuel channel. High-pressure, heavy water coolant passes through the fuel bundle string to remove heat generated from the fuel. Fuel sheath collapses down around the uranium dioxide pellet due to the coolant pressure when the fuel is loaded into the reactor. Longitudinal ridges may form in CANDU fuel element sheaths as a result of sheath collapse onto the pellets. A static analysis, finite-element (FE) model was developed to simulate the longitudinal ridging of the fuel element with use of the structural analysis computer code ABAQUS. Collapse pressures were calculated for the fifty-one cases for which test results of WCL in 1973 and 1975 are available. Calculation results under-predicted the critical collapse pressure but it showed significant relationship against test results

  3. Thermionic fuel element verification program—overview

    Science.gov (United States)

    Bohl, Richard J.; Dutt, Dale S.; Dahlberg, Richard C.; Wood, John T.

    1991-01-01

    TFE Verification Program is in the sixth year of a program to demonstrate the performance and lifetime of thermionic fuel elements for high power space applications. It is jointly funded by SIDO and DOE. Data from accelerated tests in FFTF and EBR-II show component lifetimes longer than 7 years. Alumina insulators have shown good performance at high fast fluence. Graphite-cesium reservoirs based on isotropic graphite also meet requirements. Three TFEs are current operating in the TRIGA reactor, the oldest having accumulated 15,000 hours of irradiation as of 1 October 1990.

  4. Thermionic fuel element Verification Program - Overview

    Science.gov (United States)

    Bohl, Richard J.; Dahlberg, Richard C.; Dutt, Dale S.; Wood, John T.

    The TFE Verification Program is in the sixth year of a program to demonstrate the performance and lifetime of thermionic fuel elements for high power space applications. Data from accelerated tests in FETF and EBR-II show component lifetimes longer than 7 yr. Alumina insulators have shown good performance at high fast fluence. Graphite-cesium reservoirs based on isotropic graphite also meet requirements. Three TFEs are currently operating in the TRIGA reactor, the oldest having accumulated 15,000 hr of irradiation as of 1 October 1990.

  5. Fuel Element Technical Manual

    Energy Technology Data Exchange (ETDEWEB)

    Burley, H.H. [ed.

    1956-08-01

    It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

  6. The Jules Horowitz Reactor: A New European Material Test Reactor (MTR) Open to International Collaboration: Update Description and Focus on Modern Safety Approach

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is a new Material Testing Reactor currently under construction at CEA Cadarache research centre in the south of France. It will represent a major Research Infrastructure for scientific studies dealing with material and fuel behaviour under irradiation (and is consequently identified for this purpose within various European road maps and forums; ESFRI, SNE-TP, etc.). The reactor will also be devoted to medical isotopes production. The reactor will perform R and D programs for the optimization of the present generation of NPP, support the development of the next generation of NPP (mainly LWR) and also offer irradiation capacities for future reactors. JHR is designed, built and will be operated as an international user-facility open to international collaboration. In order to comply with the evolution of safety requirements and to guarantee long term operations, the construction safety standards of JHR have been significantly improved compared to MTRs built in the 60s. The paper gives an up-to-date status of the construction and of the developments performed to build the future experimental capacity and is particularly focusing on the modern Safety approach used and its consequences on the design of the reactor. (author)

  7. Nuclear fuel element

    Science.gov (United States)

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

    A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

  8. Stationary Liquid Fuel Fast Reactor

    International Nuclear Information System (INIS)

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  9. Stationary Liquid Fuel Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  10. Results of the work on development of research reactor fuel element based on high density fuel with decreased enrichment in uranium-235

    International Nuclear Information System (INIS)

    The work is developed to the branch program on decrease of fuel enrichment in the Russian research reactors. The analysis of results of foreign works and own studies on creating dispersion fuel compositions on the basis of high density uranium compounds is accomplished. The uranium alloys U3Si, U3Si alloys, U-Zr-Nb, U6Fe and U6Fe alloys are chosen for application in further developments. Characteristics of these alloys are given from the viewpoint of their behaviour under irradiation

  11. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  12. Fast Reactor Fuel Development in Europe

    International Nuclear Information System (INIS)

    Research and development of minor-actinide-bearing fuels in Europe has made significant progress, with a number of scoping irradiation tests made on a number of candidate fuels foreseen for fast reactors and dedicated minor actinide transmutation systems, e.g. the accelerator driven system. Currently, efforts concentrate on uranium based fuels, as the deployment of fast reactor fleets requires Pu generation in order to achieve sustainability. Both homogeneous and heterogeneous concepts for minor actinide reactor recycling are considered. In the former, the minor actinides are added in small quantities to the mixed oxide fuel, while in the latter, the minor actinides are loaded in significant quantities in UO2. Irradiation programmes to test these concepts for pellet and SPHEREPAC fuel configurations are under way. (author)

  13. Optimization of the distribution of bars with gadolinium oxide in reactor fuel elements PWR; Optimizacion de la distribucion de barras con oxido de gadolinio en elementos combustibles para reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Melgar Santa Cecilia, P. A.; Velazquez, J.; Ahnert Iglesias, C.

    2014-07-01

    In the schemes of low leakage, currently used in the majority of PWR reactors, it makes use of absorbent consumables for the effective control of the factors of peak, the critical concentration of initial boron and the moderator temperature coefficient. One of the most used absorbing is the oxide of gadolinium, which is integrated within the fuel pickup. Occurs a process of optimization of fuel elements with oxide of gadolinium, which allows for a smaller number of configurations with a low peak factor for bar. (Author)

  14. Environmental concerns regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN

    International Nuclear Information System (INIS)

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the maim programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel - CCN - is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt% 235U), to supply its IEA-RI research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the Sustainable Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to CCN

  15. History of fast reactor fuel development

    Science.gov (United States)

    Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

    1993-09-01

    The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

  16. Failed fuel identification techniques for liquid-metal cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, J.D.B.; Gross, K.C.; Mikaili, R. [Argonne National Lab., IL (United States); Frank, S.M.; Cutforth, D.C.; Angelo, P.L. [Argonne National Lab., Idaho Falls, ID (United States)

    1995-06-01

    The Experimental Breeder Reactor II (EBR-II), located in Idaho and operated for the US Department of Energy by Argonne National Laboratory, has been used as an irradiation testbed for LMR fuels and components for thirty years. During this time many endurance tests have been carried out with experimental LMR metal, oxide, carbide and nitride fuel elements, in which cladding failures were intentionally allowed to occur. This paper describes methods that have been developed for the detection, identification and verification of fuel failures.

  17. Recovery of weapon plutonium as feed material for reactor fuel

    International Nuclear Information System (INIS)

    This report presents preliminary considerations for recovering and converting weapon plutonium from various US weapon forms into feed material for fabrication of reactor fuel elements. An ongoing DOE study addresses the disposition of excess weapon plutonium through its use as fuel for nuclear power reactors and subsequent disposal as spent fuel. The spent fuel would have characteristics similar to those of commercial power spent fuel and could be similarly disposed of in a geologic repository

  18. Recovery of weapon plutonium as feed material for reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Armantrout, G.A.; Bronson, M.A.; Choi, Jor-Shan [and others

    1994-03-16

    This report presents preliminary considerations for recovering and converting weapon plutonium from various US weapon forms into feed material for fabrication of reactor fuel elements. An ongoing DOE study addresses the disposition of excess weapon plutonium through its use as fuel for nuclear power reactors and subsequent disposal as spent fuel. The spent fuel would have characteristics similar to those of commercial power spent fuel and could be similarly disposed of in a geologic repository.

  19. Fueling of tandem mirror reactors

    International Nuclear Information System (INIS)

    This paper summarizes the fueling requirements for experimental and demonstration tandem mirror reactors (TMRs), reviews the status of conventional pellet injectors, and identifies some candidate accelerators that may be needed for fueling tandem mirror reactors. Characteristics and limitations of three types of accelerators are described; neutral beam injectors, electromagnetic rail guns, and laser beam drivers. Based on these characteristics and limitations, a computer module was developed for the Tandem Mirror Reactor Systems Code (TMRSC) to select the pellet injector/accelerator combination which most nearly satisfies the fueling requirements for a given machine design

  20. Intensive post operation clean out (Poco) in AREVA NC Cadarache dismantling projects: economic value and sustainable development

    International Nuclear Information System (INIS)

    After fabricating plutonium fuels for 40 years (Fast Breeder Reactor, MOX), commercial operations at the AREVA NC Cadarache plant ended in 2003 and post operation clean out of its production buildings and laboratory started. AREVA objective is to achieve IAEA level 2 cleanup of the process facilities and transfer them to their owner, the French Atomic Energy Commission (CEA). There are two main opposite dismantling scenarios: very limited cleanup to optimize man power, or intensive cleanup of the facilities to optimize the quantity of long lived level waste as much as possible. Economic and environmental considerations lead the choice between these two options. (author)

  1. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  2. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10⁷ cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m⁻¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were

  3. Fuel element development

    International Nuclear Information System (INIS)

    In capsule irradiation tests the influence was studied which is exerted by high power densities on thin oxide fuel rods. Cladding expansions have been observed which are not attributable to creep but to plastic strains. Power jumps during load cycling resulted in stress to the cladding through fuel pressure due to thermal differential strain. - Changes in geometry of oxide fuel pellets during cycling were investigated theoretically using models. The test group 5b was also studied with a view to plutonium redistribution. A very high plutonium enrichment was found at the central channel, and outer zones nearly free from plutonium soon after the beginning of irradiation, which might be due to the high specific power and central temperature and the high PuO2-content (35%) of the fuel. Two contributions include as subjects the porosity of fuel in the context of structural analyses and creep caused by irradiation. The plutonium content itself does not seem to increase substantially the creep rate. Further results of post-examinations are available from the oxide irradiation tests Mol-7B and DFR-435. The zone of maximum damage of the Mol-7B-rods occurs at the upper end of the fuel column; even here the structure of the rod has essentially remained unchanged. The amount of fuel escaping is not as great as at the damaged points of DFR-435. (orig.)

  4. Peach Bottom test element program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Saurwein, J.J.; Holzgraf, J.F.; MIller, C.M.; Myers, B.F.; Wallroth, C.F.

    1982-11-01

    Thirty-three test elements were irradiated in the Peach Bottom high-temperature gas-cooled reactor (HTGR) as part of the testing program for advanced HTGRs. Extensive postirradiation examinations and evaluations of 21 of these irradiation experiments were performed. The test element irradiations were simulated using HTGR design codes and data. Calculated fuel burnups, power profiles, fast neutron fluences, and temperatures were verified via destructive burnup measurements, gamma scanning, and in-pile thermocouple readings corrected for decalibration effects. Analytical techniques were developed to improve the quality of temperature predictions through feedback of nuclear measurements into thermal calculations. Dimensional measurements, pressure burst tests, diametral compression tests, ring-cutting tests, strip-cutting tests, and four-point bend tests were performed to measure residual stress, strain, and strength distributions in H-327 graphite structures irradiated in the test elements.

  5. Peach Bottom test element program. Final report

    International Nuclear Information System (INIS)

    Thirty-three test elements were irradiated in the Peach Bottom high-temperature gas-cooled reactor (HTGR) as part of the testing program for advanced HTGRs. Extensive postirradiation examinations and evaluations of 21 of these irradiation experiments were performed. The test element irradiations were simulated using HTGR design codes and data. Calculated fuel burnups, power profiles, fast neutron fluences, and temperatures were verified via destructive burnup measurements, gamma scanning, and in-pile thermocouple readings corrected for decalibration effects. Analytical techniques were developed to improve the quality of temperature predictions through feedback of nuclear measurements into thermal calculations. Dimensional measurements, pressure burst tests, diametral compression tests, ring-cutting tests, strip-cutting tests, and four-point bend tests were performed to measure residual stress, strain, and strength distributions in H-327 graphite structures irradiated in the test elements

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

    Energy Technology Data Exchange (ETDEWEB)

    Timothy A. Hyde

    2012-06-01

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

  7. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in highly enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less than 20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. A program is now beginning in the U.S. to develop the necessary fuel technology, but several years of work will be needed. Accordingly, as an immediate interim step, the U.S. is proposing to convert existing research and test reactors (and new designs) from the use of 90-93% enriched fuel to the use of 30-45% enriched fuel wherever this can be done without unacceptable reactor performance degradation

  8. Study and development of a pyrometric in-core measurement technique to follow the temperature of the fuel rod cladding; applied to the study of Loss of Coolant Accident (LOCA) during trial simulations in the Jules Horowitz Reactor (Material Testing Reactor)

    International Nuclear Information System (INIS)

    In both research and industry, temperature is a key parameter for understanding and characterizing the behavior of materials. To study the thermomechanical behavior of a fuel rod, a test device is designed for the Jules Horowitz Material Testing Reactor (currently under construction in the CEA Cadarache). The device will be placed under accidental conditions (Loss Of Coolant Accident, LOCA) causing rapid overheating. The temperature tracking, between 700 and 1200 C, will be measured by a fiber optic sensor. The aim of the project is to optimize temperature measurement by comparing different pyrometry techniques. This study covers the management of the main difficulties inherent to the design of the sensor.The first challenge consists of predicting optical fiber behavior in such complex environments where irradiation and high temperature are combined. The fiber will be exposed to a neutron dose rate about 1012 nfast/cm2/s and a dose rate of about 1 kGy/s. Moreover its extremity is heated to approximately 800 C. It is shown that under these conditions, light interferences, absorption bands and fluctuating attenuation are obstacles to overcome or to mitigate.The second challenge, concerning pyrometric measurement, comes from spectral variations expected for the rod emissivity. The material of study is chosen for its widespread use in France: Zircaloy-4. Under oxidizing conditions the spectral emissivity of this Zirconium alloy evolves. This thesis proves that between 700 and 800 C pyrometric measurement is possible from experimental point of view in laboratory without irradiation.In conclusion rod temperature tracking in JHR conditions may be possible providing that interferences are mastered and wavelengths are chosen. This work makes the use of optical pyrometry under civil nuclear extreme conditions more promising. (author)

  9. MOLTEN FLUORIDE NUCLEAR REACTOR FUEL

    Science.gov (United States)

    Barton, C.J.; Grimes, W.R.

    1960-01-01

    Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

  10. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

    2008-02-01

    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  11. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

    Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)

  12. Fuel condition in Canadian CANDU 6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, R.H.; Macici, N [Hydro-Quebec, Montreal, Quebec (Canada); Gibb, R. [New Brunswick Power, Lepreau, NB (Canada); Purdy, P.L.; Manzer, A.M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Hydro, Toronto, Ontario (Canada)

    1997-07-01

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO{sub 2} fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly

  13. Fuel condition in Canadian CANDU 6 reactors

    International Nuclear Information System (INIS)

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO2 fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly discuss our

  14. High Temperature Gas Cooled Reactor Fuels and Materials

    International Nuclear Information System (INIS)

    At the third annual meeting of the technical working group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), held in Vienna, in 2004, it was suggested 'to develop manuals/handbooks and best practice documents for use in training and education in coated particle fuel technology' in the IAEA's Programme for the year 2006-2007. In the context of supporting interested Member States, the activity to develop a handbook for use in the 'education and training' of a new generation of scientists and engineers on coated particle fuel technology was undertaken. To make aware of the role of nuclear science education and training in all Member States to enhance their capacity to develop innovative technologies for sustainable nuclear energy is of paramount importance to the IAEA Significant efforts are underway in several Member States to develop high temperature gas cooled reactors (HTGR) based on either pebble bed or prismatic designs. All these reactors are primarily fuelled by TRISO (tri iso-structural) coated particles. The aim however is to build future nuclear fuel cycles in concert with the aim of the Generation IV International Forum and includes nuclear reactor applications for process heat, hydrogen production and electricity generation. Moreover, developmental work is ongoing and focuses on the burning of weapon-grade plutonium including civil plutonium and other transuranic elements using the 'deep-burn concept' or 'inert matrix fuels', especially in HTGR systems in the form of coated particle fuels. The document will serve as the primary resource materials for 'education and training' in the area of advanced fuels forming the building blocks for future development in the interested Member States. This document broadly covers several aspects of coated particle fuel technology, namely: manufacture of coated particles, compacts and elements; design-basis; quality assurance/quality control and characterization techniques; fuel irradiations; fuel

  15. Refabricated and instrumented fuel rods

    International Nuclear Information System (INIS)

    Nuclear Fuel for power reactors capabilities evaluation is strongly based on the intimate knowledge of its behaviour under irradiation. This knowledge can be acquired from refabricated and instrumented fuel rods irradiated at different levels in commercial reactors. This paper presents the development and qualification of a new technique called RECTO related to a double-instrumented rod re-fabrication process developed by CEA/LECA hot laboratory facility at CADARACHE. The technique development includes manufacturing of the properly dimensioned cavity in the fuel pellet stack to house the thermocouple and the use of a newly designed pressure transducer. An analytic irradiation of such a double-instrumented fuel rod will be performed in OSIRIS test reactor starting October 2004. (Author)

  16. Thermal-hydraulic design calculations for the annular fuel element with replaceable test bundles (TOAST) on the test zone position 205 of KNK II/3

    International Nuclear Information System (INIS)

    Annular fuel elements are foreseen in KNK II as carrier elements for irradiation inserts and test bundles. For the third core a reloadable annular element on position 205 is foreseen, in which replaceable 19-pin test bundles (TOAST) shall be irradiated. The present report deals with the thermal-hydraulic design of the annular carrier element and the test bundle, whereby the test bundle required additional optimization. The code CIA has been used for the calculations. Start of irradiation of the subassembly is planned at the beginning of the third core operation. After optimization of the pin-spacer geometry in the test bundle, design calculations for both bundles were performed, whereby thermal coupling between both was taken into account. The calculated mass-flows and temperature distributions are given for the nominal and the eccentric element configuration. The calculated bundle pressure losses have been corrected according to experimental results

  17. Manufacture of uranium compounds for research reactors fuel elements. Participation of the UCPP (Uranium compound production plant) in the Egyptian project

    International Nuclear Information System (INIS)

    UCPP is an international qualified supplier of U3O8 with up to 20 % enrichment in U-235. The characteristics of this powder are those specified for fuel plates manufacture for test reactors. This paper describes the works performed in the plant since its beginning, emphasising those undertaken during the last years. The transference of U3O8 manufacturing technology to INVAP SE, the enterprise that installed a plant of similar characteristics in the Arabian Republic of Egypt, is especially described. (author)

  18. Failure of triga fuel cladding at the Berkeley Research Reactor

    International Nuclear Information System (INIS)

    On September 16, 1985, following a long maintenance shutdown, unusually high concentrations of radioisotopes were detected in the reactor-room air on a Constant Air Monitor (CAM) after two and a half hours of full power operation. It was thought that the activity could be coming from some contamination in the pool water. Thus the water was cleaned and the water conductivity was reduced fourfold. However, a full-power operation again showed high count rates on the CAM. A third test was conducted with a germanium detector. Following two hours of operation, three fission-product gasses were identified in the reactor-room air; Kr85, Kr37 and Kr88. Once again no unusual activities could be detected on the CAM filter, in the pool water, or in the demineralizer resins. It was concluded that the gasses must be coming from a leaking fuel element. Three old, instrumented elements with defective thermocouples were selected to be the first ones isolated from the core. After removing the elements, the reactor was operated at full-power for two hours with no abnormal activities detected. New standard elements were loaded and the reactor was again operated at full-power to confirm that no leaking element remained in the core. Since then, the reactor has been operated, with no abnormal activities detected. (Nogami, K.)

  19. Method and apparatus for increasing fuel efficiency in nuclear reactors

    International Nuclear Information System (INIS)

    This patent describes an improved method of producing a spectral shift in a nuclear reactor to achieve increased nuclear fuel efficiency, the nuclear reactor containing a fluid moderator juxtaposed with fuel elements containing the nuclear fuel, which comprises disposing within the fluid moderator stationary non-poison displacer rods for achieving the spectral shift, the displacer rods exhibiting a continuous reduction in volume during operation of the nuclear reactor whereby the fluid moderator increases in volume as the nuclear fuel is burned in the nuclear reactor

  20. TJ-II operation tracking from Cadarache

    International Nuclear Information System (INIS)

    The TJ-II remote participation system was designed to follow the TJ-II discharge production, even allowing the physicist in charge of operation to be in a remote location. The system has been based on both web servers and Java technology. These elements were chosen due to its open character, security properties, platform independence and technological maturity. Web pages and Java applications permit users to access experimental systems, data servers and the operation logbook. Security resources are provided by the PAPI system, a distributed authentication and authorization system. The TJ-II remote participation tools have allowed us to command and follow the stellarator operation from Cadarache. Over 1,000 digitizer channels and more than 20 diagnostic control systems were remotely available from web pages for monitoring/programming purposes. One Java application provided on-line information about the acquisition status of channels and acquisition cards. A second Java application showed temporal evolution signals that were refreshed in an automated way on the screen after each shot. A third Java application provided access to the operation logbook. In addition to these tools, we used the VRVS video conferencing system (FUSION community, X-Point room) and the EFDA Messenger Service for instant messaging (Jabber client). (Author)

  1. TJ-II operation tracking from Cadarache

    International Nuclear Information System (INIS)

    The TJ-II remote participation system was designed to follow the TJ-II discharge production, even allowing the physicist in charge of operation to be in a remote location. The system has been based on both web servers and Java technology. These elements were chosen due to its open character, security properties, platform independency and technological maturity. Web pages and Java applications permit users to access experimental systems, data servers and the operation logbook. Security resources are provided by the PAPI system, a distributed authentication and authorization system. The TJ-II remote participation tools have allowed us to command and follow the stellarator operation from Cadarache. Over 1,000 digitizer channels and more than 20 diagnostic control systems were remotely available from web pages for monitoring/programming purposes. One Java application provided on-line information about the acquisition status of channels and acquisition cards. A second Java application showed temporal evolution signals that were refreshed in an automated way on the screen after each shot. A third Java application provided access to the operation logbook. In addition to these tools, we used the VRVS videoconferencing system (FUSION community, X-Point room) and the EFDA Messenger Service for instant messaging (Jabber client). (author)

  2. Final report on the irradiation testing and post-irradiation examination of low enriched U3O8-Al and UAlx-Al fuel elements by the Netherlands Energy Research Foundation (ECN)

    International Nuclear Information System (INIS)

    Within the framework of the RERTR-programme four low-enriched (20%) MTR-type fuel elements have been irradiated in the High Flux Reactor at Petten (The Netherlands) and subjected to post-irradiation examination. Two of the elements contain UAlx-Al and two contain U3O8-Al fuel. The test irradiation has been completed up to the target burn-up values of 50% and 75% respectively. An extensive surveillance programme carried out during the test period has confirmed the excellent in-reactor behaviour of both types. Post-irradiation examination of the test elements, comprising of dimensional measurements, burn-up determination, fuel metallography and blister testing, has confirmed the irradiation experiences. Good agreement between calculated and measured power and burn-up characteristics has been found. A survey of the test element characteristics, their irradiation history, the irradiation tests and the PIE results is given in this report. (author). 5 refs, 31 figs, 3 tabs

  3. Experience with spent fuel storage at research and test reactors. Proceedings of an advisory group meeting held in Vienna, 5-8 July 1993

    International Nuclear Information System (INIS)

    Irradiated fuel from research and test reactors has been stored at various facilities for several decades. As these facilities age and approach or exceed their original design lifetimes, there is mounting concern about closure of the fuel cycle and about the integrity of ageing fuels from the materials point of view as well as some concern about the loss of self-protection of the fuels as their activity decays. It is clear that an international effort is necessary to give these problems sufficient exposure and to ensure that work begins on appropriate solutions. To obtain an overall picture of the size and extent of these problems, an Advisory Group Meeting on Storage Experience with Spent Fuel from Research Reactors was convened in Vienna 5-8 July 1993, and attended by twelve participants and three observers representing thirteen different countries. These proceedings contain the country reports presented at the meeting. Refs, figs and tabs

  4. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  5. Advanced Multiphysics Modeling of Fast Reactor Fuel Behavior

    International Nuclear Information System (INIS)

    Evaluation of fast reactor fuel thermo-mechanical performance using fuel performance codes is a key aspect of advanced fast reactors designs. Those fuel performance codes capture the multiphysics nature of fuel behavior during irradiation where different, mostly interdependent, phenomena are taking place. Existing fuel performance codes do not fully capture those interdependencies and present the different phenomena through de-coupled models. Recent developments in multiphysics simulation capabilities and availability of advanced computing platforms led to advancements in simulation of nuclear fuel behavior. This paper presents current experiences in applying different multiphysics simulation platforms to evaluation of fast reactors metallic fuel behavior. Full 3D finite element simulation platforms that include capabilities to fully couple key fuel behavior models are discussed. Issues associated with coupling metallic fuels phenomena, such as fission gas models and constituent distribution models, with thermo-mechanical finite element platforms, as well as different coupling schemes are also discussed. (author)

  6. Status of spent fuels in Japanese research reactors

    International Nuclear Information System (INIS)

    There are now eleven research and test reactors in operation in Japan. Spent fuel issues might cause problems at the JRR-3M and JMTR reactors in the near future. Increasing the number of spent fuel racks at these reactors is now under consideration because the existing capacity is almost filled. The commissioning of extra racks will allow space for the normal discharge of fuel from these reactors for several more years. The current management of spent fuel from the eleven operational reactors is suitable to meet their needs. (author). 3 tabs

  7. Transportation of irradiated fuel elements

    International Nuclear Information System (INIS)

    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.)

  8. Analysis of irradiation-induced stresses in coating layers of coated fuel particles for the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Irradiation-induced stresses in coating layers of coated fuel particles were analyzed by the MICROS-2 code for the fuels of the High Temperature Engineering Test Reactor (HTTR) under its operating conditions. The analyses were made on the standard core fuel (A-type) and the test fuels comprising the advanced SiC-coated particle fuel (B-1 type) and the ZrC-coated particle fuel (B-2 type). For the B-1 type fuel, the stresses were relieved due to the thicker buffer and SiC layers than for the A type fuel. The slightly decreased thickness of the fourth layer for the B-1 type than for the A type fuel had no significant effect on the stresses. As for the B-2 type fuel, almost the same results as for the B-1 type were obtained under an assumption that the ZrC layer as well as the SiC layer undergoes negligible dimension change within the analysis conditions. The obtained results indicated that the B-1 and B-2 type fuels are better than the A type fuel in terms of integrity against the irradiation-induced stresses. Finally, research subjects for development of the analysis code on the fuel behavior are discussed. (author)

  9. Research reactor fuel management in the Czech Republic

    International Nuclear Information System (INIS)

    Fuel management of the Czech research reactors is described. There are three research reactors in the Czech Republic: LVR-15 and LR-0 operated by the Nuclear Research Institute Rez plc, VR-1 operated by the Czech Technical University, Faculty of Nuclear Sciences and Physical Engineering in Prague, and SR-0 reactor of SKODA JS plc which is under decommissioning now. The paper describes the major features of the Czech research reactors, types of fuels used in them, and the spent fuel management principles. The participation of the LVR-15 and VR-1 reactors in the RERTR international programme (Reduced Enrichment for Research and Test Reactors) is also highlighted. (author)

  10. Speech by Prime Minister Francois Fillon. Visit of the Jules Horowitz experimental reactor works on the Commissariat a l'Energie et aux Energies Alternatives site. Cadarache, May 3, 2010

    International Nuclear Information System (INIS)

    In this speech, the French Prime Minister evokes the present context, the importance of strategic technologies, and the challenge of investing in these technologies within a context of reduction of public expenses. He comments the decision of his government to finance research and education activities in different domains, and more specifically in the energy sector with this fourth generation Jules Horowitz experimental reactor. He recalls that the nuclear sector has always been very important to the eyes of the successive French governments, and outlines how this reactor will contribute to reactor operational optimization, lifetime extension and safety, nuclear fuel development, etc.

  11. A New Fuel Design for Two Different HW Type Reactors

    Directory of Open Access Journals (Sweden)

    Daniel O. Brasnarof

    2011-01-01

    Full Text Available A new fuel element (called CARA designed for two different heavy water reactors (HWRs is presented. CARA could match fuel requirements of both (one CANDU and one unique Siemens's design Argentine HW reactors. It keeps the heavier fuel mass density and hydraulic flow restriction in both reactors together with improving both thermomechanic and thermalhydraulic, safety margins of present fuels. In addition, the CARA design could be considered as another design line for the next generation of CANDU fuels intended for higher burnup.

  12. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    Science.gov (United States)

    Makmal, T.; Aviv, O.; Gilad, E.

    2016-10-01

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections.

  13. Fabrication of particulate metal fuel for fast burner reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Lee, Sun Yong; Kim, Jong Hwan; Woo, Yoon Myung; Ko, Young Mo; Kim, Ki Hwan; Park, Jong Man; Lee, Chan Bok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

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

  14. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

  15. Spacer for fuel rods in nuclear fuel elements

    International Nuclear Information System (INIS)

    Spacers for fuel rods in nuclear reactor fuel elements are described, especially for use aboard ships. Spacers are used in a grid formed by web plates orthogonally intersecting and assembled together in a tooth-comb fashion forming a plurality of channels. The web plates are joined together and each of the web plates includes apertures through which resilient and separator members are joined. The resilient and separator members are joined. The resilient and separator members are in adjacent channels and with other similar members in the same channel, contact a fuel rod in the channel. The contact pressure between the members and fuel rod is radially directed

  16. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Center for Nuclear Engineering has shown expertise in the field of nuclear and energy systems ad correlated areas. Due to the experience obtained over decades in research and technological development at Brazilian Nuclear Program personnel has been trained and started to actively participate in the design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in the production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. The Nuclear Fuel Center is responsible for the production of the nuclear fuel necessary for the continuous operation of the IEA-R1 research reactor. Development of new fuel technologies is also a permanent concern

  17. Automated Fuel Element Closure Welding System

    International Nuclear Information System (INIS)

    The Automated Fuel Element Closure Welding System is a robotic device that will load and weld top end plugs onto nuclear fuel elements in a highly radioactive and inert gas environment. The system was developed at Argonne National Laboratory-West as part of the Fuel Cycle Demonstration. The welding system performs four main functions, it (1) injects a small amount of a xenon/krypton gas mixture into specific fuel elements, and (2) loads tiny end plugs into the tops of fuel element jackets, and (3) welds the end plugs to the element jackets, and (4) performs a dimensional inspection of the pre- and post-welded fuel elements. The system components are modular to facilitate remote replacement of failed parts. The entire system can be operated remotely in manual, semi-automatic, or fully automatic modes using a computer control system. The welding system is currently undergoing software testing and functional checkout

  18. Extended Storage for Research and Test Reactor Spent Fuel for 2006 and Beyond

    Energy Technology Data Exchange (ETDEWEB)

    Hurt, William Lon; Moore, K.M.; Shaber, Eric Lee; Mizia, Ronald Eugene

    1999-10-01

    This paper will examine issues associated with extended storage of a variety of spent nuclear fuels. Recent experiences at the Idaho National Engineering and Environmental Laboratory and Hanford sites will be described. Particular attention will be given to storage of damaged or degraded fuel. The first section will address a survey of corrosion experience regarding wet storage of spent nuclear fuel. The second section will examine issues associated with movement from wet to dry storage. This paper also examines technology development needs to support storage and ultimate disposition.

  19. Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

    1998-10-01

    The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

  20. Burn-up and Operation Time of Fuel Elements Produced in IPEN

    Science.gov (United States)

    Tondin, Julio Benedito Marin; Filho, Tufic Madi

    2011-08-01

    The aim of this paper is to present the developed work along the operational and reliability tests of fuel elements produced in the Institute of Energetic and Nuclear Research, IPEN-CNEN/SP, from the 1980's. The study analyzed the U-235 burn evolution and the element remain in the research reactor IEA-R1. The fuel elements are of the type MTR (Material Testing Reactor), the standard with 18 plates and a 12-plate control, with a nominal mean enrichment of 20%.

  1. KNK II third core: design report for the annular fuel elements on the central position to accommodate material test inserts NZ 402 and NZ 403

    International Nuclear Information System (INIS)

    Since August 1984 irradiation experiments with temperature controlled pressure tube probes are being performed in the central position of KNK II. This is part of a long-term experimental program for the development of irradiation resistant reactor materials, which shall also be continued in the third core. The necessary irradiation channel is provided by a special annular fuel element. The present report describes the annular fuel elements for the third core. Aspects of the subassembly design are considered on the basis of the annular element design for the second core and the standard elements of the third core. Two annular elements NZ 402 and NZ 403 (as reserve) are available. It is demonstrated that the expected loadings will allow an unperturbed operation of the annular elements on the central position of the third core

  2. In-pile steam oxidation of model HTGR fuel elements

    International Nuclear Information System (INIS)

    Model HTGR fuel elements were exposed to various concentrations of steam while being irradiated under several sets of temperature conditions in the Oak Ridge Research Reactor. In one test, catalysis by iron impurities in the graphite casing of the fuel element caused a highly localized attack on the graphite by the steam; this resulted in the formation of deep pits in the casing. Furthermore, the iron impurities were sufficiently mobile to cause pitting attack on the pyrolytic carbon coatings of the fuel particles as well. The presence of steam induced a rapid increase in the release of gaseous fission products. However, the cessation of steam ingress in the primary system resulted in a pronounced, but correspondingly smaller, reduction in the level of gaseous release. The incidence of fuel failure was greater than anticipated; however, even though the coatings of greater than 30% of the fuel had failed, the release of fission products beyond the fuel element itself was largely confined to iodine and the noble gases. A novel mode of fuel failure was observed under the rather severe conditions of the tests; this involved the attack of the pyrolytic carbon coatings on intact particles by uncoated fragments of uranium fuel kernel material from failed particles

  3. Gamma-ray spectroscopy on irradiated MTR fuel elements

    Science.gov (United States)

    Terremoto, L. A. A.; Zeituni, C. A.; Perrotta, J. A.; da Silva, J. E. R.

    2000-08-01

    The availability of burnup data is an important requirement in any systematic approach to the enhancement of safety, economics and performance of a nuclear research reactor. This work presents the theory and experimental techniques applied to determine, by means of nondestructive gamma-ray spectroscopy, the burnup of Material Testing Reactor (MTR) fuel elements irradiated in the IEA-R1 research reactor. Burnup measurements, based on analysis of spectra that result from collimation and detection of gamma-rays emitted in the decay of radioactive fission products, were performed at the reactor pool area. The measuring system consists of a high-purity germanium (HPGe) detector together with suitable fast electronics and an on-line microcomputer data acquisition module. In order to achieve absolute burnup values, the detection set (collimator tube+HPGe detector) was previously calibrated in efficiency. The obtained burnup values are compared with ones provided by reactor physics calculations, for three kinds of MTR fuel elements with different cooling times, initial enrichment grades and total number of fuel plates. Both values show good agreement within the experimental error limits.

  4. Particle fuel bed tests

    International Nuclear Information System (INIS)

    Gas-cooled reactors, using packed beds of small diameter coated fuel particles have been proposed for compact, high-power systems. The particulate fuel used in the tests was 800 microns in diameter, consisting of a thoria kernel coated with 200 microns of pyrocarbon. Typically, the bed of fuel particles was contained in a ceramic cylinder with porous metallic frits at each end. A dc voltage was applied to the metallic frits and the resulting electric current heated the bed. Heat was removed by passing coolant (helium or hydrogen) through the bed. Candidate frit materials, rhenium, nickel, zirconium carbide, and zirconium oxide were unaffected, while tungsten and tungsten-rhenium lost weight and strength. Zirconium-carbide particles were tested at 2000 K in H2 for 12 hours with no visible reaction or weight loss

  5. Preliminary Nuclear Analysis for the HANARO Fuel Element with Burnable Absorber

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, So Young; In, Won Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Burnable absorber is used for reducing reactivity swing and power peaking in high performance research reactors. Development of the HANARO fuel element with burnable absorber was started in the U-Mo fuel development program at HANARO, but detailed full core analysis was not performed because the current HANARO fuel management system is uncertain to analysis the HANARO core with burnable absorber. A sophisticated reactor physics system is required to analysis the core. The McCARD code was selected and the detailed McCARD core models, in which the basic HANARO core model was developed by one of the McCARD developers, are used in this study. The development of nuclear fuel requires a long time and correct developing direction especially by the nuclear analysis. This paper presents a preliminary nuclear analysis to promote the fuel development. Based on the developed fuel, the further nuclear analysis will improve reactor performance and safety. Basic nuclear analysis for the HANARO and the AHR were performed for getting the proper fuel elements with burnable absorber. Addition of 0.3 - 0.4% Cd to the fuel meat is promising for the current HANARO fuel element. Small addition of burnable absorber may not change any fuel characteristics of the HANARO fuel element, but various basic tests and irradiation tests at the HANARO core are required.

  6. Safety analysis report for packaging: the ORNL HFIR spent-fuel-element shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Eversole, R.E.; Just, R.A.; Llewellyn, G.H.

    1977-11-01

    The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) spent-fuel-element shipping cask is used to transport HFIR, Oak Ridge Research Reactor (ORR), and other reactor fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures and tests were used to determine behavior of the cask relative to the general standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations.

  7. Searching for a possible fuel element leak

    International Nuclear Information System (INIS)

    A gamma spectrum analysis of a filter paper from an Oregon State University TRIGA Reactor (OSTR) continuous air monitor (CAM) which routinely monitors the air directly over the reactor tank revealed just-detectable levels of several short-lived particulate fission products typically associated with a fuel cladding failure. This prompted an intensive.search to determine the origin of these radionuclides. A number of methods were used, including a fuel element rotation program designed to ultimately remove all of the fuel elements from the core in groups of three, and a scheme to selectively sample bubbles from different parts of the core during operation. Determination of the source was made very difficult by the fact that its presence was erratic in nature and because radioactivity levels found on filter papers were on the border of detectability even when the reactor was operated at the maximum allowable power level of 1MW. The origin and source of the fission product activity was not found, no other abnormality was identified and the reactor was therefore returned to normal operation. In addition to continuing the routine operation of the reactor-top CAM, further surveillance designed to detect a positive reappearance of the source was also implemented and currently involves a complete gamma spectrum analysis of a CAM filter paper each week after a standard (controlled) 3 hour reactor run at 1 MW. (author)

  8. Air oxidation behavior of fuel for the high temperature engineering test reactor (HTTR)

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Hironobu; Hayashi, Kimio; Fukuda, Kousaku (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)

    1992-08-01

    The oxidation behavior of the HTTR fuel was studied with respect to the scenario of an air ingress accident which had been assessed in the HTTR safety analysis. The coated fuel particles were heated under a sufficient air flow in the temperature range of 900-1400degC for maximum duration of 600 h (at 1300degC). Failure fractions of the SiC coating layer after the heat treatments remained within the fraction at the fuel production. And the failure behavior of the SiC layer did not depend on such heating conditions as the temperature and the duration in the present experiment. It was confirmed by scanning electron microscopy (SEM), X-ray diffraction and laser Raman spectroscopy that a thin oxide film was formed on the SiC layer by the heat treatments. (author).

  9. Air oxidation behavior of fuel for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    The oxidation behavior of the HTTR fuel was studied with respect to the scenario of an air ingress accident which had been assessed in the HTTR safety analysis. The coated fuel particles were heated under a sufficient air flow in the temperature range of 900-1400degC for maximum duration of 600 h (at 1300degC). Failure fractions of the SiC coating layer after the heat treatments remained within the fraction at the fuel production. And the failure behavior of the SiC layer did not depend on such heating conditions as the temperature and the duration in the present experiment. It was confirmed by scanning electron microscopy (SEM), X-ray diffraction and laser Raman spectroscopy that a thin oxide film was formed on the SiC layer by the heat treatments. (author)

  10. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    International Nuclear Information System (INIS)

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% 235U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  11. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  12. Detection of fuel element vibration at KNK II

    International Nuclear Information System (INIS)

    The reactivity signal of the KNK-II-plant shows almost harmonic oscillations of δrho <= 0.5 c. Very sensitive correlation measurements, made during the regular plant operation with the normal plant instrumentation, revealed, that these oscillations are associated with individual fuel elements. Auxiliary measurements under various operational conditions and theoretical considerations show, that this phenomenon is probably caused by flow-induced mechanical vibration. Similar characteristics with respect to the frequencies have obviously not yet been observed for fuel element vibration during tests in out-of-core loops and in other reactors. Therefore efforts have been made in order to classify the flow-induced vibration and to identify the particular excitation mechanism. Most likely seems a flow-induced vibration of whole fuel elements by vortex shedding or jet switching. This model can explain all observations without exception. (orig.)

  13. TRIGA - LEU cluster with 36 fuel elements

    International Nuclear Information System (INIS)

    Designing the TRIGA - LEU fuel cluster is part of the mechanical design of TRIGA reactor core. The latter is supported by a square frame (11 x 12 132 meshes) accommodating the 35 fuel clusters. The TRIGA fuel cluster is designed to incorporate 36 fuel elements with 3/8 inch diameter allowing the pins to be arranged into a 6 x 6 matrix. The final mechanical design of reactor zone resulted into a cluster of squared cross section with 87.5 mm side and 88.9 mm separation between the centers of the clusters. This cluster was designed by preserving the dimensions and configuration of fuel clusters with 25 elements. By the positioning of the pins inside the cluster one obtains: - a fuel element protection by reducing the failure risks; - delimitation of fixed channel of the cooling flow for each cluster; - a convenient means of manipulation; - a correct water flow for cooling the pins in a fixed channel by preserving the surface of cooling channels from the 25 fuel element cluster. The cluster has the following principal components: - casing; - bottom plug or adapter; - upper plug for maneuvering; - spacer for fuel elements. The cluster casing is made of aluminium with square cross section of 87.5 mm side and is provided at the lower part with an aluminium adapter allowing its insertion in the reactor core frame. This piece is designed to support the ends of the 36 fuel elements in a blocked position. The fuel elements are subject to asymmetric temperature distribution flux conditions, hence an asymmetric temperature distribution results concomitantly with a symmetrical (about 0.8 mm) swelling of the Incoloy 800 can. Also bending of the fuel element occurs which will be limited by the intermediate spacer. At the casing upper part an aluminium upper plug or handle is mounted allowing cluster maneuvering by means of a special tool. The cluster is provided with lateral holes in its upper part ensuring the necessary cooling water flow in case the upper part of the cluster

  14. Polarly anisotropic thermoelasticity of cylindrical and spherical fuel elements

    International Nuclear Information System (INIS)

    This paper deals with the solution for principal thermally induced stress in log solid and hollow rods and balls, taking onto account not only surface pressure load and internal heat generation, but also non equal elastic parameters and thermal strain. Closed form solutions obtained for circumferentially reinforced bodies are complementary to recently published formulae for transversely isotropic thermoelasticity of cylinders and spheres. Numerical test is gi ven using typical data for the ceramic fuel of roll and ball shape elements for high temperature gas cooled nuclear reactors, but similar values develop in the pressurised light water reactors. (author)

  15. Drying results of K-Basin fuel element 0309M (Run 3)

    International Nuclear Information System (INIS)

    An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-West Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from K-West canister 0309M during the second fuel selection campaign, conducted in 1996, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. The fuel element was broken in two pieces, with a relatively clean fracture, and the larger piece was tested. A gray/white coating was observed. This was the first test of a damaged fuel element in the furnace. K-West canisters can hold up to seven complete fuel assemblies, but, for purposes of this report, the element tested here is designated as Element 0309M. Element 0309M was subjected to drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  17. Fine lattice stochastic modeling of particle fuels in HTGR fuel elements

    International Nuclear Information System (INIS)

    There is growing interest worldwide in high temperature gas-cooled reactors (HTGRs) as candidates for next generation reactor systems. Either in a pebble type or in a prismatic type HTGR, coated particle fuel (TRISO fuel) appears to be the most promising fuel candidate to be used. For design and analysis of such a reactor, transport models, in particular, stochastic models that permit the simulation of neutron transport through the stochastic mixture of fuel and moderator materials, are becoming essential and gaining importance. Naturally, the Monte Carlo methods have been used for this situation. However, the methods reported in the literature all have their own deficiencies. In this thesis, we propose a new Monte Carlo method named fine lattice stochastic (FLS) modeling that is distinct from others. This method is based on fine lattice system in which a lattice circumscribes a fuel particle. Once the problem is given, an interface Fortran code gives out the TRISO particle fuel configurations (a set of lattice center points only) for MCNP input. The number of available lattice center points is far larger than the number of fuel particles according to packing fraction of the fuel element. We apply discrete random sampling here to choose a certain number of lattices to fill with fuel particles. In this aspect, FLS modeling allows more realistic fuel particle distributions. In this thesis, only simple cube (SC) structure is used in cubic lattice. However, FLS model can be easily extended to BCC, FCC structures or hexagonal prism type lattice. The criticality calculations for our FLS modeling were first tested on a small cube problem and compared with other models. The results indicate that the new stochastic model is an accurate and efficient approach to analyze TRISO particle fuel configurations. Then the FLS modeling was performed to analyze HTGR fuel elements for both pebble type and prismatic type and the results were also good as expected

  18. Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.

    Energy Technology Data Exchange (ETDEWEB)

    Garner, P. L.; Hanan, N. A. (Nuclear Engineering Division)

    2011-06-07

    The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

  19. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    International Nuclear Information System (INIS)

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  20. Drying Results of K-Basin Fuel Element 2660M (Run 7)

    International Nuclear Information System (INIS)

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the seventh of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 2660M. This element (referred to as Element 2660M) was stored underwater in the K-West Basin from 1983 until 1996. Element 2660M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

  1. Drying Results of K-Basin Fuel Element 6513U (Run 8)

    International Nuclear Information System (INIS)

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL)on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the eighth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 6513U. This element (referred to as Element 6513U) was stored underwater in the K-West Basin from 1983 until 1996. Element 6513U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0 and discussed in Section 6.0

  2. Radionuclide release from research reactor spent fuel

    International Nuclear Information System (INIS)

    ruled out. Solution concentrations of U were within the range of the solubility limits of the solid phase U(OH)4(am). The determined concentrations of U and Am in solution were about one order of magnitude higher for the U3Si2-Al fuel sample. Here, the formation of U/Si containing secondary phase components and their influence on radionuclide solubility cannot be ruled out. Results of this work show that the U3Si2-Al and UAlx-Al dispersed research reactor spent fuel samples dissolved completely within the test period of 3.5 years in MgCl2-rich brine in the presence of Fe2+. In view of final disposal this means that these fuel matrices represent no barrier. The radionuclides will be released instantaneously. Cs (the long-lived isotope 135Cs is of special concern with respect to final disposal) and Sr were classified as mobile radionuclide species. For U, Am, Pu and Eu, a reimmobilization was observed. Sorption is the process which is assumed to be responsible for the reimmobilization of the long-lived actinide Am and the lanthanide Eu. Solution concentrations of U and Pu seem to be controlled by their solubility controlling solid phases.

  3. Measurement of photon flux with a miniature gas ionization chamber in a Material Testing Reactor

    Science.gov (United States)

    Fourmentel, D.; Filliatre, P.; Villard, J. F.; Lyoussi, A.; Reynard-Carette, C.; Carcreff, H.

    2013-10-01

    Nuclear heating measurements in Material Testing Reactors (MTR) are crucial for the design of the experimental devices and the prediction of the temperature of the hosted samples. Nuclear heating in MTR materials (except fuel) is mainly due to the energy deposition by the photon flux. Therefore, the photon flux is a key input parameter for the computer codes which simulate nuclear heating and temperature reached by samples/devices under irradiation. In the Jules Horowitz MTR under construction at the CEA Cadarache, the maximal expected nuclear heating levels will be about 15 to 18 W g-1 and it will be necessary to assess this parameter with the best accuracy. An experiment was performed at the OSIRIS reactor to combine neutron flux, photon flux and nuclear heating measurements to improve the knowledge of the nuclear heating in MTR. There are few appropriate sensors for selective measurement of the photon flux in MTR even if studies and developments are ongoing. An experiment, called CARMEN-1, was conducted at the OSIRIS MTR and we used in particular a gas ionization chamber based on miniature fission chamber design to measure the photon flux. In this paper, we detail Monte-Carlo simulations to analyze the photon fluxes with ionization chamber measurements and we compare the photon flux calculations to the nuclear heating measurements. These results show a good accordance between photon flux measurements and nuclear heating measurement and allow improving the knowledge of these parameters.

  4. Measurement of photon flux with a miniature gas ionization chamber in a Material Testing Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fourmentel, D., E-mail: damien.fourmentel@cea.fr [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Filliatre, P.; Villard, J.F.; Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Reynard-Carette, C. [Aix-Marseille Université, LISA EA 4672, cedex 20, Marseille 13397 (France); Carcreff, H. [CEA, DEN, DRSN, Saclay, F-91191 Gif-sur-Yvette (France)

    2013-10-01

    Nuclear heating measurements in Material Testing Reactors (MTR) are crucial for the design of the experimental devices and the prediction of the temperature of the hosted samples. Nuclear heating in MTR materials (except fuel) is mainly due to the energy deposition by the photon flux. Therefore, the photon flux is a key input parameter for the computer codes which simulate nuclear heating and temperature reached by samples/devices under irradiation. In the Jules Horowitz MTR under construction at the CEA Cadarache, the maximal expected nuclear heating levels will be about 15 to 18 W g{sup −1} and it will be necessary to assess this parameter with the best accuracy. An experiment was performed at the OSIRIS reactor to combine neutron flux, photon flux and nuclear heating measurements to improve the knowledge of the nuclear heating in MTR. There are few appropriate sensors for selective measurement of the photon flux in MTR even if studies and developments are ongoing. An experiment, called CARMEN-1, was conducted at the OSIRIS MTR and we used in particular a gas ionization chamber based on miniature fission chamber design to measure the photon flux. In this paper, we detail Monte-Carlo simulations to analyze the photon fluxes with ionization chamber measurements and we compare the photon flux calculations to the nuclear heating measurements. These results show a good accordance between photon flux measurements and nuclear heating measurement and allow improving the knowledge of these parameters.

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

    Science.gov (United States)

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

    2009-08-01

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

  6. Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

    2016-06-01

    Low-enriched (U-235 reactors. In most cases, fuel plates with Al or Al-Si alloy matrices have been tested in the Advanced Test Reactor to support this development. In addition, fuel plates with Mg as the matrix have also been tested. The benefit of using Mg as the matrix is that it potentially will not chemically interact with the U-Mo fuel particles during fabrication or irradiation, whereas with Al and Al-Si alloys such interactions will occur. Fuel plate R9R010 is a Mg matrix fuel plate that was aggressively irradiated in ATR. This fuel plate was irradiated as part of the RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

  7. Conversion and standardization of university reactor fuels using low-enrichment uranium: Plans and schedules

    Energy Technology Data Exchange (ETDEWEB)

    Young, H.H.; Brown, K.R.; Matos, J.E.

    1986-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. To minimize this risk, the US Nuclear Regulatory Commission issued its final rule on ''Limiting the Use of Highly Enriched Uranium in Domestically Licensed Research and Test Reactors,'' in February 1986. This paper describes the plans and schedules developed by the US Department of Energy to coordinate an orderly transition from HEU to LEU fuel in most of these reactors. An important element in the planning process has been the desire to standardize the LEU fuels used in US university reactors and to enhance the performance and utilization of a number of these reactors. The program is estimated to cost about $10 million and to last about five years.

  8. Test system mount for ultrasonic testing of the external rotative welding seam of cylindrical construction elements especially in reactor plants

    International Nuclear Information System (INIS)

    The ultra sonic test system is used for testing the external socket welds, tube connection welds and socket edges of a pressure vessel. The test system mount consists of a centered circular frame with a revolving bed track for a transport car. At an axial adjustable sled of the transport car a trailing lever is hinged, ot which the test head is seated cardanically. The trailing lever itself can be pivoted by a piston cylinder system. (DG)

  9. Local stability tests in Dresden 2 boiling water reactor

    International Nuclear Information System (INIS)

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations

  10. Reliability Analysis of High Temperature Reactor Fuels

    International Nuclear Information System (INIS)

    This paper presents the results of reliability analysis of the TRISO -coated fuel particles for the High Temperature Test Reactor (HTTR), Japan. The reliability of fuel particle was evaluated based on the failure probability of each coating layer, and only the failure due to internal gas pressure and shrinkage of pyrolytic carbon (PyC) layer was considered The analysis results show that, no significant failure occurs up to about 45 MWd/kgU for the first core fuel particle and up to about 75 MWd/kgU for the reload core fuel particle. The fuel particle is predicted to fail completely at about 50 MWd/kgU for the first core fuel particle and at about 85 MWd/kgU for the reload core fuel particle. This results show that the TRISO -coated fuel particle for the HTTR to have high reliability. No failure occurs up to the maximum burnup design level, i.e. 33 MWd/kgU for the first core fuel particle and 60 MWd/kgU for the reload core fuel particle. The analysis results show also that the fuel particle reliability (coating layers) depends on the irradiation temperature. The failure occurs at lower burnup if the irradiation temperature increases. (author)

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

    International Nuclear Information System (INIS)

    The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O2 fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required

  12. Spent fuel situation at the ASTRA Seibersdorf and the TRIGA Vienna research reactors

    International Nuclear Information System (INIS)

    In the past decades Austria operated three research reactors, the 10 MW ASTRA reactor at Seibersdorf, the 250 kW TRIGA reactor at the Atomic Institut Vienna and the 1 kW Argonaut reactor at the Technical University in Graz. Since the shut down on July 31st, 1999 and decommissioning of the ASTRA reactor and the shut down of the ARGONAUT reactor Graz on July 31, 2004 only the TRIGA reactor remains operational. The MTR fuel elements of the ASTRA reactor have been shipped in spring 2001 to Savannah River and the fuel plates from the ARGONAUT reactor Graz in December 2005 under the DOE fuel return programme. (author)

  13. Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Daniel M. Wachs; Richard G. Ambrosek; Gray Chang; Mitchell K. Meyer

    2006-10-01

    Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progress toward element testing will be reviewed.

  14. Advanced test reactor. Testing capabilities and plans

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plants for the NSUF. (author)

  15. Thermophysical properties of reactor fuels

    International Nuclear Information System (INIS)

    A review is presented of the literature on the enthalpy of uranium, thorium, and plutonium oxide and an approach is described for calculating the vapor pressure and gaseous composition of reactor fuel. In these calculations, thermodynamic functions of gas phase molecular species (obtained from matrix-isolation spectroscopy) are employed in conjunction with condensed phase therodynamics. A summary is presented of the status of this work

  16. The Jules Horowitz Reactor : A new high Performances European MTR (Material Testing Reactor) with modern experimental capacities : Toward an International User Facility

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is a new Material Testing Reactor (MTR) currently under construction at CEA Cadarache research centre in the south of France. It will be a major Research facility in support to the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimise and demonstrate safe operations of existing power reactors as well as to support future reactor design. It will represent also an important Research Infrastructure for scientific studies dealing with material and fuel behaviour under irradiation. The JHR will contribute also to secure the production of radioisotope for medical application. This is a key public health stake. The construction of JHR which was started in 2007 is on-going. The first operation is planned before the end of this decade.The design of the reactor will provide an essential facility supporting the programs for the nuclear energy for the next 50 years. JHR is designed to provide high neutron flux (enhancing the maximum available today in MTRs), to run highly instrumented experiments to support advanced modelling giving prediction beyond experimental points, and to operate experimental devices giving environment conditions (pressure, temperature, flux, coolant chemistry, ···) relevant for water reactors, for gas cooled thermal or fast reactors, for sodium fast reactors, ···So, the reactor will perform R and D programs for the optimization of the present generation of NPP, support the development of the next generation of NPP (mainly LWR) and also offer irradiation possibilities for future reactors. In parallel to the construction of the reactor, the preparation of an international community around JHR is continuing; this is an important topic as building and gathering a strong international community in support to MTR experiments is a key-issue for the R and D in nuclear energy field. Consequently, CEA is

  17. Comparison of depletion results for a boiling water reactor fuel element with CASMO and SCALE 6.1 (TRITON/NEWT)

    Energy Technology Data Exchange (ETDEWEB)

    Mesado, C.; Morera, D.; Miro, R.; Barrachina, T.; Verdu, G., E-mail: cmesado@isirym.upv.es, E-mail: dmorera@isirym.upv.es, E-mail: rmiro@isirym.upv.es, E-mail: tbarrachina@isirym.upv.es, E-mail: gverdu@isirym.upv.es [Universitat Politecnica de Valencia (UPV), Valencia (Spain). Institute for the Industrial, Radiophysical and Environmental Safety; Concejal, Alberto, E-mail: acbe@iberdrola.es [Iberdrola Ingenieria y Construcion, S.A.U, Madrid (Spain); Soler, Amparo, E-mail: asoler@iberdrola.es [SEA Propulsion S. L., Madrid (Spain); Melara, Jose, E-mail: j.melara@iberdrola.es [Iberdrola Generacion Nuclear, Madrid (Spain)

    2013-07-01

    In this work, the results of depletion calculations with CASMO and SCALE 6.1 (TRITON) are compared. To achieve it, a region of a Boiling Water Reactor (BWR) fuel element is modeled, using both codes. To take into account different operating conditions, the simulations are repeated with different void fraction, ranging from null void fraction to a void fraction closed to one. Special care was used to keep in mind that the homogenization of the materials and the two group approach was the same in both codes. Additionally, KENO-VI and MCDANCOFF modules are used. The k-effective calculated by KENO-VI is used to ensure that the starting point was correct and MCDANCOFF module is used to calculate the Dancoff factors in order to improve the model accuracy. To validate the whole process, a comparison of k{sub eff}, and cross-sections collapsed and homogenized is shown. The results show a very good agreement, with an average error around the 1.75%. Furthermore, an automatic process for translating CASMO data to SCALE input decks was developed. The reason for the translation is the fact that SCALE's TRITON module is a new code very powerful and continuously being developed. Thus, great advantage can be taken from the current and future SCALE features. This is hoped to produce more realistic models, and hence, increase the accuracy of neutronic libraries. (author)

  18. Power Reactor Fuel Reprocessing: Mechanical Phase

    Energy Technology Data Exchange (ETDEWEB)

    Klima, B. B.

    1959-07-01

    The major events in the mechanical phase of the Power Reactor fuels reprocessing program during June were: 1. Feasibility of shearing of fuel elements without disassembly has been demonstrated in tests using porcelain-loaded prototype fuel elements. 2. Further work with the Manco shear was not deemed tb be advisable since permission has been granted to use another shear for cutting UO{sub 2}-loaded fuel elements. 3. Necessity to strip the windows in Building 3048, to sandblast, and repaint them has seriously disrupted occupancy of the cell by July 1. Start of installation probably will not be before August 1. 4. A cold SRE element should be received during July which will permit a direct look a t the problems associated with processing of these irradiated fuel elements. 5. Concurrence with AEC, Atomics International, and ORNL people on the fabrication of a poisoned carrier was obtained and all criteria for the carrier were released and the design was completed. 6. A decision was made to install and use a 24-inch Ty-Sa-Man saw which is on hand and was originally purchased for use in the Segmenting Facility for the SRE reprocessing. This will be used instead of the multipurpose saw to allow more time to refine the design of that saw. The multipurpose saw will be installed for use in subsequent reprocessing programs. This report will chronicle the changes in status which occurred during the calendar month of June. A complete description of each item is not included and may be found in the parent report. The dates indicated on the schedule have slipped since the last report primarily due to increase in scope of the work and postponement on all phases of the work except for the SRE preparations. Twenty-four new items have been added to the schedule. The status of procurement is shown. A total of 93 purchase requests have been turned in to t% Purchasing Department. A total of $199,261.83 has been committed by purchase orders, and a total of 56 purchase orders have been

  19. Topical papers on heavy water, fuel fabrication and reactors

    International Nuclear Information System (INIS)

    A total of four papers is presented. The first contribution of the Federal Republic of Germany reviews the market situation for reactors and the relations between reactor producers and buyers as reflected in sales agreements. The second West German contribution gives a world-wide survey of fuel element production as well as of fuel and fuel element demand up to the year 2000. The Canadian paper discusses the future prospects of heavy-water production, while the Ecuador contribution deals with small and medium-sized nuclear power plants

  20. System design description for the whole element furnace testing system

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, G.A. [Fluor Daniel Northwest (United States); Marschman, S.C. [Pacific Northwest National Lab., Richland, WA (United States); MacFarlan, P.J. [Duke Engineering and Services Hanford, WA (United States); King, D.A. [SGN Eurisys Services Corp. (United States)

    1998-05-01

    This document provides a detailed description of the Hanford Spent Nuclear Fuel (SNF) Whole Element Furnace Testing System located in the Postirradiation Testing Laboratory G-Cell (327 Building). Equipment specifications, system schematics, general operating modes, maintenance and calibration requirements, and other supporting information are provided in this document. This system was developed for performing cold vacuum drying and hot vacuum drying testing of whole N-Reactor fuel elements, which were sampled from the 105-K East and K West Basins. The proposed drying processes are intended to allow dry storage of the SNF for long periods of time. The furnace testing system is used to evaluate these processes by simulating drying sequences with a single fuel element and measuring key system parameters such as internal pressures, temperatures, moisture levels, and off-gas composition.

  1. Deep-Burn Modular Helium Reactor Fuel Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    McEachern, D

    2002-12-02

    This document contains the workscope, schedule and cost for the technology development tasks needed to satisfy the fuel and fission product transport Design Data Needs (DDNs) for the Gas Turbine-Modular Helium Reactor (GT-MHR), operating in its role of transmuting transuranic (TRU) nuclides in spent fuel discharged from commercial light-water reactors (LWRs). In its application for transmutation, the GT-MHR is referred to as the Deep-Burn MHR (DB-MHR). This Fuel Development Plan (FDP) describes part of the overall program being undertaken by the U.S. Department of Energy (DOE), utilities, and industry to evaluate the use of the GT-MHR to transmute transuranic nuclides from spent nuclear fuel. The Fuel Development Plan (FDP) includes the work on fuel necessary to support the design and licensing of the DB-MHR. The FDP is organized into ten sections. Section 1 provides a summary of the most important features of the plan, including cost and schedule information. Section 2 describes the DB-MHR concept, the features of its fuel and the plan to develop coated particle fuel for transmutation. Section 3 describes the knowledge base for fabrication of coated particles, the experience with irradiation performance of coated particle fuels, the database for fission product transport in HTGR cores, and describes test data and calculations for the performance of coated particle fuel while in a repository. Section 4 presents the fuel performance requirements in terms of as-manufactured quality and performance of the fuel coatings under irradiation and accident conditions. These requirements are provisional because the design of the DB-MHR is in an early stage. However, the requirements are presented in this preliminary form to guide the initial work on the fuel development. Section 4 also presents limits on the irradiation conditions to which the coated particle fuel can be subjected for the core design. These limits are based on past irradiation experience. Section 5 describes

  2. RERTR program activities related to the development and application of new LEU fuels. [Reduced Enrichment Research and Test Reactor; low-enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A.

    1983-01-01

    The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the current 1.7 g U/cm/sup 3/ to the 7.0 g U/cm/sup 3/ which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years.

  3. A CFD Simulation Process for Fast Reactor Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Kurt D. Hamman; Ray A. Berry

    2010-09-01

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k–e and SST (Menter) k–? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  4. A CFD simulation process for fast reactor fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hamman, Kurt D., E-mail: Kurt.Hamman@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Berry, Ray A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States)

    2010-09-15

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly 'benchmark' geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-{epsilon} and SST (Menter) k-{omega} were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  5. Proliferation Resistant Nuclear Reactor Fuel

    International Nuclear Information System (INIS)

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

  6. Fuel Temperature Characteristics for Fuel Channels using Burnable Poison in the CANDU reactor

    International Nuclear Information System (INIS)

    Although the CANFLEX RU fuel bundle loaded 11.0 wt% Er2O3 are originally designed focused on the safety characteristics, the fuel temperature characteristics is revealed to be not deteriorated but rather is slightly enhanced by the decreased fuel temperature in the outer ring compared with that of standard 37 fuel bundle. Recently, for an equilibrium CANDU core, the power coefficient was reported to be slightly positive when newly developed Industry Standard Tool set reactor physics codes were used. Therefore, it is required to find a new way to effectively decrease the positive power coefficient of CANDU reactor without seriously compromising the economy. In order to make the power coefficient of the CANDU reactor negative at the operating power, Roh et al. have evaluated the various burnable poison (BP) materials and its loading scheme in terms of the fuel performance and reactor safety characteristics. It was shown that reactor safety characteristics can be greatly improved by the use of the BP in the CANDU reactor. In a view of safety, the fuel temperature coefficient (FTC) is an important safety parameter and it is dependent on the fuel temperature. For an accurate evaluation of the safety-related physics parameters including FTC, the fuel temperature distribution and its correlation with the coolant temperature should be accurately identified. Therefore, we have evaluated the fuel temperature distribution of a CANFLEX fuel bundle loaded with a burnable poison and compared the standard 37 element fuel bundle and CANFELX-NU fuel bundle

  7. Development of nuclear fuel for integrated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

  8. Characteristics of fast reactor core designs and closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N. [State Scientific Center of the Russian Federation, Institute for Physics and Power Engineering (IPPE), Obninsk, Kaluga Region (Russian Federation)

    2007-07-01

    On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

  9. Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

    International Nuclear Information System (INIS)

    Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics

  10. Post-irradiation examination of Fugen reactor fuel assembly at reactor fuel examination facility

    International Nuclear Information System (INIS)

    Post-irradiation examination of the first assembly of a monitoring program for Heavy Water Reactor ''Fugen'' of PNC (Power Reactor and Nuclear Fuel Development Corporation) has been executed since Oct. 1983 at the Reactor Fuel Examination Facility, JAERI Tokai (Japan Atomic Energy Research Institute, Tokai Research Establishment). The fuel assembly is a cylindrical cluster, with 4,400mm length, composed of 28 rods in 3 concentric circles, 12 spring-grid spacers and the upper and lower tie plates. The fuel is plutonium-uranium mixed oxide (0.8 w/o), and the material of cladding tube is Zry-2. The average burnup of the fuel assembly is about 13,600 MWd/t. This paper describes the methods and some results on the post irradiation examination items as follows: 1. Radioactive measurement of water in transportation cask; 2. Visual inspection of the fuel assembly in dry cell, before and after removing the crud, by ultrasonic vibration method; 3. Chemical analyses and radioactive measurement of the crud materials; 4. Dimensional measurement of assembly length and rod-rod gaps, before and after removing the crud; 5. Disassembly and dimensional measurement of rod-rod gaps in the inner circles; 6. Several nondestructive testing techniques of fuel rods. (author)

  11. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  12. Trace metal assay of fast breeder test reactor fuel using D.C. arc and plasma emission spectrometry

    International Nuclear Information System (INIS)

    This report describes the methods developed and used for the trace metal assay of the first charge of the FBTR fuel using a glove-box adapted direct reading emission spectrometer. The group of medium and highly volatile elements are determined in (U,Pu)C fuel samples by d.c. arc carrier distillation technique while the group of lanthanide elements are determined by ICP excitation mode with prior chemical separation from the major matrix. The statistical treatment of the analytical data collected from the analysis of about one hundred samples has indicated good purity of samples and consistent and satisfactory performance of the direct reading spectrometer and associated systems during this period. (author)

  13. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    The purpose of this technical paper is to provide status of the United State domestic Research Reactor Infrastructure (RRI) Program at the Idaho National Laboratory. This paper states the purpose of the program, lists the universities operating TRIGA reactors that are supported by the program, identifies anticipated fresh fuel needs for the reactor facilities, discusses spent fuel activities associated with the program, and addresses successes and planned activities for the program. (author)

  14. Integrated Recycling Test Fuel Fabrication

    International Nuclear Information System (INIS)

    The Integrated Recycling Test is a collaborative irradiation test that will electrochemically recycle used light water reactor fuel into metallic fuel feedstock. The feedstock will be fabricated into a metallic fast reactor type fuel that will be irradiation tested in a drop in capsule test in the Advanced Test Reactor on the Idaho National Laboratory site. This paper will summarize the fuel fabrication activities and design efforts. Casting development will include developing a casting process and system. The closure welding system will be based on the gas tungsten arc burst welding process. The settler/bonder system has been designed to be a simple system which provides heating and controllable impact energy to ensure wetting between the fuel and cladding. The final major pieces of equipment to be designed are the weld and sodium bond inspection system. Both x-radiography and ultrasonic inspection techniques have been examine experimentally and found to be feasible, however the final remote system has not been designed. Conceptual designs for radiography and an ultrasonic system have been made. (author)

  15. Integrated Recycling Test Fuel Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    R.S. Fielding; K.H. Kim; B. Grover; J. Smith; J. King; K. Wendt; D. Chapman; L. Zirker

    2013-03-01

    The Integrated Recycling Test is a collaborative irradiation test that will electrochemically recycle used light water reactor fuel into metallic fuel feedstock. The feedstock will be fabricated into a metallic fast reactor type fuel that will be irradiation tested in a drop in capsule test in the Advanced Test Reactor on the Idaho National Laboratory site. This paper will summarize the fuel fabrication activities and design efforts. Casting development will include developing a casting process and system. The closure welding system will be based on the gas tungsten arc burst welding process. The settler/bonder system has been designed to be a simple system which provides heating and controllable impact energy to ensure wetting between the fuel and cladding. The final major pieces of equipment to be designed are the weld and sodium bond inspection system. Both x-radiography and ultrasonic inspection techniques have been examine experimentally and found to be feasible, however the final remote system has not been designed. Conceptual designs for radiography and an ultrasonic system have been made.

  16. Jules Horowitz Reactor: a high performance material testing reactor

    Science.gov (United States)

    Iracane, Daniel; Chaix, Pascal; Alamo, Ana

    2008-04-01

    The physical modelling of materials' behaviour under severe conditions is an indispensable element for developing future fission and fusion systems: screening, design, optimisation, processing, licensing, and lifetime assessment of a new generation of structure materials and fuels, which will withstand high fast neutron flux at high in-service temperatures with the production of elements like helium and hydrogen. JANNUS and other analytical experimental tools are developed for this objective. However, a purely analytical approach is not sufficient: there is a need for flexible experiments integrating higher scales and coupled phenomena and offering high quality measurements; these experiments are performed in material testing reactors (MTR). Moreover, complementary representative experiments are usually performed in prototypes or dedicated facilities such as IFMIF for fusion. Only such a consistent set of tools operating on a wide range of scales, can provide an actual prediction capability. A program such as the development of silicon carbide composites (600-1200 °C) illustrates this multiscale strategy. Facing the long term needs of experimental irradiations and the ageing of present MTRs, it was thought necessary to implement a new generation high performance MTR in Europe for supporting existing and future nuclear reactors. The Jules Horowitz Reactor (JHR) project copes with this context. It is funded by an international consortium and will start operation in 2014. JHR will provide improved performances such as high neutron flux ( 10 n/cm/s above 0.1 MeV) in representative environments (coolant, pressure, temperature) with online monitoring of experimental parameters (including stress and strain control). Experimental devices designing, such as high dpa and small thermal gradients experiments, is now a key objective requiring a broad collaboration to put together present scientific state of art, end-users requirements and advanced instrumentation. To cite this

  17. Space Nuclear Thermal Propulsion Nuclear Element Tests at Sandia National Laboratories

    International Nuclear Information System (INIS)

    Nuclear Element Tests (NET) are being performed as part of the U.S. Air Force Space Nuclear Thermal Propulsion (SNTP) Program to evaluate high performance fuel elements intended for use in future nuclear propulsion systems. The NET experiments are to be performed at the Sandia National Laboratories (SNL's) Annular Core Research Reactor (ACRR). Objectives of these experiments are to provide engineering validation and demonstration of critical-fuel-element-related technologies and an experimental data base to support analytical design methods for the SNTP Program. Currently, hardware for the first two fueled NET experiments has been fabricated, and cold flow tests have been accomplished with a representative set of hardware to assure the experimental capability to achieve test objectives in-reactor. Assembly of the first NET experiment to test a representative nuclear fuel element is in progress, and planned operational sequences have been defined

  18. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    Science.gov (United States)

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  19. The Conflux Fuel bundle: An Economic and Pragmatic Route to the use of Advanced Fuel Cycles in CANDU Reactors

    International Nuclear Information System (INIS)

    The CANFLEX1 bundle is being developed jointly by AECL and KAERI as a vehicle for introducing the use of enrichment and advanced fuel cycles in CANDU2 reactors. The bundle design uses smaller diameter fuel elements in the outer ring of a 43-element bundle to reduce the maximum element ratings in a CANDU fuel bundle by 20% compared to the 37-element bundle currently in use. This facilitates burnups of greater than 21,000 MW d/TAU to optimize the economic benefit available from the use of enrichment and advanced fuel cycles. A combination of this lower fuel rating, plus development work underway at Aecl to enhance the thermalhydraulic characteristics of the bundle (including both CHF3 and bundle. This provides extra flexibility in the fuel management procedures required for fuel bundles with higher fissile contents. The different bundle geometry requires flow tests to demonstrate acceptable vibration and fretting behavior of the Conflux bundle. A program to undertake the necessary range of flow tests has started at KAERI, involving the fabrication of the required bundles, and setting up for the actual tests. A program to study the fuel management requirements for slightly enriched (0.9 wt % 235 in total U) Conflux fuel has been undertaken by both Aecl and KAERI staff, and further work has started for higher enrichments. Irradiation testing of the Conflux bundle started in the NUR reactor in 1989, and a second irradiation test is due to start shortly. This paper describes the program, and reviews the status of key parts of the program

  20. Upgraded HFIR Fuel Element Welding System

    Energy Technology Data Exchange (ETDEWEB)

    Sease, John D [ORNL

    2010-02-01

    The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. In recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.

  1. New Concept of Designing Composite Fuel for Fast Reactors with Closing Fuel Cycle

    International Nuclear Information System (INIS)

    A possibility is considered for use in fast reactors in place of the base pelletized type MOX or metallic U-Pu-Zr fuel, the dispersion type fuel elements (composite of metallic high uranium content fuel, U-Mo or U-Zr with PuO2 powder distributed in Zr alloy matrix). Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a frame fuel element having a porous uranium alloy meat is filled with standard PuO2 powder of less than 50 micron fractions prepared by pyrochemical or other methods. Proposed composite fuel features higher characteristics in comparison to metallic or MOX fuel. Novel approach to reprocessing of composite fuel is demonstrated, which allows to separate uranium from plutonium as well as the newly generated fissile plutonium from burnt one without chemical processes with repeated use in fast, PWR or CANDU reactors, which simplifies the closing of the nuclear fuel cycle. (author)

  2. Repurposing an irradiated instrumented TRIGA fuel element for regular use

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Paulo F.; Souza, Luiz C.A., E-mail: pfo@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    TRIGA IPR-R1 is a research reactor also used for training and radioisotope production, located at the Centro de Desenvolvimento da Tecnologia Nuclear da Comissao Nacional de Energia Nuclear (Nuclear Technology Development Centre, Brazilian National Nuclear Energy Commission - CDTN/CNEN). Its first criticality occurred in November 1960. All original fuel elements were aluminum-clad. In 1971 nine new fuel elements, stainless steel-clad were acquired. One of them was an instrumented fuel element (IFE), equipped with 3 thermocouples. The IFE was introduced into the core only on August 2004, and remained there until July 2007. It was removed from the core after the severing of contacts between the thermocouples and their extension cables. After an unsuccessful attempt to recover electrical access to the thermocouples the IFE was transferred from the reactor pool to an auxiliary spent fuel storage well, with water, in the reactor room. In December 2011 the IFE was transferred to an identical well, dry, where it remains so far. This work is a proposal for recovery of this instrumented fuel element, by removing the cable guide rod and adaptation of a superior terminal plug similar to conventional fuel elements. This will enable its handling through the same tool used for regular fuel elements and its return to the reactor core. This is a delicate intervention in terms of radiological protection, and will require special care to minimize the exposure of operators. (author)

  3. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

  4. Radiation thermal processes in Cr13Mo2NbVB steel - the material of the fuel assembly shell in reactor BN-350 under mechanical tests

    OpenAIRE

    Larionov, A. S.; Dikov, А. S.; Poltavtseva, V. P.; Kislitsin, S. B.; Kuimova, Marina Valerievna; Chernyavski (Chernyavskiy), Aleksandr Viktorovich

    2015-01-01

    Regularities of changes of structural-phase state and mechanical properties of steel 13Mo2NbVB - the material of the fuel assembly shell in reactor BN-350 after various mechanical tests at 350°C are experimentally studied. The formation of microprecipitations FeMo, enriched or depleted with molybdenum was found in the short-time mechanical tests, which is the cause of thermal hardening of irradiated Cr13Mo2NbVB steel and its destruction by the ductile-brittle mechanism. On the basis of long-t...

  5. Fuel experience at a 37 year old TRIGA type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Boeck, H. [Atominstitut der Oesterreichischen Universitaeten, Wien (Austria)

    1999-07-01

    A survey is given on 37 years of TRIGA fuel experience at the 250 kW TRIGA Mark II reactor Vienna. Approximately 3000 fuel-years of experience have accumulated at this facility with only minor problems. Totally only 8 fuel elements had to be removed permanently from the core. Various inspection methods which have been developed throughout the years are described in this paper. (author)

  6. Fuel element tube testing with on-line signal and data processing

    International Nuclear Information System (INIS)

    The first part gives the development of the testing requirements for cladding material until today, the second part gives at the example flaw measurement and wall variation measurement the advantages of on-line data processing. The third part is a description of existing equipments including control card technic for equipment classification and production statistics. (orig.)

  7. New Concept of Designing Composite Fuel for Fast Reactors with Closing Fuel Cycle

    International Nuclear Information System (INIS)

    For fast reactors a novel type of promising composite U-PuO2 fuel is proposed which is based on dispersion fuel elements. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. Novel fuel features higher characteristics in comparison to metallic or MOX fuel its fabrication technology is readily accomplished and is environmentally clean. A possibility is demonstrated of fabricating coated steel claddings to protect from interaction with fuel and fission products when use standard rod type MOX or metallic U-Pu-Zr fuel. Novel approach to reprocessing of composite fuel is demonstrated, which allows to separate uranium from burnt plutonium as well as the newly generated fissile plutonium from burnt one without chemical processes, which simplifies the closing of the nuclear fuel cycle. Novel composite fuel combines the advantages of metallic and ceramic types of fuel and has high uranium density that allows also to implicate it in BREST types reactor with conversion ratio more than 1. Peculiarities of closing nuclear cycle with composite fuel are demonstrated that allows more effective re-usage of generated Pu as well as, minimizing r/a wastes by incineration of MA in specially developed IMF design

  8. New Concept of Designing Combined Fuel for Fast Reactors with Closed Fuel Cycles

    International Nuclear Information System (INIS)

    New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO2 powder of 2 powder. Framework fuel element having porous meat is fabricated by capillary impregnation method with the use of Zr eutectic matrix alloys, which provides metallurgical bond between fuel and cladding and protects it from interaction. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio does not interact with fuel cladding and is more environmentally clean. A possibility is demonstrated of fabricating coated steel claddings to protect from interaction with fuel and fission products when use standard rod type MOX or metallic U-Pu-Zr fuel. Novel approach to reprocessing of combined fuel is demonstrated, which allows to separate uranium from plutonium as well as the newly generated fissile plutonium from burnt one without chemical processes, which simplifies the closing of the nuclear fuel cycle. (author)

  9. Design, fabrication, and testing of gadolinium-shielded metal fuel samples in the hydraulic tube of the high flux isotope reactor

    International Nuclear Information System (INIS)

    The use of hydraulic rabbit capsules inserted into and ejected from the core of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) during full power operation allows for precise control of the neutron fluence in fueled experiments. Rabbit capsules with strong thermal neutron absorbers must be used to screen out thermal neutrons, thereby reducing the heat generation rate while maintaining the fast neutron flux that produces displacement damage similar to fast reactor type conditions. However, rapid insertion and ejection of rabbit capsules containing a strong neutron absorber causes a reactivity response in the reactor that has the potential to engage the HFIR safety response system which could result in an unplanned shutdown. Therefore, a set of tests were performed to provide the data needed to establish limits on the reactivity worth that can be ejected from the hydraulic facility without causing a reactor shutdown. This paper will describe the design, operation, and results of the reactivity measurements undertaken to understand the reactor response to insertion of the gadolinium-lined rabbit capsules. (author)

  10. The Conceptual Design for a Fuel Assembly of a New Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, J-S.; Cho, Y-G.; Yoon, D-B.; Dan, H-J.; Chae, H-T.; Park, C.

    2004-10-06

    A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibration characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.

  11. Performance and fuel cycle cost study of the R2 reactor with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Pond, R.B.; Freese, K.E.; Matos, J.E.

    1984-01-01

    A systematic study of the experiment performance and fuel cycle costs of the 50 MW R2 reactor operated by Studsvik Energiteknik AB has been performed using the current R2 HEU fuel, a variety of LEU fuel element designs, and two core-box/reflector configurations. The results include the relative performance of both in-core and ex-core experiments, control rod worths, and relative annual fuel cycle costs.

  12. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  13. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2013-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  14. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2011-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  15. Detecting, locating and identifying failed fuel in Canadian power reactors

    International Nuclear Information System (INIS)

    This document summarizes how defected fuel elements are detected, located and identified in Canadian CANDU power reactors. Fuel defects are detected by monitoring the primary coolant for gaseous fission products and radioiodines, while location in core is usually performed on-power by delayed neutron monitoring of coolant samples from individual fuel channels or off-power by gamma-ray monitoring of the channel feeder pipes. The systems and techniques used to detect and locate defected fuel in both Ontario Hydro and CANDU 6 power stations are described, along with examples provided by station experience. The ability to detect and locate defected fuel in power stations was greatly enhanced by a fundamental R and D program, which provided an understanding and models of fission-product release and transport, and the post-defect deterioration of failed fuel. Techniques and equipment used to identify and store defected fuel after it has been discharged from the reactor are briefly reviewed

  16. Research reactor de-fueling and fuel shipment

    Energy Technology Data Exchange (ETDEWEB)

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-08-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures.

  17. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

  18. Proliferation Resistant Nuclear Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-02-18

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

  19. Reactor BN-350 spent fuel handling

    International Nuclear Information System (INIS)

    In pursuance with the Decree No. 456 of the Government of Kazakhstan, dated 22 April of 1999, BN-350 reactor shall be converted to SAFSTOR state for 50 years period followed by dismantling and disposal. Nuclear fuel unloading and safe arrangement for long-term storage in a specially constructed storage facility outside the reactor plant is one of the main criteria of reactor conversion of SAFSTOR state. In accordance with principles of nonproliferation and cancellation of 'nuclear test sites' the 'Baikal-1' bench-top complex located at National Nuclear Center of the Republic of Kazakhstan site is defined by Kazakhstan side decision as a location for long-term storage of BN-350 spent fuel. Project of BN-350 spent fuel transportation and arrangement for long-term storage includes several stages for completion. Currently the spent fuel is unloaded and packed into sealed jackets filled with inert gas. Thus the first Project stage - spent fuel preparation for transportation and provision of necessary temporary storage condition in BN-350 ponds till the moment of transportation is completed. Spent fuel transportation to the place of long-term storage is suggested to conduct in transport packaging casks (TPC) by railway to Kurchatov station where casks will be reloaded for transportation by auto-trailers. For the second Project stage the works have to be carried out on development of the following features: TPC design, technological process of transportation, design of storage facility and both nuclear fuel loading and reloading platforms. This part of this stage is yet completed and main project and technical solution are reported (TPC based on the one pack metal cask, technological process of TPC handling, Silo-type storage facility. As one of the option the TPC is reported based on heavy metal-concrete cask and indented for spent fuel transportation and storage (up to seven canisters with SFAs). Advantages and disadvantages of these TPC are reported compared to that of

  20. Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [Unidad de Actividad Combustibles Nucleares Comision Nacional de Energia Atomica (CNE4), Avda. del Libertador, 8250 C1429BNO Buenos Aires (Argentina)

    2002-07-01

    One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm{sup 3}. PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

  1. Application of finite element numerical technique to nuclear reactor geometries

    International Nuclear Information System (INIS)

    Determination of the temperature distribution in nuclear elements is of utmost importance to ensure that the temperature stays within safe limits during reactor operation. This paper discusses the use of Finite element numerical technique (FE) for the solution of the two dimensional heat conduction equation in geometries related to nuclear reactor cores. The FE solution stats with variational calculus which considers transforming the heat conduction equation into an integral equation I(O) and seeks a function that minimizes this integral and hence gives the solution to the heat conduction equation. In this paper FE theory as applied to heat conduction is briefly outlined and a 2-D program is used to apply the theory to simple shapes and to two gas cooled reactor fuel elements. Good results are obtained for both cases with reasonable number of elements. 7 figs

  2. The DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification Program

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Hans Gougar; Gary Bell

    2005-05-01

    The Department of Energy has established the Advanced Gas Reactor Fuel Development and Qualification Program to address the following overall goals: Provide a baseline fuel qualification data set in support of the licensing and operation of the Next Generation Nuclear Plant (NGNP). Gas-reactor fuel performance demonstration and qualification comprise the longest duration research and development (R&D) task for the NGNP feasibility. The baseline fuel form is to be demonstrated and qualified for a peak fuel centerline temperature of 1250°C. Support near-term deployment of an NGNP by reducing market entry risks posed by technical uncertainties associated with fuel production and qualification. Utilize international collaboration mechanisms to extend the value of DOE resources. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, postirradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete fundamental understanding of the relationship between the fuel fabrication process, key fuel properties, the irradiation performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. Fuel performance modeling and analysis of the fission product behavior in the primary circuit are important aspects of this work. The performance models are considered essential for several reasons, including guidance for the plant designer in establishing the core design and operating limits, and demonstration to the licensing authority that the applicant has a thorough understanding of the in-service behavior of the fuel system. The fission product behavior task will also provide primary source term data needed for licensing. An overview of the program and recent progress will be presented.

  3. Catalyzed deuterium fueled tandem mirror reactor assessment

    International Nuclear Information System (INIS)

    This study was part of a Department of Energy supported alternate fusion fuels program at Science Applications International Corp. The purpose of this portion of the study is to perform an assessment of a conceptual tandem mirror reactor (TMR) that is fueled by the catalyzed-deuterium (Cat-d) fuel cycle with respect to the physics, technology, safety, and cost. Achievable stable betas and magnet configurations are found to be comparable for the Cat-d and d-t fueled TMR. A comparison with respect to cost, reactor performance, and technology requirements for a Cat-d fueled reactor and a comparable d-t fueled reactor such as MARS is also made

  4. Participation in IAEA proficiency test exercise on major, minor and trace elements in ancient Chinese ceramic (IAEA-CU-2006-06) using low power research reactor

    International Nuclear Information System (INIS)

    A proficiency test (PT) exercise was offered by the International Atomic Energy Agency (IAEA) for major, minor and trace elements in Chinese ceramic reference material (IAEA-CU-2006-06). Neutron activation analysis (NAA) laboratory at PINSTECH, Pakistan participated in the exercise and submitted the results for 28 elements. The aim of participation was to develop a suitable methodology for accurate measurement of as many elements as possible in ceramic material using a low power reactor (PARR-2) as this would help future investigation in a project on the authenticity of art objects, for provenance, conservation and management of ancient cultural heritage of the country. After receiving the final report of the PT exercise, a critical review of our data and final scoring of each element is made to check the suitability of our methodology and reliability of the acquired data. Most of the reported results passed different statistical evaluation criterion such as relative bias, z-score and u-scores and ratio of our results and IAEA target values. One element (Yb) falls in the unacceptable range of relative bias and z-scores. Hf and Tb showed slightly high z-scores within the questionable range. Ho, Mo and Sn were determined during this study but their results were not submitted to the IAEA. The confidence of accuracy observed for most of the elements in ceramic material has made it mandatory to report their results as information values. (author)

  5. Uranium density reduction on fuel element side plates assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rios, Ilka A. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

  6. Safety requirements applied to research reactors in France

    International Nuclear Information System (INIS)

    , the safety requirements in this case are the resistance and leaktightness of the pool and the resistance of the containment building to the water column and overpressure due to this accident. A number of tests were performed to validate the safety demonstrations relating to these issues for the above-mentioned reactors. These tests were particularly aimed at: checking the mechanical strength of the pool structures in the event of a BORAX accident; determining the operating limits of the fuels used; examining the impact on the core of a pressurized irradiation loop rupture; validating the design of clad failure detection systems. The safety reviews systematically performed at the request of the safety authority, for reactors that have been in operation for more than 10 years have not called into question the above-mentioned safety requirements. The main requests of the safety authority following the assessments made by the IRSN and presented to the standing group of experts for the nuclear reactors have concerned the need for upgrading the facilities to take into account in particular the evolutions of the requirements relating to internal and external hazards. In particular, this has led to facilities being reinforced to increase their earthquake resistance, and to improvements concerning fire protection and the management of post-accident situations. The aging of facilities has been addressed by gradually refurbishing their equipment and replacing their instrumentation, their control and protection systems and those structures affected by high radiation doses. The operating experience feedback acquired from existing research reactors, the evolution of certain safety requirements, and the progress achieved in the methodology adopted for the study of research reactor safety have been taken into account in the design of the new RJH reactor (P = 100 MW) which is planned to be constructed in the Cadarache site. The final paper will provide the detailed presentation of the

  7. High-Uranium-Loaded U3O8-Al fuel element development program. Part 1

    International Nuclear Information System (INIS)

    The High-Uranium-Loaded U3O8-Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U3O8-Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U3O8-Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U3O8). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % U involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U3O8 and aluminum

  8. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

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

  9. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  10. About a fuel for burnup reactor of periodical pulsed nuclear pumped laser

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, A.I.; Lukin, A.V.; Magda, L.E.; Magda, E.P.; Pogrebov, I.S.; Putnikov, I.S.; Khmelnitsky, D.V.; Scherbakov, A.P. [Russian Federal Nuclear Center, Snezhinsk (Russian Federation)

    1998-07-01

    A physical scheme of burnup reactor for a Periodic Pulsed Nuclear Pumped Laser was supposed. Calculations of its neutron physical parameters were made. The general layout and construction of basic elements of the reactor are discussed. The requirements for the fuel and fuel elements are established. (author)

  11. Research project ILW and HTR fuel element test disposal in boreholes (MHV project). Retrieval test storage (project part REV). Final report

    International Nuclear Information System (INIS)

    The report documents the status of work achieved by the end of 1992. Ten waste packages, i.e. six 200-l-drums with cemented dissolver sludges, cemented cladding hulls and cemented fuel hardware, from the Karlsruhe Reprocessing Plant (WAK) and four 220-l-AVR canisters containing 950 spent HTR fuel elements each from the AVR were to be emplaced in unlined boreholes on the 800-m-level of the Asse salt mine so as to be retrievable. The planned emplacement test, lasting a maximum of five years, was intended to simulate the operating phase of boreholes filled with a comparable waste spectrum in a future repository in salt. Monitoring of the emplacement boreholes by different kinds of measuring methods and set-ups was intended to provide evidence that borehole emplacement of such waste could be implemented from the safety perspective, i.e. from the radiation protection and operational safety standpoint. Furthermore, the basic principles of all package handling, shipping and emplacement procedures relevant to safety, which would also have to be carried out in a similar manner in a future repository, were to be tested and demonstrated. However, this test was not approved by the Ministry of the Environment of Lower Saxony for political reasons. (orig./HP)

  12. Microstructure and elemental distribution of americium-containing MOX fuel under the short-term irradiation tests

    International Nuclear Information System (INIS)

    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. (author)

  13. Fuel behavior comparison for a research reactor

    Science.gov (United States)

    Negut, Gh.; Mladin, M.; Prisecaru, I.; Danila, N.

    2006-06-01

    The paper presents the behavior and properties analysis of the low enriched uranium fuel, which will be loaded in the Romanian TRIGA 14 MW steady state research reactor compared with the original high enriched uranium fuel. The high and low enriched uranium fuels have similar thermal properties, but different nuclear properties. The research reactor core was modeled with both fuel materials and the reactor behavior was studied during a reactivity insertion accident. The thermal hydraulic analysis results are compared with that obtained from the safety analysis report for high enriched uranium fuel core. The low enriched uranium fuel shows a good behavior during reactivity insertion accident and a revised safety analysis report will be made for the low enriched uranium fuel core.

  14. Mixed oxide fuel for water cooled reactors

    International Nuclear Information System (INIS)

    The problems connected with introduction of plutonium extracted from spent fuels of operating NPPs into water cooled reactor fuel cycle are considered. The trends in formation of the World market of mixed fuel are illustrated taking as examples Great Britain and Japan

  15. Environmental concerns regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN; Atomos para el desarrollo de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Santos, G. R. T.; Durazzo, M.; Carvalho, E. F. U. [IPEN, CNEN-SP, P.O. Box 11049, CEP 05422-970, Sao Paulo (Brazil); Riella, H. G. [Universidade Federal de Santa Catarina, Departamento de Engenharia Quimica, Campus Universitario, Florianopolis, CEP 88040-900 (Brazil)]. e-mail: grsantos@ipen.br

    2008-07-01

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the maim programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel - CCN - is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt% {sup 2}35U), to supply its IEA-RI research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the Sustainable Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to

  16. Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

    Science.gov (United States)

    Whitmarsh, C. L., Jr.

    1973-01-01

    An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation.

  17. The International Marketing Target of Fuel Element for Research Rectors

    International Nuclear Information System (INIS)

    The International marketing efforts of PT BATAN Teknologi's fuel element for research reactors are out line. These efforts intensively started in third year marketing time since it is commenced on 24 May 1996. The market segmentation told that there are 269 research reactors in the world, I.e. 65 in USA, 27 in Russia, 18 in Japan, and the remaining are in many Countries. Many of those are 78 swimming fool type reactors, and 17 of them, I.e. 4 in Japan, 4 in USA, and each Austria, Germany, Argentina, Iran, Pakistan, Peru, Brazil, Algeria and Indonesia have the similar fuel element specifications with are close related with PT BATAN Teknologi's. It can be predicated that around 38 fuel elements and 84 fuel control can be marketed. The first feasibility study told that for countries such as Peru, Pakistan, Iran, Algeria, became the potential marketing target of the BATAN Teknologi's fuel element, because for those countries the competitors in producing such fuel elements could be minimal. The fuel elements and fuel control which could be presumably marketed in those countries are 83 and 19 respectively. The problem will be facing in near future such as packaging design and nuclear fuel transportation have to be firstly solved by collaborating with foreign companies abroad. Non technical problems including political situation have to be completely studied in order the uranium, transfer to many countries for exporting purpose could easily take place in the future. The government of the Republic of Indonesia (in this case BATAN) and the International Atomic Energy Agency (IAEA) could assist to solve the non technical problems which might be appear in the future as the chance of the exporting the fuel elements and the fuel controls come true. (author)

  18. Impact of uranium concentration reduction in side plates of the fuel elements of IEA-R1 reactor on neutronic and thermal hydraulic analyses; Impacto da reducao na concentracao de uranio nas placas laterais dos elementos combustiveis do reator IEA-R1 nas analises neutronica e termo-hidraulica

    Energy Technology Data Exchange (ETDEWEB)

    Rios, Ilka Antonia

    2013-09-01

    This master thesis presents a study to verify the impact of the uranium concentration reduction in the side plates of the reactor IEA-R1 fuel elements on the neutronic and thermal-hydraulic analyses. To develop such study, a previous IPEN-CNEN/SP research was reproduced by simulating the fuel elements burn-up, with side plate uranium density reduced to 50, 60 and 70% of the standard fuel element plates. This research begins with the neutronic analysis using the computer code HAMMER and the first step consists in the calculation of the cross section of all materials presented at the reactor core, with their initial concentration; the second step consists in the calculation of the fast and thermal neutron group fluxes and power densities for fuel elements using the computer code CITATION. HAMMER output data is used as input data. Once the neutronic analysis is finished and the most critical fuel elements with highest power density have been defined, the thermal-hydraulics analysis begins. This analysis uses MCTR-IEA-R1 thermal-hydraulics model, which equations are solved by commercial code EES. Thermalhydraulics analysis input is the power density data calculated by CITATION: it is considered the highest power density on each fuel element, where there is a higher energy release and, consequently, higher temperatures. This data is used on energy balance equations to calculate temperatures on critical fuel element regions. Reactor operation comparison for three different uranium densities on fuel side plates is presented. Uranium density reduction contributes to the cladding surface temperature to remain below the established limit, as reactor operation safety requirement and it does not affect significantly fuel element final burn-up nor reactor reactivity. The reduction of uranium in the side plates of the fuel elements of the IEA-R1 showed to be a viable option to avoid corrosion problems due to high temperatures. (author)

  19. Sodium fast reactors with closed fuel cycle

    CERN Document Server

    Raj, Baldev; Vasudeva Rao, PR 0

    2015-01-01

    Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled fast reactors with closed fuel cycles, this book:Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspectsFeatures a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspecti

  20. Modelling fuel behaviour in a reactor park using fuel cycle kinetics

    International Nuclear Information System (INIS)

    In this thesis the theory of fuel cycle kinetics is re-examined. The fuel cycle kinetics theory is a powerful tool to describe the time-dependent fuel behaviour of large populations of nuclear reactors. The fuel cycle kinetics theory is based on the point kinetics theory and the principles of a reactor park. The point kinetics theory is a simplification of the space-, energy-and time-dependent diffusion balance equation to only a time-dependent equation. A reactor park is the description of the interconnections between a population of nuclear reactors with various designs. In the fuel cycle kinetics theory the point kinetics theory is used as a model to simplify space- energy- and time-dependent burn-up equations of the reactors in a reactor park to a set of only time-dependent equations, one for every reactor type. The fuel cycle kinetics theory is verified by means of a number of test cases. In the first test case the same symbiotic system is used as was used by Maudlin. There is no difference between the two obtained results. The second test case is that of only Fast Breeder Reactor, FBR, deployment. Here the result of the fuel cycle kinetics equation is checked against the result obtained from TRITON. TRITON is a module of the SCALE code system that is used for depletion analysis of 3-D reactor models. With the use of the pseudo-initial condition the results of the fuel cycle kinetics and TRITON calculations are almost identical. The pseudo-initial condition is a correction on the initial condition to adjust for neglecting the time dependency of the parameters in the fuel cycle kinetics equations. In the third case a symbiotic system of FBRs and Pressurised Water Reactors, PWRs, is researched. There is only a small difference in the asymptotic growth between the fuel cycle kinetics results and the TRITON results. In the last test case the same system of FBRs and PWRs is used to investigate two demanded asymptotic growths obtained from the upper and lower

  1. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  2. Reactor Simulator Testing Overview

    Science.gov (United States)

    Schoenfeld, Michael P.

    2013-01-01

    Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  3. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  4. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  5. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    International Nuclear Information System (INIS)

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas release