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Sample records for cadarache fuel element testing reactor

  1. Research and Test Reactor Fuel Elements (RTRFE)

    International Nuclear Information System (INIS)

    BWX Technologies Inc. (BWXT) has experienced several production improvements over the past year. The homogeneity yields in 4.8 gU/cc U3Si2 plates have increased over last year's already high yields. Through teamwork and innovative manufacturing techniques, maintaining high quality surface finishes on plates and elements is becoming easier and less expensive. Currently, BWXT is designing a fabrication development plan to reach a fuel loading of 9 gU/cc within 2 - 4 years. This development will involve a step approach requested by ANL to produce plates using U-8Mo at a loading of 6 gU/cc first and qualify the fuel at those levels. In achieving the goal of a very high-density fuel loading of 9 gU/cc, BWXT is considering employing several new, state of the art, ultrasonic testing techniques for fuel core evaluation. (author)

  2. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  3. Development and testing of the EDF-2 reactor fuel element

    International Nuclear Information System (INIS)

    This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors)

  4. Results of the BREST-300 type reactor model fuel elements testing in the IGR reactor

    International Nuclear Information System (INIS)

    Testings of BREST-300 type fast reactor's model fuel elements with nitride fuel in the lead coolant in the central experimental channel of IGR reactor were carried out. In the testing the regime of non-controlled power burst was simulated. In the result of testing the seal failure of fuel elements with 2 % and 10 % 235U enrichment has been occurred, and fragmentation of the part of fuel pellets at interaction with coolant has been taken place. During the reactor testing the measurements and registration of experimental parameters (temperature of fuel, shell, coolant; pressure in fuel elements and testing ampoule; power release in the reactor) were conducted. The physical study of the 'fuel element - ampoule - reactor' was carried out, after-start-up spectrometric and material testing studies, calculated evaluation of temperature fields parameters in the testing ampoule were examined as well. Calculated and experimental values of breaking down specific power releases in the fuel are obtained. The assessment of both fuel fragmentation rate and it character is carried out. Distribution of fuel fragmentation within experimental ampoule volume is studied

  5. 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)

  6. NUCLEAR REACTOR FUEL ELEMENT

    Science.gov (United States)

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  7. Sipping Test: Checking for Failure of Fuel Elements at the OPAL Reactor

    International Nuclear Information System (INIS)

    Sipping measurements were implemented at the Open Pool Australian Light water reactor (OPAL) to test for failure in reactor fuel elements. Fission product released by the fuel element into the pool water was measured using both High Purity Germanium (HPGe) detection via samples and a NaI(Tl) detection in-situ with the sipping device. Results from two fuel elements are presented

  8. Tests in an MR reactor of experimental fuel elements of the VVER-1000 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Goncharov, V.V.; Dubrovin, K.P.; Ivanov, E.G.; Korneev, V.T.; Kruglov, A.B.; Lebedev, L.M.

    1987-11-01

    In an MR reactor performance tests of 16 fuel assemblies, with elements having essentially the same structure as standard VVER-100 fuel elements, were carried out. Tests of five more fuel assemblies are continuing. Of the 16 assemblies, 13 were studied in a hot laboratory. The test in the MR, carried out at high loads and with a large number of transition processes, as well as the postreactor studies, indicated the fuel elements of the specified design (with initial helium pressures of 1.96-2.45 MPa) have a high reliability. None of the elements of the fuel assemblies studied malfunctioned due to design defects or faults in their fabrication. During the tests the jackets were subject to a little oxidation and hydrogenation (zirconium-oxide film < ..mu..m thick, hydrogen content less than 0.008% by mass), and their plasticity remained high (the relative elongation at the working temperature remained at the 20% level).

  9. Tests in an MR reactor of experimental fuel elements of the VVER-1000 reactor

    International Nuclear Information System (INIS)

    In an MR reactor performance tests of 16 fuel assemblies, with elements having essentially the same structure as standard VVER-100 fuel elements, were carried out. Tests of five more fuel assemblies are continuing. Of the 16 assemblies, 13 were studied in a hot laboratory. The test in the MR, carried out at high loads and with a large number of transition processes, as well as the postreactor studies, indicated the fuel elements of the specified design (with initial helium pressures of 1.96-2.45 MPa) have a high reliability. None of the elements of the fuel assemblies studied malfunctioned due to design defects or faults in their fabrication. During the tests the jackets were subject to a little oxidation and hydrogenation (zirconium-oxide film < μm thick, hydrogen content less than 0.008% by mass), and their plasticity remained high (the relative elongation at the working temperature remained at the 20% level)

  10. Sipping tests for the irradiated fuel elements of the TR-2 research reactor

    International Nuclear Information System (INIS)

    Sipping tests have been performed for fuel elements of the TR-2 reactor at Cekmece Nuclear Research and Training Center (CNRTC), in order to find out which one failed in the core. A sipping assembly has been constructed and placed in the pool of the TR-2 reactor. The assembly identifies leaking fuel elements by collecting and measuring 137Cs that leak out from the defective fuel elements. 31 fuel elements in the reactor have been tested for the clad integrity. The measured 137Cs activity of the fuel element with an identification number S-104 is a 10247 Bq/(0.3 l). This value is approximately 234 times greater than the average of the other tested fuel elements in the reactor. (orig.)

  11. 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)

  12. Cadarache - 20 years of plutonium fuels

    International Nuclear Information System (INIS)

    The qualitative and quantitative evolution of plutonium used is reviewed; more particularly, the isotopic composition of plutonium handled at the plutonium fuel fabrication plant of Cadarache has been considerably changing during these 20 last years. The evolution of fast neutron reactor fuel assemblies explains why the fabrication unit has to. The installations have been modified, as also the fuel assembly fabrication process. Changes can be classified in 3 categories: the transformations related to mass evolution, those related more particularly to the plutonium isotopic composition, finally those related to the waste treatment process. These transformations concerned workshops, apparatus and equipments, the reinforcement of protections, criticality accident prevention, and safeguards

  13. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Ionescu, D.; Olteanu, G. [Inst. for Nuclear Research, Pitesti (Romania)

    2008-07-01

    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{sup -1} UO{sub 2}. 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)

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

    International Nuclear Information System (INIS)

    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)

  16. 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)

  17. Conceptual design of experimental LFR fuel element for testing in TRIGA reactor, ACPR zone

    International Nuclear Information System (INIS)

    In the pulsed area of the TRIGA reactor (ACPR zone), the irradiation tests called ''rapid insertions of reactivity on different types of nuclear fuel elements'' are usually realized. During these tests, in the fuel element high powers for a relatively short period of time (about few milliseconds) are generated. The generated heat in fuel pellets raise their central temperature to values over 100 deg C. The conceptual design of an experimental fuel element proposed to be developed and presented in this paper must fulfill a couple of requirements, as follows: to ensure full compatibility with irradiation device sample holder (compatibility is achieved through reduced length of the fuel stack pellets - this way assures a flow flattening on the entire length of the fuel element); to be compatible with the project of irradiated fuel bundle in Lead cooled Fast Reactors (LFR). (authors)

  18. 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.

  19. Performance and testing of refractory alloy clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO2) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at .% burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  20. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G. [Inst. for Nuclear Research (INR), Pitesti (Romania); Palleck, S. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada); Ionescu, D. [Inst. for Nuclear Research (INR), Pitesti (Romania)

    2010-07-01

    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)

  2. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    The fuel element for a BWR known from the patent application DE 2824265 is developed so that the screw only breaks on the expansion shank with reduced diameter if the expansion forces are too great. (HP)

  3. 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

  4. 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)

  5. Independent loop channel for testing the fuel elements for the BREST-OD-300 reactor

    International Nuclear Information System (INIS)

    The problems connected with irradiation tests of the fuel element simulators for the BREST-OD-300 fast lead cooled reactor in the BOR-60 reactor are discussed. The mononitride uranium-plutonium fuel in the container-type fuel elements is used in the BREST-OD-300 pilot reactor being under design. An independent loop channel with lead coolant is created in the BOR-60 reactor cell D-23 for testing the fuel element simulators. The channel thermal power is of the order of 45 kw, its power service life time amounts to 11000 hours, the fuel assembly inlet/outlet temperatures are up to 480/540 deg C, the lead coolant flow rate is up to 4.5 kg/s. The tests will give an opportunity to refine the information on the fuel properties in low-temperature range, as well as to determine the fuel can radiation and corrosion resistance under simultaneous effects of coolant, irradiation and spacing grids

  6. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-12-15

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

  8. Hydraulic and hydrodynamic tests for design evaluation of research reactors fuel elements

    International Nuclear Information System (INIS)

    During the design steps of research reactors fuel elements some tests are usually necessary to verify its design, i.e.: its hydraulic characteristics, dynamical response and structural integrity. The hydraulic tests are developed in order to know the pressure drops characteristics of different parts or elements of the prototype and of the whole fuel element. Also, some tests are carried out to obtain the velocity distribution of the coolant water across different prototype's sections. The hydrodynamic tests scopes are the assessment of the dynamical characteristics of the fuel elements and their components and its dynamical response considering the forces generated by the coolant flowing water at different flow rate conditions. Endurance tests are also necessary to qualify the structural design of the FE prototypes and their corresponding clamp tools, verifying the whole system structural integrity and wear processes influences. To carry out these tests a special test facility is needed to obtain a proper representation of the hydraulic and geometric boundary conditions of the fuel element. In some cases changes on the fuel element prototype or dummy are necessary to assure that the data results are representative of the case under study. Different kind of sensors are mounted on the test section and also on the fuel element itself when necessary. Some examples of the instrumentation used are strain gauges, displacement transducers, absolute and differential pressure transducers, pitot tubes, etc. The obtained data are, for example, plates' vibration amplitudes and frequencies, whole bundle displacement characterization, pressure drops and flow velocity measurements. The Experimental Low Pressure Loop is a hydraulic loop located at CNEA's Constituyentes Atomic Center and is the test facility where different kind of tests are performed in order to support and evaluate the design of research reactor fuel elements. A brief description of the facility, and examples of

  9. 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.

  10. Testing of experimental fuel elements for VVER-1000 reactors in MR to high fuel burnup

    International Nuclear Information System (INIS)

    Pressurized water reactors are given a commanding role in the development program for the nation's nuclear power industry. Considerable operating experience has been gained with VVER-1000 reactors. As of the start of 1990, 17 units with VVER-1000 reactors were in operation in this country and abroad. The first loadings were designed for a 2-year run with average fuel burnup of 28.5 MW-day/kg. The rod-type fuel elements used in the reactors displayed high serviceability and reliability (leakage does not exceed 0.02%). Operating experience and the results of computational and experimental work have made it possible to substantiate the possibility of switching them to a 3-year run. The fifth unit of the Novovoronezh Atomic Power Plant was the first to be switched to a 3-year run, as of 1984. The average fuel burnup achieved after three fuel cycles was 42.6 MW-day/kg. All units with VVER-1000 reactors are now being switched to a 3-year run with an average burnup of more than 40 MW-day/kg for the unloaded fuel

  11. 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)

  12. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    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. (HP)

  13. 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.

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

    International Nuclear Information System (INIS)

    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.

  15. 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

  16. Unirradiated high temperature reactor fuel element head-end reprocessing tests

    International Nuclear Information System (INIS)

    For several years, the United States and the Federal Republic of Germany (FRG) have engaged in a successful cooperative program to develop high temperature gas-cooled reactor (HTGR) fuel cycle technology. Recent tests in reprocessing pilot plant facilities at General Atomic Company have demonstrated the feasibility of performing HTGR head-end unit operations for both spherical (German) and block-type (American) fuel elements in a single process line. Because of an unexpected high fines generation and elutriation rate, extended fluidized bed primary burning of FRG fuel material was impossible to accomplish with the burner system and operating procedures optimized for U.S. fuel burning. Operational modification, including startup with a carbon-poor bed and reduction of the fluidizing velocity, resulted in dramatic improvements in FRG fuel-burning behavior and allowed extended processing campaigns. Additional modifications to the fines recycle system and burner are recommended to optimize the system for processing of FRG fuels

  17. 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)

  18. Test programme for the nuclear fuel of the Jules Horowitz reactor starts

    International Nuclear Information System (INIS)

    In Cadarache in the south of France, an advanced material test reactor is currently being built. The Jules Horowitz reactor is an initiative of the French Atomic Energy Commission ((CEA), supported by the European Commission and research centres in the Czech Republic, Spain, Finland and Belgium, plus a number of large energy companies. On the request of CEA, SCK-CEN started characterising the nuclear fuel for this reactor, in its own BR2. The article describes SCK-CEN's programme on the characterisation of new nuclear fuel elements.

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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 an experimental fuel subassembly of 37 pins has been fabricated for irradiation in Fast Breeder Test Reactor at Kalpakkam near Chennai. MOX fuel pins containing 45% PUO2 have also been also made for the hybrid core of FBTR. The experimental sub-assembly for irradiation testing in FBTR consists of 37 short length 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 UO 2. Uranium enriched with U233 was used to stimulate 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. MOX powder were mixed, milled, pre-compacted and granulated. The final compaction was done using a multistation rotary press with suitable tooling for making annular MOX pellets. The technology for making annular pellets was developed for this purpose. The pellets were sintered at reducing atmosphere at 1650 deg. C for 4 hours to obtain acceptable quality pellets. Over sized pellets were centrelessly ground without using a liquid coolant. The acceptable pellets were degassed before encapsulation. MOX fuel stack, UO2 insulation pellets, plenum spring and spring support were loaded in bottom endplug welded clad tube. The end plug welding was carried out by TIG welding technique. The welded elements after inspection were wire wrapped. 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

  4. Integrity Test on The Fuel Element Cladding of The TRIGA 2000 Reactor Using Hot Sipping Test Method

    International Nuclear Information System (INIS)

    A series of leakage detection tests have been carried out to identify integrity of the fuel element cladding of the TRIGA 2000 reactor. The hot sipping test method was utilized in the tests. The hot sipping test method is one of the techniques to find out possible leakage on the fuel claddings, by detecting the presence of fission product nuclides in fuel elements soaking water that has been well isolated both before and after the irradiation. The fission products as leakage indicator were noble gases nuclide i.e. Xe-138, Kr-87, Kr-88, Kr-85 m, Xe-135, and daughters. It has been identified that 3 fuel element releasing fission products with maximum activity 0,4 kBq/L (Kr-87), 0,6 kBq/L (Kr-88), 0,3 kBq/L (Kr-85 m), 0,3 kBq/L (Xe-135), dan 1,7 kBq/L (Cs-138). The activity of noble gases were identified and measured using gamma spectrometry device. Those fuel elements have been removed from the core for further analysis to know the source and cause of noble gas released. (author)

  5. 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

  6. Nuclear reactor fuel elements charging tool

    International Nuclear Information System (INIS)

    To assist the loading of nuclear reactor fuel elements in a reactor core, positioning blocks with a pyramidal upper face charged to guide the fuel element leg are placed on the lower core plate. A carrier equipped with means of controlled displacement permits movement of the blocks over the lower core plate

  7. 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.)

  8. Tests of experimental fuel elements by the method of nuclear-thermal pulse loadings in 'HYDRA' reactor

    International Nuclear Information System (INIS)

    The results of tests of experimental fuel elements with uranium dioxide fuel composition embedded in Al and Zr matrix with the enrichment from 90% to 36% in respect to U-235 performed at the pulse 'HYDRA' reactor are presented in this paper. Testing is performed in the frame-work of extensive research program studying the behavior of fuel elements (FE) of research and mini nuclear power systems in case of practically immediate energy release in the fuel taking place during the RIA-type accidents. Duration of the neutron pulse when testing in 'HYDRA' reactor is from 7 to 20 ms. The methods of diagnostics of the state of FE prior to and after testing in the reactor are developed and verified. Mathematical model describing temperature fields inside the FE in the process of testing. and accounting for non-uniformity of fuel composition has been developed in order to summarize experimental results. Experimental data on the limiting values of the energy density leading to deformation and degradation of FE depending on the type of fuel composition have been obtained and the mechanisms for the development of these processes have been determined. The nature of physical-chemical processes taking place in the fuel composition and fuel cladding depending on material composition under different levels of energy deposition is demonstrated. The data on hydrogen generation and radioactive product release out of fuel after failure of FE are presented. (author)

  9. Characterization of spent fuel elements stored at IEA-R1 research reactor based on visual inspections and sipping tests

    International Nuclear Information System (INIS)

    Aluminum spent nuclear fuels are susceptible to corrosion attack, or mechanical damage from improper handling, while in pool reactor storage. Storage practices have been modified to reduce the potential for damage, based on recommendations presented at second WS on Spent Fuel Characterization, promoted by IAEA. In this work, we present the inspection program proposed to the IEA-R1 stored spent fuel elements, in order to provide information on the physical condition during the interim storage time under wet condition at the reactor pool. The inspection program is based on non-destructive tests results (visual inspection and sipping tests) already periodically performed to exam the IEA-R1 stored spent fuel and fuel elements from the core reactor. To record the available information and examination results it was elaborated a document in the format of a catalogue containing the proposed inspection program for the IEA-R1 stored spent fuel, the description of the visual inspection and sipping tests systems, a compilation of information and images result from the tests performed for all stored standard spent fuel element and, in annexes, copies of the reference documents. That document constitutes an important step of the effective implementation of the referred IEA-R1 spent fuel inspection program and can be used to address regulatory and operational needs for the demonstration, for example, of safe storage throughout the pool storage period. (author)

  10. Fabrication of ORNL Fuel Irradiated in the Peach Bottom Reactor and Postirradiation Examination of Recycle Test Elements 7 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Long, Jr. E.L.

    2001-10-25

    Seven full-sized Peach Bottom Reactor. fuel elements were fabricated in a cooperative effort by Oak Ridge National Laboratory (ORNL) and Gulf General Atomic (GGA) as part of the National HTGR Fuel Recycle Development Program. These elements contain bonded fuel rods and loose beds of particles made from several combinations of fertile and fissile particles of interest for present and future use in the High-Temperature Gas-Cooled Reactor (HTGR). The portion of the fuel prepared for these elements by ORNL is described in detail in this report, and it is in conjunction with the GGA report (GA-10109) a complete fabrication description of the test. In addition, this report describes the results obtained to date from postirradiation examination of the first two elements removed from the Peach Bottom Reactor, RTE-7 and -4. The fuel examined had relatively low exposure, up to about 1.5 x 10{sup 21} neutrons/cm* fast (>0.18 MeV) fluence, compared with the peak anticipated HTGR fluence of 8.0 x 10{sup 21}, but it has performed well at this exposure. Dimensional data indicate greater irradiation shrinkage than expected from accelerated test data to higher exposures. This suggests that either the method of extrapolation of the higher exposure data back to low exposure is faulty, or the behavior of the coated particles in the neutron spectrum characteristic of the accelerated tests does not adequately represent the behavior in an HTGR spectrum.

  11. Fabrication of ORNL Fuel Irradiated in the Peach Bottom Reactor and Postirradiation Examination of Recycle Test Elements 7 and 4

    International Nuclear Information System (INIS)

    Seven full-sized Peach Bottom Reactor fuel elements were fabricated in a cooperative effort by Oak Ridge National Laboratory (ORNL) and Gulf General Atomic (GGA) as part of the National HTGR Fuel Recycle Development Program. These elements contain bonded fuel rods and loose beds of particles made from several combinations of fertile and fissile particles of interest for present and future use in the High-Temperature Gas-Cooled Reactor (HTGR). The portion of the fuel prepared for these elements by ORNL is described in detail in this report, and it is in conjunction with the GGA report (GA-10109) a complete fabrication description of the test. In addition, this report describes the results obtained to date from postirradiation examination of the first two elements removed from the Peach Bottom Reactor, RTE-7 and -4. The fuel examined had relatively low exposure, up to about 1.5 x 1021 neutrons/cm* fast (>0.18 MeV) fluence, compared with the peak anticipated HTGR fluence of 8.0 x 1021, but it has performed well at this exposure. Dimensional data indicate greater irradiation shrinkage than expected from accelerated test data to higher exposures. This suggests that either the method of extrapolation of the higher exposure data back to low exposure is faulty, or the behavior of the coated particles in the neutron spectrum characteristic of the accelerated tests does not adequately represent the behavior in an HTGR spectrum

  12. PIE of a CANDU fuel element irradiated for a load following test in the INR TRIGA reactor

    International Nuclear Information System (INIS)

    As part of the collaboration under the Romania - Canada Memorandum for co-operation in research and development of nuclear energy and technology, a load following test has been devised to demonstrate the load following capability of CANDU-6 fuel within the established design envelope for operating powers. A 37-element CANDU-6 fuel bundle element fabricated by AECL was irradiated in the TRIGA 14 MW(th) material testing reactor at the Institute for Nuclear Research (INR) in Pitesti, Romania. The load following cycle consisted of 200 daily cycles from 100% power to 50% power within the reference overpower envelope for fuel in a CANDU-6 reactor. Full power operation was 57 kW/m Element Linear Power. The paper provides the results obtained by post-irradiation examination of the fuel element in the INR hot cells. The following techniques were used: - Visual inspection and photography by periscope; - Profilometry; - Axial gamma scanning; - Fuel element puncturing and fission gas analysis; - Metallographic and ceramographic examinations by optical microscopy; - Burn-up measurement by mass spectrometry using the 235U depletion method. (authors)

  13. 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

  14. 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

  15. Soreq Nuclear Reactor Fuel Element Flow Distribution

    International Nuclear Information System (INIS)

    Flow of cold water through the Soreq Nuclear Reactor fuel element was simulated numerically. The main objective of the present study was to obtain the flow distribution among the rectangular channels of the element. The results of the simulations were compared to the overall pressure drop on the element measured in Soreq Nuclear Reactor. The numerical model chosen has succeeded in predicting the pressure drop on the fuel element of up to 5% from the measured values. Flow through the IPEN IEA-R1 MTR fuel element was also simulated as a part of a model validation procedure. The numerical results were compared to the measurements available in the literature [1]. It was found that the water pool above the fuel element has a significant influence on the flow distribution among the channels of the element. The flow distribution reported in [1] was closely predicted numerically when the water pool was included into the simulated geometry. It can be concluded that flow distribution in the Soreq Nuclear Reactor fuel element is flatter than that in the IPEN IEA-R1 MTR fuel element

  16. 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

  17. Process and device for testing the dimensions of a fuel element for nuclear reactors

    International Nuclear Information System (INIS)

    Such a process was previously only used to measure the distance between two fuel rods (OS 3542204). In order to use such a process for testing the length of a fuel element, it is provided that a wide side of a sensor away from the ultrasonic test head is slowly driven to the front of the end plate and is made to make contact there. After successful contact is made, due to the sprung construction of the sensors, a distance is set between the two sensors, which is less than the distance when the sensor does not make contact. The distance set can be seen on a VDU screen calibrated in millimetres (see also PS DE 3150249, OS DE 3542200). (orig./HP)

  18. 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

  19. Testing experimental fuel elements of the BN-600 fuel element type up to various depth of burn up in the BOR-60 reactor

    International Nuclear Information System (INIS)

    Results of the investigation of experimental fuel elements are presented. The authors discuss fuel element construction, basic testing parameters, results of measuring gas release from fuel, deformation of cladding and swelling of steel, and also data on material investigations of macro- an micro-structures of fuel and cladding with an analysis of the degree and character of their physico-chemical interaction with fission fragments

  20. Nuclear reactor with a reactor core composed of fuel elements

    International Nuclear Information System (INIS)

    A tube surrounding a fuel element projects above the liquid level. The tube is situated in a pot, whose upper edge lies between the top of the reactor core and the liquid level. A greater pressure is therefore produced, which ensures a reduction of the steam bubble proportion in the cooling liquid at the other fuel elements. (orig./HP)

  1. 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

  2. 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

  3. 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

  4. 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.

  5. Fuel element handling equipment for nuclear reactor

    International Nuclear Information System (INIS)

    The present device allows the handling of the fuel elements of a PWR type reactor when they are put in the cooling pool and when they are placed in the lead casks. The handling device includes a vertical arm, which comprises a telescopic assembly. The lower part of the telescopic assembly can slide axially, along the upper part between a retired position and a deployment position, in which the grab is at the level of the head of a fuel element in the pool or in the transport casks respectively. The grab can only be opened when it is at one of the extreme positions of the telescopic

  6. Environmental monitoring in the area of the Cadarache Nuclear Studies Centre

    International Nuclear Information System (INIS)

    The Cadarache Centre is part of the French nuclear programme providing for extensive development of research and hence of personnel and installations. It is essentially a centre for testing prototype reactors intended for use in power stations or for ship propulsion. In addition, experimental studies are carried out in connection with reactor structural materials and fuel elements. Various studies related to reactor physics, radiation protection, plant biology, etc. are also undertaken. This article describes the Cadarache site and the monitoring programme to assess the levels of radioactivity released from it.

  7. Nuclear reactor and associated fuel element

    International Nuclear Information System (INIS)

    Nuclear reactor with a high instantaneous negative reactivity temperature coefficient, comprising a vessel containing a certain quantity of water serving as coolant and moderator, a reactor core immersed in this water and comprising a series of fuel assemblies. Each fuel element contains a solid homogeneous mixture of zirconium hydride, uranium and erbium, in which the uranium constitutes 20 to 50% of the mixture by weight, the zirconium hydride 70 to 50% by weight and the erbium 0.5 to 1.5% by weight, the uranium present in the mixture being not more than 20% of U-235, the remainder being mostly U-238. The ratio of hydrogen/zirconium atom numbers is between 1.5/1 and 1.7/1 and the erbium is evenly distributed in the entire uranium-zirconium hydride mixture

  8. Experimental results of the CORA test program on the LWR fuel element behavior in severe reactor accidents

    International Nuclear Information System (INIS)

    In the framework of the CORA program the chemical interactions among fuel element (core) materials that may occur with increasing temperature up to complete melting have been examined. The high-temperature material behavior of PWR, BWR, and VVER-1000 fuel rod bundles has been studied in large-scale integral experiments and extensive separate-effects tests. In many cases, the reaction products are liquid at temperatures above 1200 C or have lower eutectic melting points than their original components. This results in a relocation of liquefied components, often far below their original melting points. Control rod materials can separate from fuel materials by a non-coherent stage-by-stage relocation process; this may cause recriticality problems during flooding of a partially degraded core with unborated water. Similarly, molten unoxidized Zircaloy cladding can relocate away from the decladded UO2 fuel rods. Significant relocation of UO2 dissolved in molten unoxidized Zircaloy can begin at the Zircaloy melting temperature (1760 C), about 1000 K below the melting point of UO2. Quenching (flooding) of the degraded bundles results in locally enhanced Zircaloy/steam reactions causing a renewed temperature rise, a meltdown of materials, and an additional strong H2 generation. The experimental results have contributed substantially to the understanding of the high-temperature core material behavior in severe reactor accidents, and provided a unique data base for the development, improvement, and validation of material-behavior models and severe accident system codes. (orig.)

  9. Testing of fuel elements and fuel element management at Kahl experimental nuclear power plant (VAK)

    International Nuclear Information System (INIS)

    The report is a survey of the different combustion elements used in the nuclear test reactor VAK; it pays special attention to their constructional characteristics and irradiation behaviour. For the first time, the feedback of plutonium as far as a one-hundred-percent MOX reactor core was demonstrated, while gadolinium was tested as a combustible neutron absorber in fuel. Components for advanced reactors, the superheated steam reactor and the project for steam cooled fast breeders were successfully tested in a special experimental loop. Moreover, the in-core fuel management with the various strategies for improving fuel utilization is described and the disposal of the burned fuel elements examined, fuel elements for which a closed fuel cycle corresponding to one for recycling uranium and plutonium was available as early as the end of the sixties. (orig./HP)

  10. Fuel elements in the core of the reactor Pegase. Description, successive improvements, actual possibilities

    International Nuclear Information System (INIS)

    The core of the research reactor Pegase, in operation at the Cadarache Nuclear Research Centre since 1983, contains fuel elements made from rolled plates of an aluminium-enriched uranium alloy whose characteristics have been changed several times. This report describes the modifications which have been made to these fuel elements with a view both to improving the technical qualities of the reactor and to decreasing its operational costs. Special attention is paid to the neutron aspects of the topic and in particular to the problem of the long-term modification of the reactivity. The 1966 results (30 per cent burn-up associated with only slight movement of the control rods) are particularly satisfying and can probably still be improved in the future. (authors)

  11. 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

  12. 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)

  13. 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.

  14. CANDU fuel elements behaviour in the load following tests

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore [Instiute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Palleck, Steve [Sheridan Park Research-AECL, Mississauga, ON (Canada). Fuel Deisgn Branch

    2011-08-15

    Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the fuel element 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 test period. New LF tests are planed 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 TRIGA Research Reactor and their relation to CANDU fuel performance in LF conditions. (orig.)

  15. Evaluation of the damage resulting from fission fragments in the fuel elements of the ETRR-2 materials testing reactor

    International Nuclear Information System (INIS)

    The damage resulting from fission fragments in the fuel elements of the ETRR-2 Materials Testing Reactor (MTR) was evaluated using the TRIM-86 computer code. The effect of very low yield fission ions (Zn and Gd) and high yield ions (Mo and La) on the structural changes of the aluminum matrix for the enriched U3O8 dispersed fuel was studied. The kinetic energy of the highly ionized fission fragments Zn, Mo, La and Gd was evaluated to be 106. 93. 61 and 48 MeV respectively. The tracks and distributions of ions and recoils were recorded and the mean ranges were calculated. The recoil and vacancy distributions as a function of track depth was studied mainly for the high yield Mo and La fission ions. The results indicated that the average number of recoils and vacancies produced from one ion of Mo and/or La amounted to 367x104 for recoils and 3.61 x 104 and 6.55 x 104 for vacancies, respectively. The replacement collisions amounted to 4% and 1% for Mo and La ions. The Total energy to aluminum recoils resulting from Mo and La ions was 3.67 and 6.77 MeV. The corresponding energies per recoil are 82 eV and 100 eV. The displacement energy required to create vacancy-recoil pair was found to be 20 5 eV and and the minimum energies with recoils to create Frenkle pairs are 102.5 and 120.5 eV for Mo and La ions

  16. Preliminary Studies On Hot Sipping Test Method For Fuel Element Cladding Integrity Test For PUSPATI TRIGA Mark-II Reactor (RTP)

    International Nuclear Information System (INIS)

    After more than 30 years of operation, some of the RTP fuel elements have been burn-up up to 20 %. Based on the ageing factor and burn-up fractions, examinations should be conducted to determine the integrity of the fuel element cladding. The test results will be used to look at the performance of the fuel element as well as the possible of fission product release into the RTP pool water. Thus, this paper discussed on the preliminary studies of hot sipping test method for RTP fuel element cladding integrity test where the hot sipping test method is one of the non-destructive techniques to find out possible leakage on the fuel elements cladding by detecting the presence of fission product nuclides in fuel element soaking water after irradiation. (author)

  17. 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)

  18. Fundamental aspects of nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    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 UO2, 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 UO2, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies

  19. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

    Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behavior and physical requirements of operating cycle sequences and fueling strategies having practical use in the management of nuclear fuel. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and maneuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy, and which govern fueling decisions normally made by the fuel manager. It is also demonstrated in this application that the simple batch size effect is not valid for non-integer fueling strategies, even in the simplest sequence configurations, and that it systematically underestimates the fueling requirements of degenerate sequences in general

  20. Development of elements simulating the fuel elements of RBMK reactors and nuclear district heating stations

    International Nuclear Information System (INIS)

    The development of elements simulating fuel elements with indirect electric heating has been going on for 20 years but work on improvements and new designs continues and is important even at the present time. Research on the thermohydraulic processes in nuclear reactor accidents is the most important application of these simulating elements. When an element simulating a fuel element is constructed, three problems are to be solved simultaneously. The design must provide the required operational parameters, it must be reliable, and it must satisfy the criteria of the necessary modelling. Simulating elements designated for research on the processes which occur in the late stages of an accident involving loss of coolant work under heat flow conditions resembling the residual energy liberation of reactors and at a high shell temperature (up to 1473 K). The number of heating cycles should amount to several tens or hundreds of cycles. When elements simulating fuel elements are developed for these processes, it is most important in regard to the modelling that the volume heat capacities of the simulating element and the fuel element coincide. The technical parameters of elements simulating the fuel elements of RBMK reactors and nuclear district heating stations were determined on samples in water and in air. A sample with an active length of 2500 mm was tested in boiling water inside a large tank under a pressure of 0.1 MPa. A heat flow q = 620 kW/m2 was obtained at a voltage U = 113 V and a current I = 560 A; this heat flow is about equal to the medium heat flow for an RBMK-1000 fuel element and the maximum for the fuel elements of nuclear district heating stations. Tests on a sample having a 1000 mm long active part and three internal thermocouples were made in air. They confirmed that these simulating elements remain functional in multiple heating cycles of up to 800-1000 degrees C and in return to load zero

  1. 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

  2. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

    Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behaviour and physical requirements of operating cycle sequences and fueling strategies having practical use in fuel management. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and manoeuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy. Numerical evaluations of degenerate equilibrium cycle sequences are then performed for a typical PWR core, and accompanying fuel cycle costs are calculated. The impact of design and operational limits as constraints on the performance mappings for this reactor are also studied with respect to achieving improved cost performance from the once-through fuel cycle. The dynamics of transition cycle sequences are then examined using the generalized theory. Proof of the existence of non-degenerate equilibrium cycle sequences is presented when the mechanics of the fixed reload batch size strategy are developed analytically for transition sequences. Finally, an analysis of the fixed reload enrichment strategy demonstrates the potential for convergence of the transition sequence to a fully degenerate equilibrium sequence. (author)

  3. Sipping tests on a failed irradiated MTR fuel element

    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. (authors)

  4. 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

  5. Fuel elements for pulsed TRIGA research reactors

    International Nuclear Information System (INIS)

    TRIGA fuel was developed around the concept of inherent safety. A core composition was sought that had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature would automatically cause the power excursion to terminate before any core damage resulted. Experiments have demonstrated that zirconium hydride possesses a basic neutron-spectrum-hardening mechanism to produce the desired characteristic. Additional advantages include the facts that ZrH has a good heat capacity, that it results in relatively small core sizes and high flux values due to the high hydrogen content, that it has excellent fission-product retentivity and high chemical inertness in water at temperatures up to 1000C, and that it can be used effectively in a rugged fuel element size. Tens of thousands of routine pulses to the range of 500 to 8000C peak fuel temperatures have been performed with TRIGA fuel, and a core was pulse-heated to peak fuel temperatures in excess of 11000C for hundreds of pulses before a few elements exceeded the conservative tolerances on dimensional change

  6. Coherence of reactor design and fuel element design

    International Nuclear Information System (INIS)

    Its background of more than 25 years of experience makes Framatome the world's leading company in the design and sales of fuel elements for pressurized water reactors (PWR). In 1994, the fuel fabrication units were incorporated as subsidiaries, which further strengthens the company's position. The activities in the fuel sector comprise fuel element design, selection and sourcing of materials, fuel element fabrication, and the services associated with nuclear fuel. Design responsibility lies with the Design and sales Management, which closely cooperates with the engineers of the reactor plant for which the fuel elements are being designed, for fuel elements are inseparable parts of the respective reactors. The Design and Sales Management also has developed a complete line of services associated with fuel element inspection and repair. As far as fuel element sales are concerned, Framatome delivers the first core in order to be able to assume full responsibility vis-a-vis the customer for the performance of the nuclear steam supply system. Reloads are sold through the Fragema Association established by Framatome and Cogema. (orig.)

  7. Preliminary study or RSG-GAS reactor fuel element integrity

    International Nuclear Information System (INIS)

    After 8 years of operation, RSG-GAS was able to reach 15 cycles of reactor operation with 116 irradiated fuels, whereas 49 fuels were produced by NUKEM; and the other 67 were produced by PEBN-BATAN. At the 15Th cycles, it have been used 40 standard fuels and 8 control fuels (Forty standard fuels and eight control fuels have been used in the 15th core cycles). Several activities have been performed in the reactor, to investigate the fuel integrity, among of them are: .fuel visual test with under water camera, which the results were recorder in the video cassette, primary water quality test during, reactor operation, fuel failure detector system examination and compared the PIE results in the Radiometallurgy Installation (RMI). The results showed that the fuel integrity, before and after irradiation, have still good performance and the fission products have not been released yet

  8. The AVR high-temperature reactor - operating experience, storage and final disposal of spent fuel elements

    International Nuclear Information System (INIS)

    The AVR is the first power plant with helium-cooled HTR to use spherical fuel elements. The experimental reactor was in successful operation for 21 years. In the first years of operation the main aim was the demonstration of the technical feasibility of high-temperature reactors. Special importance was attached to the testing and behavior of the fuel elements. The AVR was decommissioned in late 1988 and approve 170,000 spent fuel elements of various designs and compositions have been discharged. HTR fuel element reprocessing is not economically viable. Final disposal of the fuel elements is therefore envisaged after several years of intermediate storage. 3 refs., 1 tab

  9. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  10. Performance tests for integral reactor nuclear fuel

    International Nuclear Information System (INIS)

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34∼38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc

  11. Sipping test on a failed MTR fuel element

    International Nuclear Information System (INIS)

    This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes 131 I and 133 I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U3 O8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs. (author)

  12. The Calculation Of Total Radioactivity Of Kartini Reactor Fuel Element

    International Nuclear Information System (INIS)

    The total radioactivity of Kartini reactor fuel element has been calculated by using ORIGEN2. In this case, the total radioactivity is the sum of alpha, beta, and gamma radioactivity from activation products nuclides, actinide nuclides and fission products nuclides in the fuel element. The calculation was based on irradiation history of fuel in the reactor core. The fuel element no 3203 has location history at D, E, and F core zone. The result is expressed in graphics form of total radioactivity and photon radiations as function of irradiation time and decay time. It can be concluded that the Kartini reactor fuel element in zone D, E, and F has total radioactivity range from 10 Curie to 3000 Curie. This range is for radioactivity after decaying for 84 days and that after reactor shut down. This radioactivity is happened in the fuel element for every reactor operation and decayed until the fuel burn up reach 39.31 MWh. The total radioactivity emitted photon at the power of 0.02 Watt until 10 Watt

  13. RITM device for fuel element testing under power ramping

    International Nuclear Information System (INIS)

    The RITM device for studying different aspects of nuclear fuel behavior under power ramping while testing fuel elements in the SM-2 reactor is designed and tested. An irradiation rig of the device permits to conduct simultaneous irradiation of three fuel element located in individual cooling channels. Thermal neutron flux density in the rig cells varies within 0.25-1.00 of the maximum value. The rate of fuel power increase in the 0.25-1.00 and 0.5-1.0 ranges equals 3-5 and 4-12% min

  14. 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)

  15. Migration behaviour of fission products in and from spherical HTR fuel elements (SAPHIR loop experiments in the PEGASE reactor)

    International Nuclear Information System (INIS)

    The diffusion behaviour of some metallic fission products was evaluated for the irradiation experiments SAPHIR 4-10 and 11 performed by KFA Juelich in the PEGASE reactor, Cadarache (France). Diffusion coefficients for cesium in PyC coatings and for cesium and silver in graphitic fuel element matrix were obtained by analysing the fission product concentration profiles. The reason for the observed different diffusion behaviour of Cs 134 and Cs 137 is discussed as well as that for elevated concentrations of cesium, silver, and ruthenium near the fuel element surface. A simple model for assessing the trapping of metals in a graphite matrix is introduced. Release from the fuel element surfaces into the surrounding gas was found to be governed by direct recoil at low temperatures. (orig.)

  16. Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

    International Nuclear Information System (INIS)

    In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples

  17. Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

    Energy Technology Data Exchange (ETDEWEB)

    Kjellstroem, B.; Larsson, A.E.

    1967-04-15

    In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples.

  18. Handling system for nuclear reactor fuel and reflector elements

    International Nuclear Information System (INIS)

    A system for canning, inspecting and transferring to a storage area fuel and reflector elements from a nuclear reactor is described. The canning mechanism operates in a sealed gaseous environment and visual and mechanical inspection of the elements is possible by an operator from a remote shielded area. (UK)

  19. The AVR as a test bed for fuel elements

    International Nuclear Information System (INIS)

    One of the important tasks of the AVR experimental power-station was the testing of the spherical fuel elements, which had been newly developed and were used for the first time here. This testing in the AVR differs from the previous irradiation tests in material testing reactors by the fact that fuel elements from mass production were used here in large numbers. It took place in the genuine operating conditions of a nuclear powerstation. This included particularly the mechanical stesses due to fuelling equipment, the chemical interactions with the impurities of the cooling gas, accelerated by the catalytic effect of fission products. This also included the charge of temperature and power due to load changes of the powerstation and due to the fuel elements passing through the reactor several times. (orig.)

  20. Design of the Fuel Element for the RRR Reactor (Australia)

    International Nuclear Information System (INIS)

    The supply to the Replacement Research Reactor ( RRR ) to Australia represents a technological goal for our country, as much for the designers and manufacturers of this irradiation facility ( Invap SE ), as well for the responsibles of the fuel elements ( FE ) design and the suppliers of the first core ( CNEA ).In relation with the FE, although the conceptual design and fabrication technology of the FE are similar to the just developed and qualified by CNEA ( plane plates MTR fuel type ), the characteristics of this new reactor imposes most severe operation conditions on them than in previous supplies.In that sense, two distinguishing characteristics deserve to be shown: a) The magnitude of the hydrodynamics loads acting on the FE due to the coolant ascendent flow direction, and mainly, the very high flow velocities between the fuel plates ( aproximately five times higher than which presents in others Argentine FE actually in operation. b) The use of U3Si2 as fuel material.CNEA has started a programme to qualify this type of fuel.As result of these higher loads under irradiations and with the objective to maintain the high reliability level reached by our FE ( very low failure rates ), it was necessary to introduce FE mechanical-structural design modifications respect to the ECBE or standard design version, and to verify these changes through hydrodynamics tests on a 1:1 scale prototype.In this paper it is described the mechanical-structural FE design with special emphasis in the innovatives aspects incorporated.The design criteria established in function of the solicitations and limitating effects present under irradiation conditions.Also, a brief description of the proposed programme to verify and evaluate this design is presented, including analytical and numerical calculus of stresses acting on the fuel plates and others FE components, pressure loss hydrodynamics tests and endurance essays

  1. Management of Rossendorf research reactor spent fuel elements

    International Nuclear Information System (INIS)

    At the Rossendorf site, spent fuel elements have been in storage since 1957, at latest a total no. of 951. Transfer of spent fuel elements into CASTOR MTR 2 casks is the first major step of decommissioning of RFR. This paper will shortly describe the reactor, the fuel elements, their present storage, the loading procedure into CASTOR MTR 2 casks and a short-time storage at the Rossendorf site. At the beginning of this year the loading of the casks begun. The final aim is to transfer the loaded CASTOR MTR 2 casks to the Ahaus interim storage facility. (author)

  2. Design verification testing for fuel element type CAREM

    International Nuclear Information System (INIS)

    The hydraulic and hydrodynamic characterization tests are part of the design verification process of a nuclear fuel element prototype and its components. These tests are performed in a low pressure and temperature facility. The tests requires the definition of the simulation parameters for setting the test conditions, the results evaluation to feedback mathematical models, extrapolated the results to reactor conditions and finally to decide the acceptability of the tested prototype. (author)

  3. Progress of the Russian RERTR program: Development of new-type fuel elements for Russian-built research reactors

    International Nuclear Information System (INIS)

    The new design of pin-type fuel elements and fuel assembly on their basis for Russian research reactors has been developed. The number of following activities has been performed: computational and experimental substantiation of fuel element design; development of fabrication process of fuel elements; manufacturing of experimental assembly for lifetime in-pile tests. The relevant fuel assemblies are considered to be perspective for usage as low-enriched fuel for Russian research reactors. (author)

  4. Testing the surface contamination resuspension of a fuel element

    International Nuclear Information System (INIS)

    The aim of the tests is to verify if radioactive aerosols can be resuspended in the atmosphere after surface contamination of a fuel plate. These tests are part of a program for dry storage of fuel plates without container. Tests are realized in a hot cell of OSIRIS reactor in a special device. The tested element is placed in a container and compressed air sweep the surface at a speed of about 5 m/s. Sampling on a filter placed at the outlet is used for analysis of air flowing between fuel plates. Nature and activity of products are determined by gamma spectrometry and found negligible

  5. 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.

  6. Irradiation of Fuel Elements in the Belgian BR3 Reactor

    International Nuclear Information System (INIS)

    Under a contract concluded by EURATOM and CEN-BelgoNucléaire, fuel rods containing plutonium-enriched uranium were irradiated in the Belgian BR3 reactor with the object of evaluating the behaviour of plutonium fuel elements in power reactors. The first experiment consisted in introducing 12 fuel elements fabricated by vibration and compacting followed by swaging into a core assembly of the BR3 pressurized-water power reactor. Irradiation was carried out for a period corresponding to 4820 h at full power. Subsequent examination of the fuel rods showed that they had been unaffected by irradiation. A second series of experiments is being carried out in collaboration with the United Kingdom Atomic Energy Authority. These experiments involve irradiating an assembly of 37 plutonium-enriched fuel elements, some compacted and others of the pellet type, in the BR3/VN power reactor. The fabrication of the vibrocompacted elements and the thermal studies relating to the assembly are briefly described. (author)

  7. 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)

  8. 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)

  9. METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    Science.gov (United States)

    Layer, E.H. Jr.; Peet, C.S.

    1962-01-23

    A method is given for preparing a fuel element for a nuclear reactor. The method includes the steps of sandblasting a body of uranium dioxide to roughen the surface thereof, depositing a thin layer of carbon thereon by thermal decomposition of methane, and cladding the uranium dioxide body with zirconium by gas pressure bonding. (AEC)

  10. Methodology and results of operational calculations of fuel temperature in fuel elements of the BN-600 reactor fuel assemblies

    International Nuclear Information System (INIS)

    The article presents methodology of peak fuel temperature determination and computational investigations of fuel temperature condition in fuel elements of fuel assemblies of various types during the BN-600 reactor operation. The effect of sodium uranate in the gap between fuel and cladding of the fuel element on the heat transfer processes is considered

  11. Storage rack for reactor fuel elements

    International Nuclear Information System (INIS)

    Each FE-position of quadriform cross-section is provided with two diagonally arranged centering bodies. They serve as guide of a refuelling machine associated to the storage rack of the reactor- or waste disposal plant. The refuelling machine has centering fingers that fit in the centering body. (DG)

  12. Commercial Aspect of Research Reactor Fuel Element Production

    International Nuclear Information System (INIS)

    Several aspects affecting the commercialization of the Research Reactor Fuel Element Production Installation (RR FEPI) under a BUMN (state-owned company)have been studied. The break event point (BEP) value based on total production cost used is greatly depending upon the unit selling price of the fuel element. At a selling price of USD 43,500/fuel element, the results of analysis shows that the BEP will be reached at 51% of minimum available capacity. At a selling price of US$ 43.500/fuel element the total income (after tax) for 7 years ahead is US $ 4.620.191,- The net present value in this study has a positive value is equal to US $ 2.827.527,- the internal rate of return will be 18% which is higher than normal the bank interest rare (in US dollar) at this time. It is concluded therefore that the nuclear research reactor fuel element produced by state-owned company BUMN has a good prospect to be sold commercially

  13. 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)

  14. Testing of cermet fuel pins in the MIR reactor (intermediate results)

    International Nuclear Information System (INIS)

    Fuel elements with cermet fuel are designed and the technology for their manufacturing is developed. The fuel elements of two types are tested in the MIR reactor: 1) the elements with metallurgical provision of a cermet fuel pin-cladding contact (brazing); 2) the elements with a cladding tightly pressed to a cermet pin (rolling). The test results show successful operation of the first type fuel elements with the fuel consisting of uranium dioxide powder, dispersed in a matrix of zirconium base or aluminium base alloy. All fuel elements of the second type appear to be leaking

  15. Reactor fuel element heat conduction via numerical Laplace transform inversion

    Energy Technology Data Exchange (ETDEWEB)

    Ganapol, Barry D.; Furfaro, Roberto [University of Arizona, Tucson, AZ (United States). Dept. of Aerospace and Mechanical Engineering], e-mail: ganapol@cowboy.ame.arizona.edu

    2001-07-01

    A newly developed numerical Laplace transform inversion (NLTI) will be presented to determine the transient temperature distribution within a nuclear reactor fuel element. The NLTI considered in this presentation has evolved to its present state over the past 10 years of application. The methodology adopted is one that relies on acceleration of the convergence of an infinite series towards its limit. The inversion will be applied to the prediction of the transient temperature distribution within an MTR type nuclear fuel element through a novel formulation of the solution to the transformed heat conduction equation. (author)

  16. 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

  17. Impact tests with fuel element cans

    International Nuclear Information System (INIS)

    Impact tests with storage tanks for irradiated HTR-fuel balls have been carried out. The determination of the damages of the storage tanks falling from a heigth of 7 m and the graphite balls, which have been used in place of the fuel elements, has been the aim of these tests. The main results are: 1. The leakage of the three impact tested tanks (2 of type ASSE, 1 of type AVR-TL) has not increased due to the impact. 2. The deformation of the tanks caused by the impact exceed the tank specification for dimensions and shape. 3. The graphit ball damages depend on the type of tank and on the angle of impact. The damages of graphit balls in the tank of type AVR-TL have been neglectable small. (orig.)

  18. 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.

  19. 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)

  20. Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

    International Nuclear Information System (INIS)

    The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

  1. 77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Science.gov (United States)

    2012-03-22

    ... COMMISSION Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors,'' is temporarily identified... verifying the quality of plate-type uranium-aluminum fuel elements used in research and test reactors...

  2. 78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Science.gov (United States)

    2013-06-03

    ... COMMISSION Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test... Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). ADDRESSES:...

  3. 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)

  4. 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

  5. Radial heat conduction in a power reactor fuel element

    International Nuclear Information System (INIS)

    Two radial conduction models, one for steady state and another for unsteady state, in a nuclear power reactor fuel element are developed. The objective is to obtain the temperatures in the fuel pellet and the cladding. The lumped-parameter hypothesis are adopted to represent the system. Both models are verified and their results are compared with similar ones. A method to calculate the conductance in the gap between the UO2 pellet and the clad and its associated uncertainty is included in the steady state model. (author)

  6. Compaction of spent fuel elements from light water reactors

    International Nuclear Information System (INIS)

    To reduce the expenditures required for shipping and interim storage of spent fuel elements, a compaction technique has been designed which can be applied to pressurized water and boiling water reactor fuels. The highly mechanized and automated procedure achieves high throughputs while requiring little manpower. For the waste management pathway with reprocessing this means considerable savings in the costs for shipping and interim storage over the life of a plant. There are other cost advantages, which are not the subject of this article. (orig.)

  7. Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

    Science.gov (United States)

    Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

    2013-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen 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. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

  8. Fuel Element Mechanical Design for CAREM-25 Reactor

    International Nuclear Information System (INIS)

    The Fuel Element mechanical design and spider-control reactivity and security rods assembly for the CAREM-25 reactor is introduced. The CAREM-25 Fuel Element has a hexagonal cross section with 127 positions, in a triangular arrangement.There are 108 positions for the fuel rods while the guide tubes and instrumentation tube occupy the 19 remaining positions.From the structural point of view, the fuel element is being composed by a framework formed by the guides and instrumentation tubes, 4 spacer grids and the upper and lower coupling pieces.The spider is a plane piece, with a central body and six radial branches in T form, which has holes where the absorber rods are fitted.The central body ends in a joint in the upper side, which allows connect the assembly whit the reactor control mechanisms.The absorber rods are made of a neutron absorber material (Ag-In-Cd) hermetically closed in a stainless steel cladding. In this work are determined, in addition to the basic design, the operational conditions, the functional requirements to be satisfied and in agreement with those, the adopted criteria and limits to avoid systematics failure during normal operation conditions. The proposed program for the verification and evaluation of design is detailed.To consolidate the design, a prototype was manufactures, based on drawings and specifications needed for its construction

  9. 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.)

  10. 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

  11. Possibilities of Kazakhstan experimental base for space nuclear reactors elements testing

    International Nuclear Information System (INIS)

    To the mid of 70-th in Kazakhstan the surface developing base for space nuclear reactors elements testing was created. The base consists of three test complexes. Two of them - the complexes of test reactors 'Baikal-1' and IGR - are situating on the Semipalatinsk test site, and the third one - complex of WWR-K research reactor - is situating in Alatau village nearby to Almaty city. On 'Baikal-1' and IGR complexes the testings for fuel elements, fuel assemblies, modules and prototypes of nuclear rocket engine reactor and nuclear energetic engine units with turbine-engine energy transmission on the base solid-phase reactor were carrying out. On the WWR-K reactor complex the testing of power generating channels of thermal-emission transmission reactors were conducted. In the paper the assessment of up-to-date experimental base status and it possibilities for further using in space nuclear energy field are given

  12. Development of a transport cask for spent fuel elements of research reactors

    International Nuclear Information System (INIS)

    This article presents an overview of the development of a research reactor spent fuel transport cask. Through a project funded by the IAEA, Argentina, Brazil and Chile have collaborated to enhance regional capacity in the management of spent fuel elements from research reactors operated in the region. A packaging for the transport of research reactors spent fuel was developed. It was designed by a team of researchers from the countries mentioned and a 1:2 scale model for MTR type fuel was constructed in Argentina and subsequently tested in CDTN facilities in Belo Horizonte, Brazil. There were three test sequences to test the cask for normal transport and hypothetical accident conditions. It has successfully passed the tests and the overall performance was considered satisfactory. As part of the licensing process, a test sequence with the presence of regulatory authorities is scheduled for December, 2012 (author)

  13. The design of a fuel element for the RA-3 reactor (Ezeiza Atomic Center)

    International Nuclear Information System (INIS)

    Some features of the mechanical design of the low enrichment fuel element for the RA-3 reactor are described, with emphasis in those aspects of the original design that have been modified considering the experience acquired in the design of other fuel elements. The proposed modification is based fundamentally on the replacement of all welded joints by screwed joints, which facilitates the manufacture of the fuel element, avoiding the distortions produced by the welds used at present and contributing to the fulfillment of the foreseen tolerances. A basic characteristic of this design is a careful manufacture of the fuel element's structural components in order to assure an assembling of the fuel element that fulfills the tolerances intrinsically required. The fuel is designed for the RA-3 reactor and uses U3O8 or U3Si2 as carrying phase of the fissile material with an enrichment of 19.70% of 235U. The design verification was performed by analytical and numerical methods, and is supported by testing of materials in laboratory, hydrodynamics tests and performance evaluations of the fuel elements in the RA-3 reactor. (author)

  14. Burnup determination of power reactor fuel elements by gamma spectrometry

    International Nuclear Information System (INIS)

    This report describes a method for determining by γ spectrometry the burn up and the specific power of fuel elements irradiated in power reactors. The energy spectrum of γ rays emitted by fission products is measured by means of a simple equipment using a sodium iodide detector and a multichannel analyzer. In order to extract from the spectrum a quantity proportional to the burn up, it is necessary to: - isolate an activity specific of one emitter,- give the same importance to fissions in uranium and plutonium - take into account the radioactive decay during and after irradiation. One hundred fuel elements were studied and burn up values obtained by γ spectrometry are compared to results given by chemical analyses. Preliminary measurements show that the accuracy of the results is greatly increased by the use of a germanium detector, due to its good resolution. (authors)

  15. 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

  16. 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)

  17. 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

  18. Fabrication of zero power reactor fuel elements containing 233U3O8 powder

    International Nuclear Information System (INIS)

    Oak Ridge National Laboratory, under contract with Argonne National Laboratory, completed the fabrication of 1743 fuel elements for use in their Zero Power Reactor. The contract also included recovery of 20 kg of 233U from rejected elements. This report describes the steps associated with conversion of purified uranyl nitrate (as solution) to U3O8 powder (suitable for fuel) and subsequent charging, sealing, decontamination, and testing of the fuel elements (packets) preparatory to shipment. The nuclear safety, radiation exposures, and quality assurance aspects of the program are discussed

  19. 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.

  20. 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

  1. 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.

  2. Experimental studies on coolant technology and fuel elements of high temperature reactors in the helium loop PG100 and in irradiation channels

    International Nuclear Information System (INIS)

    An important part of experimental researches in SSSR on fuel element fabrication and equipment of the primary coolant circuit of high temperature reactors are made on the following test benches: an installation for the study of leak-tight fuel elements (OSA installation), an helium loop (PG100) in the MR reactor, tube channels for testing ''Kashtan'' pebble fuels, tube channels for testing ''Karat'' fuel microelement. After testing fuel elements and equipment are examined in a hot laboratory

  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. 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)

  5. Reactors G2/G3: residual power of the fuel elements

    International Nuclear Information System (INIS)

    After describing the experimental method used an account is given of tests carried out on active fuel elements, either in air, or in groups of four placed in a discharge container. A comparison with the theoretical results is also made in the case of different irradiation and decay times. Information is obtained concerning: the residual power of the fuel elements, the can temperature which may be attained by these elements after extraction from the reactor, and the risk of ignition of the elements during normal discharging operations. (author)

  6. Irradiation of MEU and LEU test fuel elements in DR 3

    International Nuclear Information System (INIS)

    Irradiation of three MEU and three LEU fuel elements in the Danish reactor DR 3. Thermal and fast neutron flux density scans of the core have been made and the results, related to the U235-content of each fuel element, are compared with the values from HEU fuel elements. The test elements were taken to burn-up percentages of 50-60%. Reactivity values of the test elements at charge and at discharge have been measured and the values are compared with those of HEU fuel elements. (author)

  7. 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)

  8. 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)

  9. Integral needs for MOX powders: state of the art at Cea Cadarache on MOX fuel experiments

    International Nuclear Information System (INIS)

    Several experimental programmes have been conducted at CEA connected with the use of MOX in commercial PWR or innovative concepts, such as high conversion light water reactors (HCLWRs) and at various moderation ratios. Measurements of integral parameters of interest were performed using fission chambers and gamma-scanning techniques, or by oscillation techniques. This paper presents an overview of the experimental programmes performed at the EOLE and MINERVE facilities located at CEA Cadarache. (author)

  10. Status of HTR fuel element development in the FRG and its potential for nuclear process heat reactors

    International Nuclear Information System (INIS)

    In the FRG three basic fuel element designs for application in high-temperature gas cooled reactors are being persued: the spherical element, the graphite block element, and the moulded block element (monolith). This report gives the state of development reached with the three types of elements but also views their specific merits and performance margins and presents aspects of their future development potential for operating in advanced HTGR plants. The development of coated feed and breed particles for application in all HTGR fuel elements is treated in more detail. - Summarizing, it can be said that all the fuel elements as well as their components have proved their aptitude for the steam cycle systems in many fuel element and particle performance tests. Further testing is still necessary, however, to adapt these fuel elements and coated particles for advanced reactor concepts and to develop them up to full technical maturity. Ways of overcoming problems arising from the more stringent requirements are shown. (orig.)

  11. LWR Fuel Gas Characterization at CEA Cadarache LECA-STAR Hot Laboratory

    International Nuclear Information System (INIS)

    The aim to improve LWR fuel behaviour led CEA to improve its PIE capacities in term of test devices and characterization techniques in the shielded hot cells of the LECA-STAR facility, located in Cadarache. A presentation of these capacities is made, focusing on gas characterization : - annealing test devices used for transient simulations. These devices are test facilities allowing to simulate a very large range of conditions (temperature, atmosphere, pressure etc); - characterization techniques : optical microscopy, for microstructure characterizations on polished samples; scanning electron microscopy (SEM) used both on polished samples and on fractographs; Electron Probe Micro-Analyzer (EPMA) for quantitative elementary analysis and elementary mappings; Secondary Ion Mass Spectrometer (SIMS), for isotopic analysis, mapping depth profiles and gas measurements. It shows an overview of their main applications for the study of fuel behaviour under nominal operating conditions but also under simulated accidental conditions and storage conditions. Two detailed examples of the use of these techniques working together on the same samples, are then presented, focusing on fission gas characterizations. The first one is a 72 GW.d.t-1 high burnup UO2 PWR fuel for which detailed characterizations have been performed before and after a LOCA type condition annealing test. The second one is a 35 GW.d.t-1 UO2 PWR fuel for which these characterizations have been done before and after a ramp test at 100 W.cm-1.min-1 up to a 90 s maximum power hold time at 520 W.cm-1. In both cases, the complementarities of these various techniques show the interest of such detailed characterizations. (author)

  12. 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)

  13. Method of testing fuel assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    The stresses occurring in the fuel assemblies are simulated by power excursions. For this purpose the fuel assembly is placed in the neutron field of a test reactor and for a short time can be exposed to the much higher neutron field of a pulsed reactor. One possibility of design provides for the test and the pulsed reactor lying one above the other, separated by a neutron absorber and penetrated by a common irradiation channel. The fuel assembly then is to be moved from the position in the test reactor to the position in the pulsed reactor. The other possibility is to make the irradiation duct pass along the gap between both reactors and, by means of a tube-shaped absorber, open one or the other irradiation field. (DG)

  14. 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)

  15. 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

  16. Testing of HTR UO2 TRISO fuels in AVR and in material test reactors

    International Nuclear Information System (INIS)

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO2 TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO2 TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO2 TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C

  17. 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)

  18. 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

  19. 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.

  20. Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

    International Nuclear Information System (INIS)

    Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO2SO4) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

  1. Qualification in the reactor Siloe of low enriched fuels for research and test reactors

    International Nuclear Information System (INIS)

    For nearly two years now, in the scope of the Reduced Enrichment Research and Test Reactor (RERTR) Program (CEA-ANL-CERCA Agreements), low enriched fuel has been irradiated in Siloe. In 1981 a complete 45% enriched fuel element (U Alx compound) was irradiated. A burn-up of 50% was obtained without any difficulty. Since June 1982 4 U3Si fuel plates are being irradiated. These plates, with a density of 5.5g of total uranium per cubic centimeter (two plates), and 6.0g per cubic centimeter (the other two plates) have already reached a mean burn-up of about 20% and their behaviour up till now is excellent. The fuel element and the plates have been manufactured by CERCA

  2. 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)

  3. Testing of research reactor fuel in the high flux reactor (Petten)

    International Nuclear Information System (INIS)

    The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium (3. However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

  4. Post reactor researches of fuel pins, tested under alternating NEMF reactor functioning modes

    International Nuclear Information System (INIS)

    Changing of rod ceramic fuel pins state under their exploitation conditions changing influence at alternating of three-mode nuclear energy-moving facility reactor functioning has been examined. There are presented the results of researches of fuel pins, tested in the reactor IRGIT and RA, firstly under moving mode, then - under energy mode of minor power of NEMF reactor. (author)

  5. 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.

  6. 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

  7. Analysis of possibilities for functional capacity for work rise of reactor fuel elements at nuclear engine regime

    International Nuclear Information System (INIS)

    The principle results of carbide fuel rods testing during series of IVG.1 reactor starts up at regime simulating nuclear engine regime of nuclear moving power unit are given. Considerable degradation of initial fuel elements status increasing from start up to start up and which could resulted fail of separate technological channels is shown. Origin case of extreme degradation of fuel elements status are insufficient thermal strength of fuel elements operation in the field brittle state of sintered carbide material, Possible ways of artificial reinforce of fuel elements of low temperature sections, increasing its thermal strength up to required level

  8. 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%

  9. 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

  10. 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

  11. Development and performance of fuel elements for sodium-cooled breeder reactors in Germany

    International Nuclear Information System (INIS)

    The first sodium-cooled reactor commissioned in Germany, KNK, serves now as test facility for plutonium bearing oxide fuel elements. The target is to provide reliable fuel for the SNR-300 project (Kalkar Nuclear Power Plant). The long-range target is fuel for burnups above 100,000 MW d/t, which moreover can easily be fabricated and reprocessed. As in the U.K., the line of grid-spaced bundles is favorised, being promising as regards the possibility of replacement of a defected pin and reinsertion of the bundle. (orig.)

  12. 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

  13. Sipping test of fuel assemblies in LVR-15 reactor

    International Nuclear Information System (INIS)

    The LVR-15 reactor is a light water research type which is situated at NRI in Rez near Prague. The poster describes the procedure and methodology used for sipping test of the fuel assemblies. These tests are designed to evaluate the leakage of fuel and fission products from the tested fuel assembly. From 1995 to 2003 there have been performed about 200 tests. Examples of results of sipping water activity measurements are presented. The values of activities of 137Cs and 134Cs are used for decision if the fuel assembly can be used in reactor core, transported to storage pool or if it is necessary to put the fuel assembly into the special protective can. The used limits of activities are discussed. (author)

  14. 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)

  15. Fuel-to-cladding heat transfer coefficient into reactor fuel element

    International Nuclear Information System (INIS)

    Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. (orig.)

  16. Application of CO2 laser beam weld for repair of fuel element of nuclear reactor 'YAYOI'

    International Nuclear Information System (INIS)

    The present studies are to develop CO2 laser beam welding techniques in order to apply for repoint of nuclear reactor fuel of Fast Neutron Source Reactor YAYOI. For that purpos, many experiments were conduted to obtain various effects of laser welding variables with use of SUS 304 plates, pipes and simulated dumy fuels. These experiments provided us an optimal welding condition through metallurgical observations, non-destructive and mechanical tests. It was found that the laser welds exhibited properties equivalent to those of the base metal, in addition they provided us a favorable system than that of electron beam welds against a cladding of radioactive nuclear fuel in a hot cell. The present paper reports on the characteristics of laser welds, structural analysis of fuel element and a system design of remotely operated devices setting in a hot cell. (author)

  17. 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)

  18. 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)

  19. 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)

  20. 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

  1. Fuel element burn-up calculation in ITU TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    The reactivity defect of fuel elements in ITU TRIGA Mark-II reactor core at 250 kW power have been calculated by considering the reactor operation history. A two-dimensional, four-group diffusion computer code TRIGLAV is used for the calculations. The unit-cell macroscopic cross sections and diffusion coefficients are generated with the WIMS-D/4 code. Two dimensional effects like vicinity of control rods, water gaps, dummy graphite elements, void channels are considered. The calculated reactivity worth of the fuel elements at known burn up are in agreement with experimental values of the fuel elements located in the reactor core without two dimensional effects. (author)

  2. 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)

  3. Development and introduction of automated lines for fabrication of vibrocompacted fuel elements for BN-600 reactor

    International Nuclear Information System (INIS)

    In the framework of international program modernization of technological complex for fabrication granular, suitable for vibration compacting fuel, fuel elements and fuel assemblies is realized. The aim of modernization is to provide BN-600 reactor with MOX fuel on the basis of weapon plutonium

  4. Conversion and evaluation of the THOR reactor core to TRIGA fuel elements

    International Nuclear Information System (INIS)

    The THOR reactor is a pool type 1 MW research reactor and has been operated since 1961. The original MTR fuel elements have been gradually replaced by TRIGA fuel elements since 1977 and the conversion completed in 1987. The calculations were performed for various core configurations by using computer codes, WIMS/CITATION. The computing results have been evaluated and compared with the core measurements after the fuel conversion. The analysis results are in good correspondence with the measurements. (author)

  5. 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.

  6. 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

  7. Fuel-element simulator for investigating thermal-hydraulic accidents in water-water reactors

    International Nuclear Information System (INIS)

    A fuel-element simulator should provide the necessary environmental parameters (thermal flux, and temperature at the cladding surface) and satisfy the requirements of reliability and modeling an actual fuel element, according to a formulated research problem. A universal simulator design, which could be used in a wide range of research, does not exist up to now and it is hardly useful in general. In developing fuel-element simulators to study loss-of-coolant accidents in water-water reactors, the most important condition from the modeling point of view is that the overall heat capacity of the simulator should correspond to that of the fuel element. The overall heat capacity and the temperature distribution over the reactor cross section determine the reserve of accumulated energy, which cannot be modeled by simply increasing the supplied electrical power. Experiments showed the magnesium oxide, as compared to other materials, is the best model of uranium oxide due to the closeness of the heat transfer coefficient and the thermal conductivity of these materials. Moreover, MgO has a high coefficient of thermal expansion, close to that of stainless steel. The construction of fuel-element simulators often uses boron nitride powder, which is densified by one means or another. Boron nitride has the highest thermal conductivity (besides beryllium oxide), but it has a lower electrical conductivity than magnesium oxide. These materials simultaneously fulfill the function of electrically insulating the heating element from the cladding. The basic disadvantage of this design is that the simulator has no gas gap; however, this is compensated by its simplicity, reliability, and long lifetime. This article presents several test designs for analysis and solving problems characteristic of loss-of-coolant accidents. Test results from VVER-440 fuel rod simulators using 19-rod assemblies an presented

  8. 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.

  9. 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.

  10. 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.

  11. 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...

  12. Safety analysis of spent fuel element storage in 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Approximately 90000 spent fuel elements will be discharged from a 10 MW high temperature gas-cooled reactor (HTR-10) in its lifetime. The activity of the radioactive fission products in these spent fuel elements will reach 1.0 x 1016 Bq, so these spent fuel elements should be properly managed. HTR-10 spent fuel elements will be discharged into lead-steel containers, with each container designed to receive 2000 fuel elements. These containers will be stored in a concrete compartment inside the reactor building and cooled by air. The author analyzes the release of the radioactive nuclides, the critical safety parameters and the irradiation shielding. The results show that the safety requirements can be met in the HTR-10 spent fuel element storage compartment

  13. 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

  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. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  16. Analysis of principle possibilities of intermediare storage of fast breeder reactor fuel elements

    International Nuclear Information System (INIS)

    The principle possibilities of intermediate storage of fast breeder reactor fuel elements were analyzed and compared on the basis of 4 different concepts of storage. The SNR-2 fuel element was chosen as reference. Only the pool (wet) storage could be used to store fuel elements of less than 18 months precooling time. The other concepts (dry storage and container storage) have distinct advantages at precooling times longer than 18 months. (orig./HP) With 22 tabs., 8 figs

  17. Water flow characteristics of Baumkuchen type fuel elements for Kyoto University high neutron flux reactor

    International Nuclear Information System (INIS)

    The Kyoto University high neutron flux reactor is a light water-moderated and cooled, divided core type reactor with heavy water reflector. In the core, six inside fuel elements and twelve outside fuel elements are arranged in double ring form, and two cylindrical, divided cores are placed at 15 cm distance. The flow rate distribution and pressure loss in the fuel elements constitute the base of the thermo-hydraulic design of the core, therefore the model fuel elements of full size were made, and the water flow experiment was carried out to examine their characteristics. It was found that the flow velocity in channels was strongly affected by the accuracy of channel gaps. The calculation of pressure loss in fuel elements, the experiments on inside fuel elements and outside fuel elements, and the results of experiments such as the calibration of the cooling channels in outside fuel elements, the relation between total flow rate and pressure loss, and the characteristics of flow at the time of reverse flow are reported. The general characteristics of flow in fuel elements were in good agreement with the prediction. In the pressure loss in fuel elements, the friction between fuel plates and the resistance of nozzles were the controlling factors under the rated operating conditions of the HFR. (Kako, I.)

  18. Postirradiation examination and evaluation of Peach Bottom fuel test elements FTE-14 and FTE-15

    International Nuclear Information System (INIS)

    Peach Bottom fuel test elements FTE-14 and FTE-15 were companion nonaccelerated tests of fuel rods and fuel particles representative of the Large High-Temperature Gas-Cooled Reactor (LHTGR). The purpose of the tests was to broaden the data base of H-327 graphite and various fuel types; specifically, UO2, UC2, weak acid resin UC/sub x//O/sub y/, and several fertile fuel types were tested. The irradiation reached peak fuel temperatures of 16000C volume- and time-averaged temperatures of 13000C, and fast fluence exposures up to 2 x 1025 n/m2 (E > 29 fJ)/sub HTGR/. Experimental results were compared with predictions based on accelerated irradiation tests, postirradiation heating, and other Peach Bottom test elements to validate HTGR design codes. The nuclear design predictions were modified by measurements which allowed the verification of thermal design calculations and thermocouple readings

  19. 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)

  20. 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)

  1. Results of out-of-pile tests of the MIR reactor irradiated fuel at high temperatures

    International Nuclear Information System (INIS)

    In the paper are presented the results of out-of-pile tests of the MIR reactor irradiated fuel elements, to determine blistering initiation temperature. Fuel elements were made by extrusion method and had a round tube shape. Fuel meat - 90% enriched uranium dioxide particles, dispersed in an aluminum matrix, with density of uranium ∼ 1 g/cm3, fuel elements cladding material - aluminum alloy SAV-6. A standard MIR reactor fuel element assembly consists of 4 round shape tubes, concentrically inserted in one another, with their outside diameters being 70, 61, 52, 43 mm correspondingly, cladding thickness being ∼ 0.72 mm, fuel meat thickness ∼ 0,56 mm, core height ∼ 1000 mm. Fuel gaseous swelling with hemispheric ∼ (2/5) mm blisters formation on the fuel elements surface starts at temperatures ∼ (450/480) deg. C. Porosity formation and fuel meat cracking started at the boundary of diffusion fuel-matrix interaction layer and matrix material out of fission products, accumulated and located in outside of fuel particles. (author)

  2. 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)

  3. Fuel element container for transporting and/or storing nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    The container consists of cast iron with spheroidal graphite for transporting and/or storing irradiated fuel elements. The front opening is closed so as to be gastight by a lid. In order to be able to weld the container after the lid is fitted, without any subsequent heat treatment being necessary, a ring made of material which can be cold welded is melted on the end of the container forming the opening when casting it via a connecting section. After loading it, the ring can be cold welded to a lid with a similar structure. (orig.)

  4. The key role of critical mock-up facilities for neutronic physics assessment of advanced reactors: an overview of Cea Cadarache tools

    International Nuclear Information System (INIS)

    The Experimental Physics section of CEA Cadarache operates three critical facilities devoted to neutronic studies of advanced reactors (EOLE, MINERVE and MASURCA) covering a large scope of interests. These include 100% MOX core in ABWR qualification, knowledge improvement of basic nuclear data for heavy nuclides for new options of the fuel cycle - especially the multi-recycling of plutonium - and accelerator-driven systems neutronic behaviour for transmutation studies. The paper describes these facilities, the scientific programmes associated and the progressive improvement of experimental techniques, the aim being to significantly reduce the uncertainties regarding the evaluation of the physical parameters. (authors)

  5. 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)

  6. Analysis of cocked fuel elements in the AFRRI TRIGA Mark-F reactor

    International Nuclear Information System (INIS)

    The Armed Forces Radiobiology Research Institute (AFRRI) TRIGA Mark-F pulsing reactor has experienced eight cocked fuel elements during the period 5 November 1974 through 17 February 1982. Although there are no adverse health and safety consequences associated with their occurrence and there is no credible potential for system damage, cocked TRIGA fuel elements do cause inconvenience to the reactor staff and a temporary delay in operations. This paper presents the history of cocked TRIGA fuel elements at AFRRI, discusses possible mechanisms for their occurrence, and outlines a plan to isolate and ultimately determine their actual cause

  7. 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

  8. 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.)

  9. 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)

  10. 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

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

    International Nuclear Information System (INIS)

    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

  12. Process for locating leaking cans of fuel elements of a nuclear reactor

    International Nuclear Information System (INIS)

    A mixture of He, Ne 20 and Ne 22 gases is introduced into the fuel element cans of the sodium-cooled reactor before setting to work. Defective cans can be located by continuous mass spectroscopic monitoring of the reactor protective gas for these indicator gases. (DG)

  13. Postirradiation examination of Peach Bottom fuel test element FTE-4

    International Nuclear Information System (INIS)

    The report presents the irradiation results and their evaluation for Peach Bottom fuel test element FTE-4. It describes in detail the efforts by General Atomic Company over the last two years to establish a system for extracting meaningful performance information from a fuel test element. This has been done with the goal of making direct comparisons between as-measured data and core design code predictions. Special emphasis has been placed on determining the 95% confidence limits on most of the preirradiation and postirradiation measurements in order to allow a better comparison with GAUGE, FEVER, and TREVER code calculations which are used in HTGR core thermal and mechanical design

  14. 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)

  15. 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.

  16. Modeling and Simulation of a Nuclear Fuel Element Test Section

    Science.gov (United States)

    Moran, Robert P.; Emrich, William

    2011-01-01

    "The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

  17. Some aspects of statistic evaluation of fast reactor fuel element reliability

    International Nuclear Information System (INIS)

    Certain aspects of application of statistical methods in forecasting operating ability of fuel elements of fast reactors with liquid-metal-heat-carriers are considered. Results of statistical analysis of fuel element operating ability with oxide fuel (U, Pu)O2 under stationary regime of fast power reactor capacity are given. The analysis carried out permits to single out the main parameters, considerably affecting the calculated determination of fuel element operating ability. It is shown that parameters which introduce the greatest uncertainty are: steel creep rate - up to 30%; steel swelling - up to 20%; fuel ceep rate - up to 30%, fuel swelling - up to 20%, the coating material corrosion - up to 15%; contact conductivity of the fuel-coating gap - up to 10%. Contribution of these parameters in every given case is different depending on the construction, operation conditions and fuel element cross section considered. Contribution of the coating temperature uncertainty to the total dispersion does not exceed several per cent. It is shown that for the given reactor operation conditions the number of fuel elements depressurized increases with the burn out almost exponentially, starting from the burn out higher than 7% of heavy atoms

  18. Use of fuel elements and fuel rod arrays of WWER-type with 20 % enriched cermet fuel for reactors of floating power plant KLT-40S

    International Nuclear Information System (INIS)

    It was carried out numerical analysis of the physical characteristics of change from normal active zone to fuel elements and fuel rod arrays using fuel cycle of WWER-1000 type as well as at replacement of oxide fuel to cermet fuel (60%UO2+40% of silumin) with 20% enrichment. At that the main physical characteristics of active zone and reactor are kept - geometric sizes, power, coolant properties etc. It was given the main physical properties of fuel elements and fuel rod arrays of active zone with cermet fuel. Calculation of neutron physical characteristics was carried out. The reactor has internal self-protectability

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

    International Nuclear Information System (INIS)

    In the fuel stack test section (T1) 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 (T2). 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

  20. 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)

  1. Nuclear criticality safety of fuel element storage for upgraded research reactor JRR-3

    International Nuclear Information System (INIS)

    Criticality aspects of storing 20 % enriched U Alsub(x) - Al fuel elements were evaluated for the upgraded research reactor JRR-3. Criticality calculations were carried out as a function of the number of fuel elements, lattice pitch, and water density in the moderator. The effects of neutron absorbers on neutron multiplication were also examined for storage arrays of the fuel element. Results show that the arrays in the storage racks proposed for the JRR-3 are subcritical enough. The fuel elements can be safely stored against any possible storage circumstances. The obtained data are presented in a form in which interpolation may be made to estimate the neutron multiplication factor of any element storage configulations of the fuel elements of the JRR-3. (author)

  2. 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)

  3. 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

  4. 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)

  5. Calibration of the Failed-Fuel-Element Detection Systems in the Aagesta Reactor

    International Nuclear Information System (INIS)

    Results from a calibration of the systems for detection of fuel element ruptures in the Aagesta reactor are presented. The calibration was carried out by means of foils of zirconium-uranium alloy which were placed in a special fuel assembly. The release of fission products from these foils is due mainly to recoil and can be accurately calculated. Before the foils were used in the reactor their corrosion behaviour in high temperature water was investigated. The results obtained with the precipitator systems for bulk detection and localization are in good agreement with the expected performance. The sensitivity of these systems was found to be high enough for detection and localization of small defects of pin-hole type (ν 10-8/s ). The general performance of the systems was satisfactory during the calibration tests, although a few adjustments are desirable. A bulk detecting system for monitoring of activities in the moderator, in which the γ-radiation from coolant samples is measured directly after an ion exchanger, showed lower sensitivity than expected from calculations. It seems that the sensitivity of the latter system has to be improved to admit the detection of small defects. In the ion exchanger system, and to some extent in the precipitator systems, the background from A41 in the coolant limits the sensitivity. The calibration technique utilized seems to be of great advantage when investigating the performance of failed-fuel-element detection systems

  6. Reactor vessel model flow tests for 145-fuel assembly core

    International Nuclear Information System (INIS)

    Hydraulic tests on a one-sixth-scale model of a two-loop pressurized water reactor with 145 fuel assemblies are described. Core inlet and outlet flow distributions and reactor vessel pressure drop were investigated. The core inlet flow distribution was developed to be independent of the flow conditions in the inlet annulus. A flow distribution system, consisting of several flow splitters in the inlet annulus and a spherical plate flow distributor in the lower head region, was developed to obtain a symmetric and stable core inlet flow distribution. A minimum core inlet flow factor of 0.99 was established in the core. Reactor vessel unrecoverable pressure drops were measured on the model to predict losses that will occur in the prototype

  7. Measurements of fuel burnup for the RA reactor spent fuel elements stored in the stainless steel containers (Draft version)

    International Nuclear Information System (INIS)

    According to the Radiological Characterisation Plan of the RA reactor, the accurate data on fuel burnup are very important for the radiation safety provisions during removal of spent fuel elements from the RA reactor as well as for verification of methods, geometry models and historically reviewed data concerning fuel irradiation. These data and methods will be used for neutron flux calculations in the RA reactor cores, reflector and biological shield, and finally for activity calculations of hard-to-detect radionuclides in the graphite reflector and concrete shields. Since the comparison of previous experimental data with the calculations showed discrepancy of 25% , fuel burnup of all fuel elements stored in the stainless steel containers was measured recently (from february to August 2006). This progress report summarizes the techniques and methods used for fuel burnup measurements of both type fuel elements (2% enriched metal uranium and 80% enriched uranium dioxide). It presents results for some maximum burned fuel elements and contains results of multichannel scanning of gamma ray emission from all stainless steel containers with spent fuel elements in storage pool

  8. 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)

  9. 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

  10. Isotope correlation studies relative to high enrichment test reactor fuels

    International Nuclear Information System (INIS)

    Several correlations of fission product isotopic ratios with atom percent fission and neutron flux, for highly enriched 235U fuel irradiated in two different water moderated thermal reactors, have been evaluated. In general, excellent correlations were indicated for samples irradiated in the same neutron spectrum; however, significant differences in the correlations were noted with the change in neutron spectrum. For highly enriched 235U fuel, the correlation of the isotopic ratio 143Nd/145+146Nd with atom percent fission has wider applicability than the other fission product isotopic ratio evaluated. The 137Cs/135Cs atom ratio shows promise for correlation with neutron flux. Correlations involving heavy element ratios are very sensitive to the neutron spectrum

  11. A Study of the Temperature Distribution in UO2 Reactor Fuel Elements

    International Nuclear Information System (INIS)

    Thermal conductivity is one of the most important properties of nuclear reactor fuels. Accurate knowledge of this property is vital because, among other things, it determines the maximum power that can be taken out of the fuel element per unit length of the material without exceeding the safety limits of the fuel elements. This report consists of a study of the thermal behaviour of uranium dioxide in the form of reactor fuel. The experimental part of the report describes measurements performed at the OECD Halden Reactor Project, Halden, Norway. The experiment was originally set up in order to measure the temperature at the center of a UO2 fuel element as a function of element power, in order to determine the safe operation limit of the fuel assembly. However, in analysing the data obtained, very interesting thermal conductivity values were obtained and comparison with existing correlations could be performed. This comparison shows that a certain agreement is obtained between the measured data at Halden and a theory published by J.L. Bates in 1961, which predicts an increase in the thermal conductivity above 1500 deg C. The data obtained below 1300 deg C are also in good agreement with measurements performed by Vogt, Grandell and Runfors in 1964. The report contains a mathematical description of the heat transfer mechanisms in cylindrical fuel elements. The model is coded in FORTRAN IV-code and referred to as FUELTEMP

  12. 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.)

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

    International Nuclear Information System (INIS)

    Thermal flux irradiation testings of small mixed oxide pellets UPuO2 fuel elements were performed in support of the fuel reference design for the Phenix fast reactor. The effects of different parameters (stoichiometry, pellet density, pellet clad gap). on the behaviour of the oxide (temperature distribution, microstructural changes, fission gas release) were investigated in various irradiation conditions. In particular, the effect of fuel density decrease and power rate increase on thermal performances were determined on short term irradiations of porous fuels. (authors)

  14. 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)

  15. Study of experiments with complex reactor core consisting of two types of fuel elements

    International Nuclear Information System (INIS)

    Criticality parameters and neutron flux distributions were measured at the heavy water RB reactor with the mixed reactor core. Cylindrical natural uranium fuel and tubular two-percent enriched metal fuel elements were used. Results were obtained under assumption of simple volume homogenization and procedure based on two-zone supercell estimation. Theoretical results obtained by homogenization are in good agreement with the experimental results, especially in case of criticality parameters of the system. This conclusion could not be generalized, because there is no significant difference in micro-distribution of neutron flux due to small difference in total thermal neutron absorption in each type of fuel elements, although the differences in multiplication and macroscopic cross sections are rather high. The obtained results are promising for studying the two-zone cell model in case of mixed core with significant mutual interference of different fuel elements, as well as for further improvement of this model

  16. 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)

  17. 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

  18. Radiological safety assessment during repackaging and transporting of the RA reactor spent fuel elements (Draft version)

    International Nuclear Information System (INIS)

    This report summarises the geometry models and calculation methods developed at the Vinca Institute of Nuclear Sciences for analysis of gamma ray and neutron source terms and equivalent dose rates on the outer surfaces of existing and transport containers with irradiated fuel elements of the RA reactor. The burnup data of most irradiated fuel elements with 2% 235U enriched metal uranium (LEU) are based on the fuel burnup measurements of fuel elements stored in the stainless steel containers in the water pools of the RA reactor spent fuel storage, performed by using the semiconductor detector with the CdZnTe crystal shielded with tungsten. The methodology for three-dimensional (3D) fuel burnup analysis of RA reactor cores founded on coupling Monte Carlo method for 3D calculation of node power distribution and transport method for depletion calculation in one-dimensional (1D) equivalent cell for each node independently was used for most irradiated fuel elements with 80% 235U uranium dioxide (HEU). The gamma rays, neutron and beta particle source terms analysis was founded: on the application of design-oriented SAS2H sequence (from the SCALE-4.4a code system) for 1D geometry models; and reference methodologies MOCUP (MCNP-4C/ORIGEN2.1) and KWO2 (KENOV. a/ORIGEN2.1) in 3D geometry models of the RA reactor unit cells. The MORSESGC code (from the SCALE-4.4a code system) was used for design oriented shielding analysis. For reference shielding calculation the MCNP-4C code and detailed geometry models of spent fuel elements containers were used. The study representing the experimental validation of methods and geometrical models for shielding analysis are also presented. Finally, the basic data obtained with presented methodologies for fuel burnup of most irradiated LEU and HEU spent fuel elements, for gamma ray and neutron source terms of these fuel elements; and for gamma ray and neutron equivalent dose rates on the outer surfaces of transport casks with the RA reactor

  19. 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

  20. Fuel-element failures in Hanford single-pass reactors 1944--1971

    International Nuclear Information System (INIS)

    The primary objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emissions since 1944 from the US Department of Energy's (DOE) Hanford Site near Richland, Washington. To estimate the doses, the staff of the Source Terms Task use operating information from historical documents to approximate the radioactive emissions. One source of radioactive emissions to the Columbia River came from leaks in the aluminum cladding of the uranium metal fuel elements in single-pass reactors. The purpose of this letter report is to provide photocopies of the documents that recorded these failures. The data from these documents will be used by the Source Terms Task to determine the contribution of single-pass reactor fuel-element failures to the radioactivity of the reactor effluent from 1944 through 1971. Each referenced fuel-element failure occurring in the Hanford single-pass reactors is addressed. The first recorded failure was in 1948, the last in 1970. No records of fuel-element failures were found in documents prior to 1948. Data on the approximately 2000 failures which occurred during the 28 years (1944--1971) of Hanford single-pass reactor operations are provided in this report

  1. 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)

  2. 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

  3. 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)

  4. French Fuel Elements for Natural Uranium-Graphite-CO2 Reactors: Experience Gained and Future Prospects

    International Nuclear Information System (INIS)

    The purpose of this paper is to summarize fuel element development for this reactor type in France in the light of the experience gained with these elements. Progress in the laboratory in work on basic materials (uranium alloys and magnesium alloys) and on the design and assembly of the fuel elements is very briefly recapitulated from the first G2 element studies (1954) to the current work on EDF3 and beyond. The following are then considered: (1) Experience gained in fabricating the G2, G3. EDF1 and EDF2 fuel elements, i.e. (a) the technical problems of mass production, (b) the production cost and the effects on it of the ever-increasing complexity of proposed elements and (c) planning problems, transition from prototype to mass production; (2) In-Pile operating experience with elements actually used in the G2 and G3 reactors, and the experimental irradiations carried out on EDF1 and EDF2. Finally, an attempt is made to determine the cost of the fuel cycle for this reactor type and future developments. (author)

  5. Irradiation device for power cycling testing of CANDU type fuel elements

    International Nuclear Information System (INIS)

    At INR Pitesti an irradiation device (capsule-C9) was designed and realized for testing the fuel element behaviour at reactor power variations occurring during normal operation of CANDU reactors in load following regime. This device allows the study of the phenomena at which the fuel elements in CANDU reactor are subject in conditions of: - normal restarting after shutdown and reactor de-poisoning; - variations of reactor power within 50-100% rated power; - return to rated power after operation at reduced power to prevent xenon poisoning; - restart within 30 minutes from the shutdown to prevent xenon poisoning; - adjusting reactivity during the return to rated power after reactor operation at reduced power; - displacement of fuel clusters in the channel by reactor loading. The power cycling can entail failure mechanisms specific to reactor operation in load following regimes, such as: - deformation of fuel element can by fuel-can interaction; - stress crevice corrosion; - corrosion assisted can fatigue; - can thinning in the vicinity of cracked pellets; - can cracking due to relocation of pellet fragments. Power cycling on the fuel element subjected to irradiation in capsule C-9 is performed by displacing the tested section in the experimental channel. Displacing the tested section under flux allows obtaining the required power values on the tested fuel element while the capsule instrumentation allows the monitoring of irradiation parameters, namely: - the linear power on the fuel element; - instant neutron flux at the force tube level; - coolant pressure within the tested section; - coolant activity; - chemical characteristics of the coolant. The main thermal-hydraulic characteristics of the capsule C-9 are: - working fluid, demineralized and degassed water; - coolant pressure, 120 bar; - coolant temperature, 150-160 deg. C; - maximum temperature on fuel element can, 325 deg. C; - thermosyphon flow rate at the tested section level, 0.15 kg/s; - disposable maximum

  6. 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.

  7. Thermal and hydraulic test plan of TRU fuel element for transmutation process

    International Nuclear Information System (INIS)

    JAERI is developing processes to partition long-lived transuranic elements (TRU) from high-level radioactive waste and transmutation processes to transform TRU into shorter-lived or stable nuclides under the OMEGA program. To promote developments of transmutation processes, thermal and hydraulic tests were planed to optimize a fuel element of an actinide burner fast reactor (ABR) cooled by helium gas. Along the test plan, a simulated fuel element in which simulated fuel particles were filled up in the porous annular space of 11.7mm in gap width and of 600mm in length was manufactured experimentally, and also a test apparatus which could circulate helium gas or nitrogen gas at a maximum flow rate of 400 m3/h under 1 MPa was designed and fabricated. Hydraulic performance of the test apparatus was confirmed through preliminary operations. This paper presents mainly a thermal and hydraulic test plan of the fuel element for developing ABR core design, outlines of the simulated fuel element and the test apparatus, and preliminary operation results. (author)

  8. 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.

  9. 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.

  10. 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.

  11. Release of radioactive fission products from BN-600 reactor untight fuel elements

    International Nuclear Information System (INIS)

    The experimental data on the release of radioactive fission products from BN-600 reactor untight fuel elements are given in the report. Various groups of radionuclides: inert gases Xe, Kr, volatile Cs, J, non-volatile Nb, and La are considered. The results of calculation-experimental study of transfer and distribution of radionuclides in the reactor primary circuit, gas system and sodium coolant are considered. It is shown that some complex radioactivity transfer processes can be described by simple mathematical models. (author)

  12. 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. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. 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, there 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

  13. 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

  14. 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

  15. Heat transfer and pressure drop of a reactor fuel element model with polyzonal spiral finning

    International Nuclear Information System (INIS)

    Heat transfer and pressure drop of a reactor fuel element model with polyzonal spiral finning have been investigated. The St-number distribution over length and perimeter of he finning are given. The mean and minimum Stk-number are plotted against the Re-number. The influence of the gap between two fuel elements upon heat transfer and pressure drop, in dependence on the Re-number, and the influence of the length of the fuel element on pressure drop across the gap are shown. The influence of the relative position of the splitters of two neighboring fuel elements on pressure drop and heat transfer is shown. The investigations were performed in the Re-number range 15,000 to 100,000 (author)

  16. Reactivity temperature coefficient evaluation of uranium zirconium hydride fuel element in power reactor

    International Nuclear Information System (INIS)

    Highlights: ► We develop an in-core fuel management code package for uranium zirconium hydride power reactor. ► The influence of changes on U–ZrHx fuel element is calculated and analyzed theoretically. ► Increased uranium contents in U–ZrHx reduce prompt negative temperature coefficient markedly. ► Additional poison erbium makes prompt negative temperature coefficient much more negative. ► The characteristics of inherent safety of U–ZrHx core can be retained in power reactors. -- Abstract: An in-core fuel management code package for uranium zirconium hydride power reactor, which is developed on the basis of the assembly lattice code TPFAP and the core calculating code BMFGD for LWR, is firstly introduced in this paper. The inherent safety of the U–ZrHx element which is mainly caused by the high prompt negative temperature coefficient is then evaluated, because the weight percentage of uranium, fuel rod radius and fuel temperature of U–ZrHx element will be different in power reactor from those in research reactor, and these changes may make obvious effect on the prompt negative temperature coefficient. The influence of weight percentage of uranium, fuel rod radius, fuel temperature, content of hydrogen and additional poison on prompt negative temperature coefficient for uranium zirconium hydride element are calculated respectively in this paper, and then the results are analyzed theoretically. The study shows that the absolute value of prompt negative temperature coefficient reduces observably along with the increasing of Uranium weight percentage from 10 wt% in research reactor to maximum 45 wt% in power reactor. Smaller radius, higher operating temperature and longer core life make little effect on the prompt negative temperature coefficient in the condition of high weight percentage of U. Additional poison erbium in fuel makes prompt negative temperature coefficient much more negative. Anyway, high prompt negative temperature coefficient can

  17. Design and fabrication of the instrumented fuel elements for the Annular Core Research Reactor (ACRR)

    International Nuclear Information System (INIS)

    This report describes the design and fabrication techniques for the instrumented fuel elements of the Annular Core Research Reactor (ACRR). The thermocouple assemblies were designed and fabricated at Sandia Laboratories while the instrumented elements were assembled at Los Alamos Scientific Laboratory. In order to satisfy the ACRR's Technical Specifications, the thermocouples are required to measure temperature in excess of 18000C under rapid heating conditions. Because of the potentially high failure rates for thermocouples in such environments, the instrumented fuel elements are designed so that the thermocouples can be replaced easily

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-02-01

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

  20. 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)

  1. ITER at Cadarache

    International Nuclear Information System (INIS)

    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.)

  2. Fuel element in rod form for nuclear reactors

    International Nuclear Information System (INIS)

    The fuel and/or breeding zone and the fission gas plenum are arranged so that the heat transfer medium sodium can be evaporated in the axial zone of the fuel and/or breeding area with spherical particles or hollow pellets and can be recondensed in the fission gas plenum. Due to gravity and the capillary effect, the liquid sodium flows back again (effect of a heat pipe). (DG)

  3. Review of behavior of mixed-oxide fuel elements in extended overpower transient tests in EBR-II

    International Nuclear Information System (INIS)

    From a series of five tests conducted in EBR-II, a substantial data base has been established on the performance of mixed-oxide fuel elements in a liquid-metal-cooled reactor under slow-ramp transient overpower conditions. Each test contained 19 preirradiated fuel elements with varying design and prior operating histories. Elements with aggressive design features, such as high fuel smear density and/or thin cladding, were included to accentuate transient effects. The ramp rates were either 0.1 or 10% ΔP/P/s and the overpowers ranged between ∼60 and 100% of the elements' prior power ratings. Six elements breached during the tests, all with aggressive design parameters. The other elements, including all those with moderate design features for the reference or advanced long-life drivers for PNC's prototype fast reactor Monju, maintained their cladding integrity during the tests. Posttest examination results indicated that fuel/cladding mechanical interaction (FCMI) was the most significant mechanism causing the cladding strain and breach. In contrast, pressure loading from the fission gas in the element plenum was less important, even in high-burnup elements. During an overpower transient, FCMI arises from fuel/cladding differential thermal expansion, transient fuel swelling, and, significantly, the gas pressure in the sealed central cavity of elements with substantial centerline fuel melting. Fuel performance data from these tests, including cladding breaching margin and transient cladding strain, are correlatable with fuel-element design and operating parameters. These correlations are being incorporated into fuel-element behavior codes. At the two tested ramp rates, fuel element behavior appears to be insensitive to transient ramp rate and there appears to be no particular vulnerability to slow ramp transients as previously perceived

  4. Review of behavior of mixed-oxide fuel element in extended overpower transient tests in EBR-II

    International Nuclear Information System (INIS)

    From a series of five tests conducted in EBR-II, a substantial data base has been established on the performance of mixed-oxide fuel elements in a liquid-metal-cooled reactor under slow-ramp transient overpower conditions. Each test contained 19 preirradiated fuel elements with various designs and prior operating histories. Elements with aggressive design features, such as high fuel smear density and/or thin cladding, were included to accentuate transient effects. The ramp rates were either 0.1 or 10% ΔP/P/s, and the overpowers ranged from approx. = 60 to 100% of the elements' prior power ratings. Six elements, all with aggressive design parameters, breached during the tests. The other elements, including all those with moderate design features for the reference or advanced long-life drivers for PNC's prototype fast reactor Monju, maintained their cladding integrity during the tests. Posttest examination indicated that fuel/cladding mechanical interaction (FCMK) was the most significant mechanism causing the cladding strain and breach. In contrast, pressure loading from the fission gas in the element plenum was less important, even in high-burnup elements. During an overpower transient, FCMI arises from fuel/cladding differential thermal expansion, transient fuel swelling, and, significantly, the gas pressure in the sealed central cavity of elements with substantial centerline fuel melting. Fuel performance data from these tests, including cladding breaching margin and transient cladding strain, are correlatable with fuel-element design and operating parameters. These correlations are being incorporated into fuel-element behavior codes. At the two tested ramp rates, fuel element behavior appears to be insensitive to transient ramp rate and there appears to be no particular vulnerability to slow-ramp transients, as previously perceived. (author)

  5. Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). 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. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. 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. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

  6. Research and test reactor fuel treatment at AREVA NC la Hague

    International Nuclear Information System (INIS)

    The La Hague plant has a long successful history of treatment power reactor spent fuel. To accommodate a new contract, the process needed to be adapted to fuel from Research Test Reactors (RTR) taking into account the dimensional characteristics, the chemical composition and the high uranium enrichment of RTR fuel. Several treatment options resulted have been studied and proposed. The chosen approach for the detailed design consists of dissolving the RTR fuel in a dissolution pit in one line of dissolver facility (T1B). The process was designed for small RTR fuel elements composed of aluminium and uranium alloy. 'Taxi' baskets containing the RTR fuel elements are transferred from the fuel storage pools, to the feed cell in T1B. The fuel elements are loaded into canisters and transferred to a rack in the general maintenance cell. The RTR fuel canisters are picked up one by one and the fuel element is dropped into a transfer tube leading to the dissolution pit. The RTR dissolution process is fundamentally different from the continuous power reactor fuel dissolution process and is done in batches. The RTR fuel elements are completely dissolved in nitric acid. It appeared that the dissolution rate is higher than expected allowing more flexibility in operation. At the end of a batch, the solution is drained from the dissolver and diluted with UOX solutions from line T1A. The resulting UOX/RTR mixture complies with the specifications for downstream processing operations, including high level waste vitrification, where aluminium is incorporated into glass in accordance with specified limits. The innovative nature of the process, demanded by the special characteristics of the RTR fuel, required a major qualification program. The main objectives of the qualification program were to validate the dissolver pit concept, to verify basic RTR process data for the T1B production environment and to acquire data showing control of the process. The first active batch dissolution was

  7. 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)

  8. 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

  9. 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)

  10. 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)

  11. 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.

  12. 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)

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

    International Nuclear Information System (INIS)

    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)

  14. Detection of fuel element failures at the gas cooled reactors

    International Nuclear Information System (INIS)

    This paper describes the system for measuring the concentration of released Kr and Xe in the cooling gas. The system developed in the Jozef Stefan Institute is efficient and suitable for application The method is based on electrostatic collection of daughter elements from radioactive decay of Xe and Kr

  15. Development of IRT-type fuel assembly with PIN-type fuel elements for LEU conversion of WWR-SM research reactor in Uzbekistan

    International Nuclear Information System (INIS)

    The design of new IRT-type fuel assembly (FA) with pin-type fuel elements has been developed. The IRT-type FA consists of the end details and a square shroud, in which 176 pin-type fuel elements are placed by means of two spacer grids. Fuel element is a square cross-section pin with four ribs curled around longitudinal axis. A large set of technological investigations of fuel element and FA fabrication processes has been completed, including: - development of drawing-design and technological documentations; - development of the design of spacer as a top and bottom grid; -manufacturing of needed tools and rigs; - development of fabrication process with fuel element simulators; - development of fabrication process for outer and inner shrouds; - development of the process of FA assembling. As a result two full-scaled dummy of FA for hydraulic tests have been manufactured and ability for manufacturing of fuel element and assembly have been confirmed under conditions of industrial production. To define the hydraulic characteristics of IRT-type FA with pin-type fuel elements the hydraulic tests of two full-scaled dummy of FA have been carried out. A large set of neutronics, thermal hydraulic and strength calculations for the WWR-SM reactor in Uzbekistan has been performed on substantiation of FA design. As a result: - the dispersion LEU fuel U-Mo in Al matrix has been chosen - the optimum U-235/FA loading has been defined; - the optimal parameters of pin-type fuel element and fuel assembly have been determined. These research of FA design are being performed in close cooperation with ANL within Russian and international RERTR programs. (author)

  16. 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)

  17. 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.

  18. Applied monitoring methods for the control of fuel elements and reactor internals in Argentine nuclear power plants poolside facilities

    International Nuclear Information System (INIS)

    Two pressurized heavy water reactors are being operated in Argentina: Atucha 1, a Siemens KWU designed and built pressure vessel 345 MWe reactor, started in 1974, and Embalse, a CANDU-600 MWe, started in 1983. Post-irradiation tests have been performed in accordance with the needs of fuel development and control programs for internal parts, such as the coolant channels (CC) where the fuel elements (FE) dwelled. The bays built in both NPP offer a good place to perform a large variety of pool-side inspection and testing activities. During the last 20 years more than 2000 CANDU-type FE, more than 1000 Atucha FE and above 100 Atucha CC, have been inspected in these bays. The usual inspection activities consist of: visual inspection, sipping tests, disassembling of FE, metrology, gamma scanning and eddy current testing. In this paper all of these techniques are briefly described and the obtained experience is shown. (author)

  19. Shipment of 255 DIDO fuel elements to the Savannah River Site to empty the storage and reactor pools at Risoe National Laboratory

    International Nuclear Information System (INIS)

    The DR-3 reactor, owned and operated by the Danish National Laboratory, was built in the late 1950's and initiated operation in January 1960. At that time the DR-1 and DR-2 reactors were already in operation. The main purpose if of Danish research reactor DR-3 was material and fuel testing. Until 1989 the reactor utilized HEU fuel elements. Conversion to the LEU fuel cycle was accomplished in 1990. DOE restarted the return program of for Foreign Research Reactor fuel elements to the United States in 1994. From that time, through 1998, three IUO4 casks (one cask in 1994) operated by Transnucleaire (now named Cogema Logistics, ACL) were used to transport Risoe's fuel to the Savannah River Site (SRS) near Aiken, SC in the USA. In 1999, Risoe elected to issue a request for proposal to transport DR-3 the DIDO fuel elements to SRS with a new licensed cask designed to replace the IUO4 cask. ACL was awarded the contract to transport the irradiated fuel from DR-3 to SRS for the remainder of the FRR Fuel Return program (2009). However, on September 28, 2000, the Board of Governors of Risoe National Laboratory decided to shut down the Danish research reactor of DR3. There had been of R2 technical problems (corrosion on the aluminum reactor tank) and, due to anticipated increasing operational expenses, the Board elected to close the reactor facility. Shortly thereafter, the Danish Government asked the National Laboratory to empty the reactor and its reactor and storage pools containing a total of 255 Dido irradiated fuel elements and ship them to Savannah Rive Site. At that time, ACL was in the process of licensing the new TN-MTR package in the USA. The early shut down of the DR-3 reactor and consequently the resultant new shipping schedule was not compatible with ACL's equipment and licensing schedule for the cask. (author)

  20. 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

  1. Tritium distribution between the fuel can and the oxide of fuel elements of light-water reactors

    International Nuclear Information System (INIS)

    The study on the measurement of tritium and other radionuclide contained in zircaloy fuel cans of the water cooled reactor fuel elements had two aims: the first was to estimate with accuracy the distribution of tritium in a fuel element (can + oxide). The measurement of tritium in the zircaloy fuel cans of the BORSSELE fuel elements associated with the measurement of tritium in the oxide allowed the establishment of a complete tritium balance on an industrial spent fuel element. This result has been compared to the values calculated by the code CEA/SEN and will allow to validate or adjust this calculation. The second aim delt with the characterization of the other radionuclides gaseous (Kr85) or not (Cs 134 and 137) contained in the solid zircaloy wastes (hulls) coming from the industrial reprocessing of ''water cooled'' fuel elements. These activity measurements in the hulls allowed to estimate the residual content of tritium, Kr 85 and other radionuclides which may be found in these solid wastes (high-level βγ radioactive wastes). Original experimental methods have been developed to reach these aims (dissolution in ammonium bifluoride medium and quantitative recovery of gases produced, radiochromatography, and liquid scintillation after double distillation). One tries to explain the presence of Kr 85 in the irradiated can

  2. Status of research reactor fuel test in the High Flux Reactor (Petten)

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; Casalta, S. [European Commission, JRC, NL-1755 ZG Petten (Netherlands); Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.; Dassei, G. [NRG Petten (Netherlands); Vacelet, H. [Cerca Framatome, F-26104 Romans (France); Languille, A. [CEA Cadarache, F-13108 Saint Paul Lez Durance (France)

    2001-07-01

    Even if the research reactors are using very well known MTR-fuel, a need exists for research in this field mainly for the reasons of industrial qualification of fuel assemblies (built with qualified fuel), improvement or modifications on a qualified fuel ( e.g. increase of density), and qualification of a new fuels such as UMo. For these types of tests, the High Flux Reactor located in Petten (the Netherlands) has a lot of specific advantages: 1) a large core with various interesting positions ranging from high to low fluence rate; 2) a high number of operating days (>280 days/year) that gives - with the high flux available - a possibility to reach quickly high burnup; 3) a downward coolant flow that simplifies the device engineering; 4) all possibilities of non-destructive and destructive examinations in the hot-cells (visual inspection, swelling, {gamma}-scanning, macro- and light microscopy, SEM and EPMA examinations, tomography). Two types of tests can be performed at the plant: either a full-scale test or a test of plates in dedicated devices. A presentation is made of the irradiation test on four UMo plates, begun in March 2000 in the device UMUS. A status report is provided of the full-scale test to be done in the near future, especially the UMo tests to begin the next year. In conclusion it appears that the HFR, that had already given an excellent contribution to silicide fuel qualification in the 1980s, will also give a significant contribution to the current UMo qualification programs. (author)

  3. Preliminary results of the BTF-105A test: an in-reactor instrument development and fuel behaviour test

    International Nuclear Information System (INIS)

    The BTF-105A test, performed in 1996 March, was the first of a pair of in-reactor tests planned to investigate fuel behaviour and fission-product release from CANDU-type fuel for the high-temperature conditions expected following large-break loss-of-coolant accidents (LOCA) with coincident loss-of-emergency-core-cooling (LOECC). The BTF-105A test assembly consisted of an instrumented fuel stringer containing a single, unirradiated, fuel element; while the second test, denoted BTF-105B, will use a similarly instrumented assembly containing a single, previously irradiated, fuel element. As the first of these two tests, the primary objectives of BTF-105A were to test instrumentation and procedures planned for use in BTF-105B, and to obtain data on the relationship between fuel-centreline and sheath temperatures under transient conditions with steam cooling. This paper describes the conduct and preliminary results of the BTF-105A test, and also indicates improvements to be made for the upcoming BTF-105B test. (author)

  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. 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.

  7. An expert system to analyze homogeneity in fuel element plates for research reactors

    International Nuclear Information System (INIS)

    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. (author)

  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. Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

    International Nuclear Information System (INIS)

    The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

  11. Process for distinguishing and sorting spherical fuel elements from high temperature nuclear reactors

    International Nuclear Information System (INIS)

    Each fuel element is heated to the outlet temperature of the high temperature reactor or to a temperature of 3000 or 6000C in a furnace. Helium flows through the furnace for three minutes and takes up the fission products released. The active carbon kept at the temperature of liquid nitrogen absorbs the fission products and the Xsub(e)-133 content is then measured. (GL)

  12. 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.)

  13. 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

  14. 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.

  15. Formation of Corrosive Deposits and Their Impact on Operational Safety of Fuel Elements in Candu Reactor

    International Nuclear Information System (INIS)

    Interaction between fuel element cladding and water coolant plays an important role in normal operation, can have a dominant role in accidental situations and can lead to failure of fuel rods and activity release. For the future, the tendency will be to increase the coolant temperature, extend fuel residence time in the reactor core (for higher burnup) and increase the heat flux. This can lead to increased probability of fuel failures due to waterside corrosion, corrosion products accumulation and deposition. In order to prevent cladding failures, the coolant chemistry must be monitored and controlled in order to reduce the amount of deposited crud and the oxygen potential. Corrosive deposits together with aqueous corrosion influence the performance of fuel elements by increase of temperature on cladding surface or changes in the coolant chemistry (increase of water pH), phenomena which lead to cladding failures. The process of corrosion products formation on zircaloy-4 fuel cladding surface and their consequences was evidenced by performing of experiments in: autoclaves circuits assembled in a by-pass loop of a CANDU-6 Reactor at NPP Cernavoda; irradiation loop of the TRIGA Reactor, and in laboratory static autoclaves. The determination of corrosion and the characterization of crud deposits on the zircaloy-4 surfaces were performed using gravimetric method, metallographic and electronic microscopy, and gamma spectrometry analysis and impedance electrochemical spectroscopy (EIS) determinations. The experimental results showed that the composition, thickness and evolution of corrosive deposits on fuel assembly surfaces depend very much on operational conditions, such as steady state operation, water chemistry conditions (pH and oxygen concentration) and different oxidation conditions of cladding surface. (author)

  16. Optimization of the distribution of bars with gadolinium oxide in reactor fuel elements PWR

    International Nuclear Information System (INIS)

    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)

  17. Technological Developments in Safe and Efficient Fabrication of Fast Reactor Fuel Elements

    International Nuclear Information System (INIS)

    The Fuel for 500 MWe Prototype Fast Breeder Reactor ( PFBR ) coming up at Kalpakkam, India consists of Mixed Oxide ( MOX ) fuel containing PuO2 and UO2.The fabrication MOX fuel elements for this reactor core is a challenging task as it involves issues related to radiological safety due to Plutonium handling, radiation exposure concerns and issues like efficient production and quality assurance. This paper deals with the technological developments carried out and their incorporation in the fabrication line to achieve higher throughput with low man-rem consumption. Vibratory bowl and linear feeders are being utilized for arranging the stack of small size i.e 5.5 mm diameter pellets and loading the stack inside the clad tube. Inactive bottom end plug welding has been successfully implemented using laser welding technique. The top end plug welding is carried out inside specially designed chamber in a glove box using TIG welding technique. The top end plug welding by laser welding technique has been demonstrated successfully and is going to be implemented shortly. Ultrasonic and laser decontamination techniques have been used to reduce transferable contamination on welded fuel pins. Issues related to radiological and criticality safety , safe handling of fuel elements and measures taken for exposure control are also discussed in this paper. (author)

  18. 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)

  19. 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

  20. Thermo-mechanical analysis of SEU 43 fuel element in CANDU reactor normal operational conditions

    International Nuclear Information System (INIS)

    The main direction of developing the CANDU type fuel is designing a new type of fuel cluster with the number of elements increased from 37 to 43. This work presents results of the research done in INR Pitesti on the new concept of SEU 43 fuel cluster designed for burnups as high as 25 Mw·day/kgU using slightly enriched uranium (up to 1.1% U235). By using ROFEM 1.0 code the behaviour of two types of SEU 43 fuel was analyzed in normal conditions of CANDU 6 reactor operation. The main performance parameters of SEU fuel were analyzed. These are: temperature distribution; the volume and pressure of fission gases; stresses in the fuel can; can deformations. Comparisons with the standard CANDU fuel are done. The results show the adequacy of the design solutions implemented for the SEU 43 fuel. The power-burnup history required by the ROFEM computations was obtained from the overpower envelope of the fuel cluster, with the radial power distribution on cluster taken into account. Evolution of the main performance parameters during irradiation is given

  1. 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.)

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

    International Nuclear Information System (INIS)

    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.

  3. 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)

  4. Sipping equipment for WWER-440 type reactors to make in-core studies on the tightness of fuel elements

    International Nuclear Information System (INIS)

    At the meeting of the Kraftwerk Union A.G. a survey on the sipping equipment originally designed to detect fuel element failures of BWRs was presented. Possibilities to apply the device for tightness inspection of fuel cladding in WWER-440 type reactors by means of 1 or 8 fuel assemby sipping bells were pointed out. (V.N.)

  5. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    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. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. 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. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program

  6. 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

  7. Fuel subassembly leak test chamber for a nuclear reactor

    International Nuclear Information System (INIS)

    A container with a valve at one end is inserted into a nuclear reactor coolant pool. Once in the pool, the valve is opened by a mechanical linkage. An individual fuel subassembly is lifted into the container by a gripper; the valve is then closed providing an isolated chamber for the subassembly. A vacuum is drawn on the chamber to encourage gaseous fission product leakage through any defects in the cladding of the fuel rods comprising the subassembly; this leakage may be detected by instrumentation, and the need for replacement of the assembly ascertained

  8. Fuel integrity project: analysis of light water reactor fuel rods test results

    International Nuclear Information System (INIS)

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  9. Channel blockage accident analysis for research reactors with MTR- type fuel elements

    International Nuclear Information System (INIS)

    It is the purpose of this study to investigate the feasibility of removing the residual decay heat from core of TR-2 ,which is a pool-type research reactor, after a channel blockage accident event and to identify the principal factors involved in cooling process. To analyze this accident scenery, THEAP-I computer code, which is a single phase transient 3-D structure/1-D flow thermal hydraulics code developed with the aim to contribute mainly to the safety analysis of the open pool research reactors, was modified and used. All of the analysis results figured out the fact that the core melting was inevitable in case of an uninterrupted operation (continuous operation) preceding a channel blockage accident of the TR-2 Reactor. Such a result will even be met if the blockage occurs only in a single fuel element. The results of analysis are expressed in terms of temperature field distribution as a function of time

  10. Behavior of mixed-oxide fuel elements during the TOPI-1E transient overpower test

    International Nuclear Information System (INIS)

    A slow-ramp, extended overpower transient test was conducted on a group of nineteen preirradiated mixed-oxide fuel elements in EBR-II. During the transient two of the test elements with high-density fuel and tempered martensitic cladding (PNC-FMS) breached at an overpower of ∼75%. Fuel elements with austenitic claddings (D9, PNC316, and PNC150), many with aggressive design features and high burnups, survived the overpower transient and incurred little or no cladding strain. Fuel elements with annual fuel or heterogeneous fuel columns also behaved well

  11. 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)

  12. 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,

  13. 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,.

  14. 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)

  15. Uranium dioxide caramel fuel. An alternative fuel cycle for research and test reactors

    International Nuclear Information System (INIS)

    The work performed in France on Caramel fuels for research reactors reflects the reality of a program based on non proliferation criteria, as they have already appeared several years ago. This work actually includes the following different aspects: - identification of the non proliferation criterion defining this action; - determination of the economical and technical goals to be reached; - realization of research and development studies finalized in a full scale demonstration; - transposition to an industrial and commercial level. The Caramel fuel goals have been defined by comparison with existing reactors: to keep the same performance level in the same safety and reliability conditions, without substantial increase in the fuel cycle cost. Taking into account the wide range of the reactors in operation, these goals will be reached, totally or partially, by assemblies with various geometries. The Caramel fuels utilize slightly enriched uranium, because of the high density of the uranium dioxide 10.25 g/cm.3 The reactivity control capacity of the core is consistent with the behaviour under irradiation so as to keep the operation cycle lengths with the same values as the present ones; the average burn-up being limited to about 30,000 MWd/t, the enrichment is maintained lower than 10%. A study of the Caramel behaviour under irradiation has been undertaken. It started with individual Caramel, and followed successfully with fuel assemblies, irradiated in the reactor Osiris within a significant environment: maximum specific power higher than 3000 W/cm3, and maximum burn up about 30 000 MWd/t. Safety experiments have led to creation of deliberate defects such as clad failure in order to test the irradiation behaviour to study its evolution. The results are positive, the kinetics being rather slow. The full change of a fuel cycle connected with nonproliferation goals appears to be a very wide program with political, technical and industrial implications. The development

  16. Qualification of the on-line power determination of fuel elements in irradiation devices in the BR2 reactor

    International Nuclear Information System (INIS)

    Fuel irradiation tests require an on-line monitoring of the fuel power. In the BR2 reactor, this is performed by continuously measuring the enthalpy change in the coolant of the irradiation device and complementing this information with data on power losses, heating of structure parts and spatial power profiles from mock-up test experiments and from calculations. Since a few years Monte Carlo codes (MCNP) are used, describing the BR2 core in great detail for every reactor cycle with its specific core load, yielding not only reliable relative values, but also calculated absolute local power values in agreement with data from PIE analyses. Several methods were conceived to combine the experimental and calculated data for the on-line calculation of the local linear power in the fuel elements; their internal consistency and the consistency with gamma spectroscopy data and data from radiochemical fission product analysis was checked. The data show that fuel irradiations in BR2 can be performed in a well-controlled way, with an accurate and reliable on-line follow-up of the fuel power. (author)

  17. 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

  18. Pulsed Nd-YAG laser welding of Prototype Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    End plug welding of Prototype Fast Breeder Reactor (PFBR) fuel elements involves welding of fully Austenitic Stainless Steel (ASS) of grade D9 clad tube with 316M end plug. Pulsed Gas Tungsten Arc Welding (GTAW) is being used for the production of PFBR fuel elements at Advanced Fuel Fabrication Facility (AFFF). GTAW is an established process for end plug welding and hence adopted by many countries. GTAW has got certain limitations like heat input, arc gap sensitivity and certain sporadic defects like tungsten inclusion. Experiments have been carried out at AFFF to use Laser Beam Welding (LBW) technique as LBW offers a number of advantages over the former process. This report mainly deals with the optimization of laser parameters for welding of PFBR fuel elements. To facilitate pulsed Nd-YAG laser spot welding, parameters like peak power, pulse duration, pulse energy, frequency and defocusing of laser beam on to the work piece have been optimized. On the basis of penetration requirement laser welding parameters have been optimized. (author)

  19. Reacteur Jules Horowitz (RJH): A new material testing reactor. Status report

    International Nuclear Information System (INIS)

    The 'REACTEUR JULES HOROWITZ' (RJH) is a new research reactor dedicated to material and nuclear fuel testing. This reactor, which will be erected in the CEA Cadarache nuclear research Center is now at a feasibility study stage. At the beginning of the next century, at a time when most of existing material testing reactors will have to be shutdown or will be at the end of their lifetime, le RJH will offer outstanding neutron flux levels (twice those of existing french reactors). This paper deals with the following topics - functional specifications of the project, - safety approach, - design and construction codes, - alternative designs under consideration at the feasibility stage. (author)

  20. 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

  1. 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)

  2. 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)

  3. Uranium-fuel thermal reactor benchmark testing of CENDL-3

    International Nuclear Information System (INIS)

    CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

  4. The fabrication and performance of Canadian silicide dispersion fuel for test reactors

    International Nuclear Information System (INIS)

    Fuel fabrication effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer material accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been Irradiated successfully to burnups in the range 65-80 atomic percent. Irradiations have also been conducted on mini-elements having 1.2 mm diameter holes In their mid-sections, some drilled before irradiation and others after irradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated silicide aspersion fuels. Small segments of fuel cores released 85Kr starting at about 520 deg. C and peaking at about 680 deg C. After a holding period of 1 hour at 720 deg. C a secondary 85Kr peak occurred during cooling (at about 330 deg. C) probably due to thermal contraction cracking. Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally, with encouraging results. After about 0.25 h at 530 deg. C the aluminum cladding developed very localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. (author)

  5. 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

  6. 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.

  7. Pressure drop calculation in a fuel element of a pool type reactor

    International Nuclear Information System (INIS)

    Even with the advances of hardware in computer sciences, sometimes it is necessary to simplify the simulation in order to optimize the results given the same calculation runtime. The object of this study is a thermodynamic analysis of the core of a pool type research reactor, focusing on natural circulation. Due to the high geometrical complexity of the core, the scale transfer process becomes an essential step to the thermodynamic study of the reactor. This process takes place by determining the effective equivalent properties obtained from a detailed simulation of the core and transferring them to a porous medium having a coarse mesh while preserving the overall characteristics. In this way, it will be able to obtain the quadratic resistance coefficient KQ by calculating the pressure drop inside the fuel element. To observe in detail the behavior of this flow, longitudinal and transversal cross sections will be made in different points, thereby observing the velocity and pressure distributions. The analysis will provide detailed data on the fluid flow between the fuel plates enabling the observation of possible critical points or undesired behavior. The whole analysis was made by using the commercial code ANSYS CFX ver. 12.1. This is study will provide data, as a first step to enable future simulations which will consider the entire reactor. (author)

  8. The future of spent TRIGA fuel elements from European TRIGA reactor stations

    International Nuclear Information System (INIS)

    The paper gives a summary of the information collected and presented to the General Atomics about TRIGA fuel elements available at European TRIGA stations under the initiative to solve the problem of the future of spent TRIGA fuel elements

  9. 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.

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

    International Nuclear Information System (INIS)

    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

  11. Loss of coolant accident analysis of supercritical water-cooled reactor fuel qualification test loop

    International Nuclear Information System (INIS)

    The supercritical water-cooled reactor fuel qualification test (SCWR-FQT) intends to test a small scale fuel assembly under supercritical water environment in a research reactor. The modified ATHLET code was applied to model the supercritical water-cooled experimental loop containing this fuel assembly and to perform the calculation analysis of the loss of coolant accident induced by the coolant pipe break. The results indicate that the design of existing safety system can practically ensure the effective cooling of the fuel rod experimental section in the accident scenario. The results also show that the modified ATHLET code has good suitability in simulation of supercritical water-cooled system. (authors)

  12. 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)

  13. Development of computer models for fuel element behaviour in water reactors

    International Nuclear Information System (INIS)

    Description of fuel behaviour during normal operation transients and accident conditions has always represented a most challenging and important problem. Reliable predictions constitute a basic demand for safety based calculations, for design purposes and for fuel performance. Therefore, computer codes based on deterministic and probabilistic models were developed. Possibility of comprehensive descriptions of the phenomena is precluded in view of the great number of individual processes, involving physical, chemical, thermohydraulical and mechanical parameters, to be considered in a wide range of situations. In case of fast thermal transients predictive capability is limited by the kinetics of evolution of the system and its eventual dynamic behaviour. Evidently, probabilistic approaches are also limited by the sparcity and limited breadth of the impirical data base. Code predictions have to be evaluated against power reactor data, results from simulation experiments and, if possible, include cross validation of different codes and validation of sub-models. Progress on this subject is reviewed in this report, which completes the co-ordinated research programme on 'Development of Computer Models for Fuel Element Behaviour in Water Reactors' (D-COM), initiated under the auspices of the IAEA in 1981

  14. 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.)

  15. 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)

  16. Radiotoxicity and Risk Reduction of TRU Elements from Spent Fuel by Transmutation in the Light Water Reactor

    International Nuclear Information System (INIS)

    A conventional PWR of type VVER-440 operating in a sustainable advanced fuel cycle mode with complete recycling of TRU elements in an Inert Matrix Combined Fuel Assembly (IMC-FA) in the same reactor was investigated. A preliminary assessment with the differences between various nuclear fuel cycles in terms of the risk analysis and its indicators has been conducted. The results indicate that the sustainable advanced fuel cycle option can, for the same amount of energy generation, significantly reduces both the amounts and radiotoxicity of the spent nuclear fuel in comparison with the conventional once-through UO2 or MOX fuel cycles

  17. 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

  18. 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.

  19. 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)

  20. 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.)

  1. Process and device for processing used fuel elements of water-cooled nuclear reactors

    International Nuclear Information System (INIS)

    The fuel elements are transported dry in a transport container to an opening into a hot cell. A fuel element manipulator takes the fuel elements from the transport container and moves them to a handover shaft into a fuel element storage pond filled with water. The manipulator lowers the fuel element into a fixed cooling container, where it is first cooled, before it is finally deposited in the storage basin. The cooling container has special water cooling and is immersed in the water of the storage pond. (DG)

  2. Characteristics and performance of the plant for refabrication of vibrocompacted fuel elements for the BOR-60 reactor

    International Nuclear Information System (INIS)

    The paper deals with the characterization of an automated pilot plant for the refabrication of fuel elements for the Soviet experimental BOR-60 fast breeder reactor. In addition, it describes the results, experience, and problems encountered in respect of the technological process of fuel element manufacture and quality assurance. The electrodynamic vibration procedure for densification of the uranium-plutonium mixed oxide is given special attention. (orig.)

  3. 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

  4. Full conversion of the core of a research, nuclear fuel testing and material testing reactor of TRIGA SSR 14 MW type

    International Nuclear Information System (INIS)

    All the TRIGA type reactors have power levels between 100 kW and 14 MW and use a nuclear fuel formed of a uranium - hydride zirconium alloy. The reactor cores made of this fuel are intrinsically secure against any accident of reactivity insertion. This type of fuel presents a low coefficient of fission product release and a remarkable stability to the temperature cycling up to high burnups. All the TRIGA type reactors except the 14 MW reactors, as the type of the TRIGA reactor of INR Pitesti is, make use of low enrichment uranium (19.9%), LEU. Due to the international policy of non-proliferation and nuclear safeguard requirements as well, the highly enriched uranium fuel fabrication ceased in 1978. Subsequently, only LEU type fuel continued to be used. The process of using LEU fuel instead of HEU fuel, coined as conversion, was supported by DOE of USA and the IAEA member states in the frame of the 'Reduced Enrichment in Research and Testing Reactors' (RERTR project) together with the technical cooperation projects (TC) between IAEA and the member states. INR Pitesti enjoyed the both forms of international support since 1984. The conversion process at the TRIGA reactor of INR Pitesti began in 1992 by feeding the reactor core with LEU fuel produced by General Atomics. The report of General Atomics from 1998, titled 'Final Results from TRIGA - LEU fuel Postirradiation Examination and Evaluation following Long-term Irradiation Testing in the OAK Ridge Reactor (ORR)' (UZR-22) presents the results of irradiation testing of the LEU f 13,22 fuel for a burnup of about 65% and an irradiation period of about 900 FPD (full power days). The report confirmed the reliability of the LEU fuel replaced in the conversion process. Similar results were obtained by irradiation of LEU fuel, produced by General Atomics, in the 14 MW TRIGA reactor of INR Pitesti and by post irradiation examination of some fuel element selected from the batch of fuel provided by the Argonne National

  5. 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).

  6. 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)

  7. ENEA TRIGA RC-1 Reactor spent fuel elements shipment in USA

    International Nuclear Information System (INIS)

    In the second half of July 1999, 140 spent fuel elements of ENEA TRIGA RC-1 reactor were shipped in the United State for the final storage. DOE, NAC, TRANSNUCLEAIRE, MIT and ENEA competencies for own questions are described. The activities planning, the organizing aspect study, the analysis and valuations of both nuclear safety and radioprotection have been suitable for the TRIGA RC-1 fuel element shipment. In fact, no operative anomaly appeared with respect to the approved procedures. Personnel engagement has been as expectations and the personnel absorbed gamma dose resulted to be negligible. Lastly, the NAC disposable narrow time (only one week at the end of July) has not produced heavy organizational problems but it has been a strong goad per all operative structures involved in the TRIGA RC-1 elements shipment. Further, the report describes the Italian Regulatory Body (ANPA) requests in order to obtain the maximum safety conditions. A brief description of the safety analysis prepared by TRIGA staff is also presented. (authors)

  8. 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

  9. 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)

  10. 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)

  11. The IRSN Institute of radiation protection and nuclear safety Cadarache Center

    International Nuclear Information System (INIS)

    The research programs of the IRSN in the Cadarache Center concern the nuclear safety (reactor safety, nuclear fuels behavior during accidents, fires in nuclear installations), the protection and the control of radioactive materials, the human and the environment protection. The programs are presented and discussed. This presentation includes also the regional impact of the Institute at Cadarache (economical impact, relations with the universities and international meetings). (A.L.B.)

  12. 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)

  13. Analytical study on effective coolant flow rate of flange type fuel element for very high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The objective of present study is to evaluate and to increase the effective coolant flow rate of VHTR (very high temperature gas-cooled reactor) core whose thermal output is 50 MW and core outlet gas temperature is 950 DEG-C, without changing the sizes of the core and fuel element. Four types of fuel element are discussed, they are two flange type pin-in-block fuel elements which have 18 fuel rods (P18) and 36 fuel rods (P36), a flange type multihole fuel element (MH8) and a dowel type pin-in-block fuel element (P15). The flow distribution analysis was carried out by flow network code which calculated coolant pressure, flow rate and temperature. As the result, the followings were obtained: Effective coolant flow rate (W-bareff) became lower in the order, MH8, P36, P18 and P15 fuels. W-bareff of the P15 fuel was the lowest, because crossflow loss coefficient of dowel type fuel element was lower than that of flange type. W-bareff of the P36 fuel was higher than that of the P18 fuel, because of larger total cross section of coolant flow channels. W-bareff of the MH8 fuel was higher than that of the P36 fuel, because of larger equivalent diameter of the channels. This analysis clearly revealed that the effective coolant flow rate increased by using the fuel element with large values of crossflow loss coefficient, cross section and equivalent diameter of coolant flow channels. (author)

  14. 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.)

  15. 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

  16. Worldwide survey of dry interim storage of fuel elements from power reactors

    International Nuclear Information System (INIS)

    Spent fuel elements with approx. 10,500 t of heavy metal are discharged annually from power reactors worldwide. The total amount existing is approximately 220,000 t. The capacity increase of decay storage pools by compact storage is about to meet limits in technical terms and with respect to licensing. External interim storage of spent fuel elements is becoming more and more important. There is a certain tendency worldwide to use interim stores on the sites of nuclear power plants in order to overcome the need for transports on public traffic routes. In Germany, this trend is enforced by the lack of central stores accessible where and when the need arises. Most interim stores use dry interim storage. In Central and Western Europe, casks are employed almost exclusively; in the United States, more canisters than casks are used. More and more systems are employed which allow both storage and transports (dual purpose systems). In the future, multipurpose systems could well become the systems of choice which, in addition, allow direct disposal of the inventory in the packages used for transport and interim storage. In Germany, important preconditions have been created in the development of the POLLUX trademark cask. No safety-related problems are expected to arise for either wet or dry interim storage. As a consequence, interim storage periods of up to 100 years seem to be possible without any technical problems especially for dry interim storage. Tentative plans along these lines are being developed in France and in the United States. (orig.)

  17. 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

  18. Calculated and measured behaviour of zircaloy fuel sheaths during in-reactor LOCA tests

    International Nuclear Information System (INIS)

    Six CANDU type fuel elements, containing UO2 fuel and sheathed in Zircaloy, have been subjected to transient conditions simulating a hypothetical large break loss of coolant accident (LOCA) in a CANDU reactor. The maximum transient sheath temperature was about 1273 K. Two single-element experiments were conducted at Chalk River Nuclear Laboratories in the X-2 loop of the NRX reactor. The other four elements experienced a single transient in the Power Burst Facility at Idaho National Engineering Laboratory. Comparisons are presented between data from these experiments and calculations by ELOCA, a computer code simulating fuel performance during transient conditions. The parameters evaluated included fuel sheath strain, internal element gas pressure, the mechanisms and timing of fuel element failure, fuel centreline temperature, sheath microstructure, and the thicknesses of zirconia and oxygen stabilized alpha-Zr layers on the sheaths. ELOCA calculations agreed well with the data. Some of the work described here was jointly funded by Atomic Energy of Canada Limited and Ontario Hydro, a Canadian utility, through the co-operative research and development program, CANDEV

  19. 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)

  20. 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)

  1. Non-Reactor testing stands for investigation of interaction of the fuel and constructive materials with the coolant

    International Nuclear Information System (INIS)

    In 1991 in the United Expedition of Scientific and Industrial Corp. Luch the non-reactor experiments were beginning. The appearances accompanied by water cooling reactor heavy accident were studied. There are 'Ruchej', 'LAVA', 'SLAVA' experimental facilities working out for such purposes. The 'Ruchej' facility is intended for conducting of the investigation of behavior of water cooling reactor core constructive elements in the high temperature gas-steam media. There were 27 testing start-up of 'Ruchej' facility and 70 fuel elements shells samples and 2 models fuel elements. 'LAVA' facility is intended to study the processes of the interaction of the melting composition of WWER-1000 reactor core with water. The 'SLAVA' facility is destined for study of corium jet characteristics and the processes of interaction of corium with WWER-1000 reactor constructive materials. The corium generation is realized in the electric melting furnaces (EPP-1, EPP-2) and both of them could be using for the 'LAVA' facility and the 'SLAVA' facility. The expenses, temperature, pressure of the water in the facility's cooling highway, pressure of gas within device, temperature of the corium or its imitator, geometrical parameters of stream' temperature of construct device's elements, electric parameters (voltage, current) has being registered

  2. 10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), located in the ATR Complex of the Idaho National Laboratory (INL), was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. The ATR is fueled with high-enriched uranium (HEU) matrix (UAlx) in an aluminum sandwich plate cladding. The National Nuclear Security Administration Global Threat Reduction Initiative (GTRI) strategic mission includes efforts to reduce and protect vulnerable nuclear and radiological material at civilian sites around the world. Converting research reactors from using HEU to low-enriched uranium (LEU) was originally started in 1978 as the Reduced Enrichment for Research and Test Reactors (RERTR) Program under the U.S. Department of Energy (DOE) Office of Science. Within this strategic mission, GTRI has three goals that provide a comprehensive approach to achieving this mission: The first goal, the driver for the modification that is the subject of this determination, is to convert research reactors from using HEU to LEU. Thus the mission of the ATR LEU Fuel Conversion Project is to convert the ATR and Advanced Test Reactor Critical facility (ATRC) (two of the six U.S. High-Performance Research Reactors (HPRR)) to LEU fuel by 2017. The major modification criteria evaluation of the project pre-conceptual design identified several issues that lead to the conclusion that the project is a major modification.

  3. 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.

  4. 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

  5. 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. The reduction of the 235U enrichment from above 90% to below 20% for such fuels would lessen the risk of diversion of the fuel for nonpeaceful uses. Fuel alloy powder prepared with low-enrichment uranium (235U) is dispersed in an aluminum matrix, and metallurgically roll bonded within a cladding of 6061 aluminum alloy. Miniplates with up to 55 vol% fuel alloy (up to 7.0 g total U/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 UAl /SUB x/ or U3O8, the current fuels for plate-type elements. A target loading of up to 7.0 g U/cm3 in the fuel zone was selected. To date the fabrication and irradiation results with the silicide fuels have been encouraging, and 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-sized plate fabrication and the irradiation of miniplates to a burnup of about 90% 235U depletion

  6. 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. The reduction of the 235U enrichment from above 90% to below 20% for such fuels would lessen the risk of diversion of the fuel for nonpeaceful uses. Fuel alloy powder prepared with low-enrichment uranium (235U) is dispersed in an aluminum matrix, and metallurgically roll bonded within a cladding of 6061 aluminum alloy. Miniplates with up to 55 vol% fuel alloy (up to 7.0 g total U/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 UAl/SUB x/ or U3O8, the current fuels for plate-type elements. A target loading of up to 7.0 g U/cm3 in the fuel zone was selected. To date the fabrication and irradiation results with the silicide fuels have been encouraging, and 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-sized plate fabrication and the irradiation of miniplates to a burnup of about 90% 235U depletion

  7. Determination of transverse ductility of fuel element canning from ring test

    International Nuclear Information System (INIS)

    of the tube. If, however, the strains are concentrated to such regions in the fuel as radial or axial cracks at the interface between the can and the fuel or at the interface between two fuel pellets the total strains might exceed the ductility. Such development of cracks due to localized straining is considered to be one of the most common normal causes of fuel element failure. Its consequences are serious inasmuch as they might lead to more or less complete disruption of the fuel can, in extreme cases resulting in the release of fuel and fission products into the coolant. In this paper the deformation of Zircaloy tubes under different testing procedures is analyzed. It is concluded that the ring expansion test is in the most cases suitable for the determination of the ductility of the cladding

  8. Analysis of the temperature field in a reactor fuel element of complex geometry

    International Nuclear Information System (INIS)

    An effective analytical method for determining the steady integral thermal conductivity and temperature distributions in cluster fuel elements has been developed. This method takes into account: distribution of heat generation, given by nonsymmetric function over the fuel rod cross section, q = q(r,φ); the thermal conductivity of the fuel and cladding material dependent on temperature, λ = λ(t), λk = λk (t); the fuel element cooling conditions defined by boundary conditions of the first, second or third kind. The second part of the paper presents the application of the developed method to a given fuel element. (author)

  9. Determination of reactor parameters by single fuel rod experiments

    International Nuclear Information System (INIS)

    Reactor parameters were measured by measuring neutron flux distribution in the vicinity of single fuel rod and it was shown that it can be applied for testing fuel elements. Four types of fuel elements were used, natural uranium rod, 2% enriched fuel element, and fuel clusters with 19 and 27 fuel rods. Neutron flux distribution was measured by irradiating Au and Dy foils. results were compared to results obtained by two-group and three-group neutron diffusion theory

  10. 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.

  11. Light water reactor fuel element suitable for thorium employment in a discrete seed and blanket configuration with the aim to attain conversion ratios above the range of one

    International Nuclear Information System (INIS)

    The thorium resources in the world are relatively large. According to the IAEA-NEA-publication ''Red Book'' they amount to 4.5 10E6 metric tons and are about 4 times greater than the resources of Uranium. The fuel element described in this paper could be used in light water reactor (LWR) preferably in pressurized water reactor (PWR). The seed (feed) rods contain uranium 235 as fissionable material and the blanket (breed) rods contain thorium and uranium. The thorium in the blanket rods is converted to fissionable U-233 by irradiation with thermal neutrons. The U-233 produced is a valuable fissionable material and is characterized by high revalues, where t is defined as the number of fission neutrons per absorption in fissile materials. By optimized configuration and loading of the seed- and blanket rods the thorium is converted to U-233 and the U-238 is converted to fissionable Plutonium isotopes. Consequently more fissionable material is generated than is used. The fuel cycle is also flexible. Thus U-235, Pu-239 or weapons-grade Plutonium can be used.Based on knowledge obtained in the development of fuel elements for material test reactors (MTR), high temperature reactors (HTR) and light water reactors (LWR), a new design of fuel element suitable for thorium employment in PWR is described.

  12. 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.

  13. Determination of Selected U-ZrH1.6 Fuel Element Depletion at PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Determination of nuclear fuel depletion in reactor after a long time of operation is very important. This allows the remaining fuel material in the reactor core to be fully utilized before transferred into storage area. Indeed, in order to properly handle nuclear spent fuel, we need to determine the level of burn-up and radiation generated prior to any activity. In this study, the fuel depletion of selected TRIGA fuel elements was determined at the Reaktor TRIGA PUSPATI (RTP), Malaysian Nuclear Agency. Parameter used was based on operational data of particular fuel elements until December 2012 where some of the fuel elements have been used since June 1982. Between these periods, the reactor core has been reloaded and/or reshuffled 14th times. Power distribution and burn-up level was calculated using TRIGLAV computer code. Further calculation was also made using ORIGEN2 computer code particularly in determining the fission product concentrations and radioactivity levels. Results of calculation are analysed and discussed in this paper. (author)

  14. 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)

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

    International Nuclear Information System (INIS)

    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

  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. Severe-fuel-damage experiments in the Canadian in-reactor Blowdown Test Facility

    International Nuclear Information System (INIS)

    The Blowdown Test Facility consists of an instrumented in-reactor irradiation site plus an out-reactor piping system. These are used to irradiate CANDU fuel under conditions representative of a loss-of-coolant accident (LOCA) or LOCA with Loss-Of-Emergency-Core-Cooling (LOECC) in order to study fuel performance, fission-product release from the fuel and the transport of fission products through the piping system. An overview of the facility and the experimental program is given in this paper. (author)

  18. 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

  19. 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

  20. 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

  1. 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

  2. Mesh generation for fast reactor fuel assembly in velocity and temperature field calculations using finite element method

    International Nuclear Information System (INIS)

    A set of computer codes for triangular six-node finite element mesh generation in fast reactor fuel assemblies (or their parts) is described as are directions for input data preparation and operator's actions. The generated meshes are utilized for velocity and temperature calculations. (author)

  3. In reactor performance of defected zircaloy-clad U3Si fuel elements in pressurized and boiling water coolants

    International Nuclear Information System (INIS)

    The results of two in-reactor defect tests of Zircaloy-clad U3Si are reported. In the first test, a previously irradiated element (∼5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at ∼270oC. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U3Si at 300oC is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, K.W.

    1980-05-02

    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. (ACR)

  5. 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

  6. 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.

  7. 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

  8. Process for manipulating and/or storing nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    In order to comply with the regulations for the subcriticality of the geometric arrangement of a transport and storage device, additional bodies made of neutron-absorbing material are introduced into the intermediate spaces between the fuel rods. The fixings of the additional bodies are situated at a skeleton end of the fuel element, which simplifies the manipulation, transport and storage of the fuel element. (DG)

  9. An overview of the fuels and materials testing programme at the OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    The fuels and materials testing programme of the OECD Halden Reactor Project is aimed at investigations of fuel and cladding properties at high burnup, water chemistry effects and in-core materials ageing problems. For the execution of this programme, different types of irradiation rigs and experimental facilities providing typical power reactors conditions are available. Data are obtained from in-core sensors developed at the Halden Project; these are shortly described. An overview of the current test programme and the scope of the following years are briefly presented. (author). 5 refs, 3 figs

  10. Sipping equipment for leak testing of fuel assemblies in VVER-440 reactors

    International Nuclear Information System (INIS)

    Sipping equipment for the Soviet-type VVER-440 pressurized water reactors was developed on the basis of the proven in-core sipping technique used for boiling water reactor fuel assemblies. The main components of the system are the sipping hood with seven test positions, the control panel for system operation and sample collection, and the manifold connection line. During testing the upper ends of the hexagonal fuel assemblies are lifted into the air-filled sipping hood to interrupt the coolant flow by means of pneumatically actuated grippers. The first equipment of this kind has been in use in the nuclear plant Jaslovske-Bohunice, Czechoslovakia, since 1986. (orig.)

  11. A Fast Test for Excessive U235 Concentrations in Enriched Fuel Elements

    International Nuclear Information System (INIS)

    This paper describes a new non-destructive method of detecting areas of enriched fuel elements containing excessive U235 , and an instrument designed to inspect uranium-aluminium alloy fuel for the NRU reactor using the new method. The new method is as precise as conventional methods but is m u c h faster. The improvement in test speed depends on the application, but typically m a y be a factor of four. Alternatively, the method m a y be used to improve the precision of the test without increasing the testing time. The method may be used when the U235 concentration in selected areas of a fuel element is estimated by measuring the intensity of 184-keV γ-rays (from the natural decay of U235 ) and when an area is considered acceptable if the measured intensity from it is less than a limiting intensity equivalent to the maximum allowable concentration. It the measured intensity is much less or much greater than the limiting intensity, much less time is required to make this decision than when they are nearly the same. The new method takes advantage of this. It thus inspects each area only as long as is necessary to make a decision with the required confidence. The average time taken in practical cases is less than that taken by conventional methods which inspect each area for a pre-selected time chosen to give the required confidence in the most difficult cases - when the measured and limiting intensities are nearly the same. The new method is based on sequential sampling theory. It may also be used to detect areas with undesirably low U235 concentration. (author)

  12. 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.)

  13. A kinetic model for fission-product release and fuel oxidation behaviour for zircaloy-clad fuel elements under reactor accident conditions

    International Nuclear Information System (INIS)

    During a severe reactor accident fission products will be released from the degraded fuel in the reactor core. In addition, hydrogen will be generated at high-temperature by the steam oxidation of the core materials. This oxidation process will also influence the rate of fission-product release. Separate-effects tests performed out-of-pile at the Chalk River Laboratories (CRL) have provided a better understanding of the processes of fission product release during severe accident conditions. The annealing experiments were conducted in steam at temperatures ranging from 1200 to 1700 deg C with irradiated fuel specimens of uranium dioxide in the form of bare fuel fragments and a short-length Zircaloy-clad fuel element. The fission product release was monitored by online gamma ray spectrometry. The oxygen partial pressure was also measured with solid-state oxygen sensors, providing a calculation of the rate of oxygen consumption and hydrogen production in the fuel specimens. Based on the CRL tests, an analytical model has been developed to describe the kinetic release behaviour of the volatile fission-product species (cesium) during high-temperature accident conditions. The physically-based model accounts for the kinetics of fuel oxidation as a rate-determining reaction at the fuel/steam interface. A more general framework is therefore provided to detail the influence of the atmosphere (i.e. oxygen potential) on the behaviour of the fission product release. Solid state diffusion in the fuel matrix is shown to be the rate-controlling mechanism in the early stages of release. The enhanced diffusivity of fission products in the hyperstoichiometric fuel is modelled with the assumption that diffusion takes place on vacant cation lattice sites. When the fuel reaches a state of oxidation of x ∼ 0.07 for the UO2+x phase, a more rapid release process occurs in accordance with first-order rate kinetics. The retarding influence of the hydrogen production on the fuel oxidation

  14. Study of the thermal drop at the uranium-can interface for fuel elements in gas-graphite reactors

    International Nuclear Information System (INIS)

    The report reviews the tests now under way at the CEA, for determining the thermal contact resistance at the uranium-can interface for fuel elements used in gas-graphite type reactors. These are laboratory tests carried out with equipment based on the principle of a heat flow across a stack of test pieces having planar contact surfaces. The following points emerge from this work: - for a metallic uranium element canned in magnesium, of the type G-2 or EDF-2, a value of 0.2 deg C/W/cm2 seems reasonable for can temperatures of 400 deg C and above. - this value is independent of the micro-geometric state of the uranium surface in a range of roughness which easily includes those observed on tubes and rods produced industrially. - for the internal cans of elements cooled internally and externally, the value of the contact resistance for temperatures of under 400 deg C as a function of the stresses in the can has not yet been measured exactly. (authors)

  15. LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

    International Nuclear Information System (INIS)

    The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

  16. In-Reactor Densification of Dual Cooled Annular Fuel Pellet during Irradiation Test at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, Young Woo; Kim, Dong Joo; Kwon, Hyoung Mun; Kim, Keon Sik; Kim, Jong Hun; Oh, Jang Soo; Yang, Jae Ho; Koo, Yang Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    These advantages result in a considerably low pellet centerline temperature. Because of this considerably low pellet temperature, in-reactor behavior of an annular pellet, such as densification and swelling may be significantly different from that of the conventional PWR solid pellet. Since the pellet temperature of an annular fuel rod is lower than that of a PWR solid fuel rod by several hundred degrees, the in-reactor densification and swelling of a dual cooled annular fuel pellet might be considered as athermal phenomena due to a low pellet temperature. In order to investigate the in-reactor behavior of the annular UO{sub 2} pellet, HANARO irradiation test was planned and conducted for annular pellets with 5 different types. Post irradiation test is being carried out in the KAERI's PIE facility. In this study, we are going to report the preliminary results of PIE test on the inreactor densification behavior of a dual cooled annular fuel pellet. Irradiation test of dual cooled annular UO{sub 2} pellet was conducted at the OR-4 hole in HANARO by using a non-instrumented test rig. The preliminary results of PIE test on the in-reactor densification behavior showed that the irradiated pellets densified much more than expected values based on MATPRO relations of inreactor densification at low temperature in the annular pellet with low initial sintered density. It might be attributed to the higher fission rate during HANARO irradiation.

  17. 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

  18. SPACE-R nuclear power system SC-320 thermionic fuel element performance tests

    International Nuclear Information System (INIS)

    In 1993 and 1994, the Russian Scientific Research Institute NII NPO ''LUCH'' and Space Power, Inc., (SPI), of San Jose, California, developed a prototype of the single-cell thermionic fuel element (TFE) for the SPACE-R space nuclear power system (NPS). The SPACE-R system was designed as a part of the US Department of Energy's (DOE) Space Reactor Development Program to develop a long life, space reactor system capable of supplying up to 40 kW(e) output power. The jointly developed SC-320 TFE is a prototype of the next generation thermionic converter for nuclear applications in space. This paper presents the results of the initial demonstration tests and subsequent parametric evaluations conducted on the SC-320 TFE as compared to the calculated performance characteristics. The demonstration tests were conducted jointly by Russian and American specialists at the Thermionic Evaluation Facility (TEF) at the New Mexico Engineering Research Institute (NMERI) of the University of New Mexico in Albuquerque

  19. Transient thermal analysis of a 1:2 scale cask for research reactors nuclear spent fuel elements considering thermal contacts and irradiation

    International Nuclear Information System (INIS)

    This work shows the approach used to the numerical simulation of the thermal test of a 1:2 scale model of a dual purpose cask (transportation and/or storage) for spent fuel elements from nuclear research reactors. Conservatively, the cask impact limiters are not modeled. This test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. Also, it is part of an IAEA sponsored project which includes Latin American countries with research reactors. This cask model has a stainless steel double wall cylinder (which contains the biological lead shielding) with flat heads and internal structures to accommodate the fuel elements. The cask project is described briefly as well as the developed finite element model and the main adopted hypothesis to consider the non-linearities as thermal contacts, properties varying with the temperature, phase change (thermal shielding lead) using the enthalpy method, and radiation among the internal parts. The analysis will cover the 30 min heating condition at 800 deg C and about 2 hours of the cooling phase. As the main purpose of the paper is to present the proposed approach for the thermal test numerical simulation, only some preliminary numerical results are shown without any comparison to the experimental ones. (author)

  20. 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)

  1. 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.)

  2. 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

  3. 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

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

    International Nuclear Information System (INIS)

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

  5. 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)

  6. Corrosion tests of candidate fuel cladding and reactor internal structural materials

    International Nuclear Information System (INIS)

    Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor (SCWR) in static and flowing supercritical water (SCW) autoclave at the temperatures of 550, 600 and 650°C, pressure of about 25MPa, deaerated or saturated dissolved hydrogen (STP). Samples are nickel base alloy type Hastelloy C276, austenitic stainless steels type 304NG and AL-6XN, ferritic/martensitic (F/M) steel type P92, and oxide dispersion strengthened steel MA 956. This paper focuses on the formation and breakdown of corrosion oxide scales, and proposes the future trend for the development of SCWR fuel cladding materials. (author)

  7. Corrosion tests of candidate fuel cladding and reactor internal structural materials

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, L.; Zhu, F.; Bao, Y. [Shanghai Jiao Tong Univ., School of Nuclear Science and Engineering, Shanghai (China); Tang, R. [Nuclear Power Inst. of China, National Key Lab. for Nuclear Fuel and Materials, Chengdu, Sichuan (China)

    2010-07-01

    Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor (SCWR) in static and flowing supercritical water (SCW) autoclave at the temperatures of 550, 600 and 650°C, pressure of about 25MPa, deaerated or saturated dissolved hydrogen (STP). Samples are nickel base alloy type Hastelloy C276, austenitic stainless steels type 304NG and AL-6XN, ferritic/martensitic (F/M) steel type P92, and oxide dispersion strengthened steel MA 956. This paper focuses on the formation and breakdown of corrosion oxide scales, and proposes the future trend for the development of SCWR fuel cladding materials. (author)

  8. 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

  9. 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.

  10. 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

  11. 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)

  12. Aluminum hydroxide coating thickness measurements and brushing tests on K West Basin fuel elements

    International Nuclear Information System (INIS)

    Aluminum hydroxide coating thicknesses were measured on fuel elements stored in aluminum canisters in K West Basin using specially developed eddy current probes . The results were used to estimate coating inventories for MCO fuel,loading. Brushing tests successfully demonstrated the ability to remove the coating if deemed necessary prior to MCO loading

  13. Development of sodium facilities for NSRR fast breeder reactor fuel tests. 2. Sodium capsule

    International Nuclear Information System (INIS)

    In order to commercialize fast reactors, which are expected to be long-term transmutes of plutonium and long half life radioactive wastes (such as americium) from light water reactors, safety research under accident conditions and establishment of the safety guidelines are essential. Sodium facilities, such as, (1) Purification/charging loop and test loop, and (2) Proto-type Sodium capsule, were developed and fabricated in order to pulse irradiate fast breeder reactor fuels in the Nuclear Safety Research Reactor (NSRR) of JAERI for investigation on fuel behavior under transient over-power conditions. This report presents the purpose, outlines, specifications, capabilities and operation results of the proto-type sodium capsule. Two kinds of capsule, i.e., the stagnant sodium capsule and the sodium loop, were designed to pulse irradiate Fast Reactor (FR) fuels in the NSRR under sodium cooling conditions with and without flow, respectively. Because the capsules have to safely contain chemically active sodium at high temperature and stand the pressure pulses by the sodium hummer which might be generated at fuel failure, the development of the capsule is essential for realizing the research. Thus, proto-type sodium loop, which consisted of doubly sealed container, sodium pump and flow meter, was developed. In addition, two type of flange structure for the stagnant capsule and loop was leak tested at high pressures, in order to confirm its sealing capability at room and high temperature conditions. (author)

  14. 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)

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

    Science.gov (United States)

    Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

    1999-03-23

    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. 7 figs.

  16. MTR (Materials Testing Reactors) cores fuel management. Application of a low enrichment reactor for the equilibrium and transitory core calculation

    International Nuclear Information System (INIS)

    This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun number of movements by rechange. The methodology proposed allows to determine the best options through conceptual relations, prior to a detailed calculation with the core code, reducing the test number with these codes and minimizing in this way CPU cost. The way to better systematized search of transient cores from the first one to the equilibrium one is presented. (Author)

  17. 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.

  18. 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.

  19. 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

  20. 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)

  1. Testing of irradiated spherical fuel elements at HTR MODUL relevant accident conditions

    International Nuclear Information System (INIS)

    It is reported that the German 200 MWth MODUL HTR uses spherical fuel elements having 10% enriched UO2 TRISO coated particles. Since 1984 the behavior of such elements of modern design under accident conditions has been studied at the Research Centre Juelich, FRG. By help of the Cold Finger Apparatus even the smallest release of fission products during testing up to 1800 deg. C can be analysed. Post heating examinations allowed important correlations between the distribution within the fuel element and the measured sphere release. The results of heating tests are described. Further work was carried out to simulate water and air ingress in a HTR. AN apparatus was built and is now commissioned. Tests with special samples and fuel spheres, and also with USA fuel are planned, to examine the influence of humidity on the fission product release. 14 refs, 13 figs, 7 tabs

  2. Application of gaseous absorbers for testing the fuel elements in nonstationary regimes

    International Nuclear Information System (INIS)

    A possibility of operative local control of power release in loop channels of a research reactor MR by means of a gaseous neutron absorber with 3He being of use as absorber is studied. The gas control system intended for the study of the WWER-1000 reactor fuel element serviceability in transients is described. The calculation-experimental investigation results of the dependence of power change of an experimental heat-generating assembly and reactivity on 3He pressure in the channel are presented. A conclusion on satisfactory efficiency of the described gas control system is made on the base of the analysis of the results obtained

  3. Application of gaseous absorbers for testing the fuel elements in nonstationary regimes. [/sup 3/He

    Energy Technology Data Exchange (ETDEWEB)

    Andreev, V.I.; Egorenkov, P.M.; Kolyadin, V.I.; Kosilov, A.N.; Kuznetsov, Eh.M.; Potapenko, P.T.; Sivokon, V.P.; Fokin, A.B.; Yakovlev, V.V.

    1981-11-01

    A possibility of operative local control of power release in loop channels of a research reactor MR by means of a gaseous neutron absorber with /sup 3/He being of use as absorber is studied. The gas control system intended for the study of the WWER-1000 reactor fuel element serviceability in transients is described. The calculation-experimental investigation results of the dependence of power change of an experimental heat-generating assembly and reactivity on /sup 3/He pressure in the channel are presented. A conclusion on satisfactory efficiency of the described gas control system is made on the base of the analysis of the results obtained.

  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. The management of spent fuel from materials testing reactors in the United Kingdom

    International Nuclear Information System (INIS)

    Four MTR Reactors, DIDO, PLUTO, DMTR and HERALD have been constructed and operated in the UK. Of these DMTR and HERALD have now been shut down. The spent fuel elements from DIDO and PLUTO are treated by chemical reprocessing at Dounreay to recover the unused U235 and concentrate the fission product activity. All of the fuel from the DMTR reactor has already been reprocessed at Dounreay, whilst the final core from HERALD will be reprocessed soon. After reprocessing the first cycle raffinate, which contains the bulk of the fission products from the spent fuel, is currently stored in high integrity, stainless steel storage tanks which are housed below ground in heavily shielded vaults. A new plant has been constructed at Dounreay which will first neutralize with caustic soda and then immobilize this liquor in cement. The resulting products will then be kept in an engineered store until an intermediate level waste repository is available. The uranyl nitrate product from reprocessing is evaporated and then blended up by the addition of U235. The uranium is then processed at Dounreay to make more MTR fuel, which is recycled to the remaining reactors. In the longer term uncertainty exists over the continued availability and economic viability of reprocessing, particularly if alternative, cheaper sources of fuel are used. The alternative of long term storage of the spent fuel, followed by direct disposal, is therefore being examined as a contingency. A review of the options has concluded that the preferred route would involve the interim storage of the elements in special cans in a cross-flow, natural convection, air-cooled vault. The use of free standing, concrete storage casks to store the fuel would be significantly more expensive and is not favoured. (author). 2 refs, 7 figs

  6. 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

  7. 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

  8. Process and device for testing vertical fuel rods of water-cooled nuclear reactors, which are collected into a fuel rod bundle

    International Nuclear Information System (INIS)

    To avoid high point loads on the frame and storage pond, a holding device for the fuel element is fitted in two unoccupied frame positions of a frame. A third frame position for accommodating a fuel element to be tested is kept free between the two unoccupied frame positions. After interlocking the fuel element magnetically with the holding device, the fuel element is lifted through the latter in the vertical direction, so that a sensor can drive between the fuel ords. The individual frame position is therefore subjected to a smaller load, as the whole device and the fuel element have a lower weight than two fuel elements. (orig./HP)

  9. Effect of high-density fuel loading on criticality of low enriched uranium fueled material test research reactors

    International Nuclear Information System (INIS)

    The effect of high-density fuel loading on the criticality of low enriched uranium fueled material test reactors was studied using the standard reactor physics simulation codes WIMS-D/4 and CITATION. Three strategies were considered to increase the fuel loading per plate: (1) by substituting the high-density fuel in place of low-density fuel keeping meat thickness and water channel width constant, (2) by substituting the high-density fuel in place of low-density fuel keeping fuel meat thickness fixed and optimizing the water channel width between the fuel plates and (3) by increasing the fuel meat thickness of fixed density fuel and optimizing the water channel width between the fuel plates. The fuel requirements for critical and first high power cores were determined in each case for higher fuel loadings per plate. It has been found that in the first case, core volume reduces with increasing fuel loadings per plate but requirement of fuel also increases. In the second and third case, core volume as well as fuel requirement decreases with increasing fuel loadings per plate. However in the second case, core volume reduces more rapidly than in case 3 with increasing fuel loadings per plate. Employing standard computer code PARET, steady state thermal hydraulic analysis of all these cores was performed. The thermal hydraulic analysis reveals that cores with higher densities and fixed water channel width are better from thermal hydraulic point of view and have fuel and clad temperatures within the acceptable limits. But the core with higher densities and optimum water channel width is a better choice in terms of core compaction, less 235U loading and higher neutron fluxes. Finally, the core was compacted in three steps to exploit the benefits of both types of cores. The strategy resulted in 36% reduction in the core volume, 50% increase in thermal neutron flux for irradiation and isotope production and a slight reduction in 235U loading. All this was achieved with

  10. 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).

  11. 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)

  12. 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.

  13. Qualification program for JHR fuel elements: Irradiation of the first JHR test assembly in the BR2-Evita loop

    International Nuclear Information System (INIS)

    An experimental program has been designed by CEA to qualify the behaviour of the JHR fuel under conditions representative of the reactor operating ones. This program uses the SCK.CEN facilities, irradiating JHR lead test elements in the BR2 reactor, inside its central channel which has been particularly arranged for this objective (Evita loop). As a first step in the program, a two cycle irradiation (4 weeks by cycle) started mid-July 2009 and ended mid-November (EVITA-1). After a cooling phase, this first JHR lead test element will be submitted to post-irradiation examination. The second JHR test element began its irradiation in the first quarter of 2010; its unloading is planned before the end of 2010, after 5 cycles in the BR2 reactor. The results of these two experiments are expected as input information for the Safety Authority Report. This paper presents the qualification program with the objectives assigned to each phase (irradiation, examination). A first interpretation of the irradiation data for the first element is presented, so as the information available on the progress of the following phases of the programme. (author)

  14. 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)

  15. Effects of alloys elements, impurities and microstructural factors in austenitic stainless steel to utilize in fuel rod of nuclear reactors

    International Nuclear Information System (INIS)

    Austenitic Stainless Steel is used as cladding material of pressurized water reactor fuel rods because of its good performance. The addition of alloy elements and the control of impurities make this to happen. Fission products do not contribute to corrosion. Dimensional changes are not critical up to 1,0 x 1022n/cm2 (E>0,1 MeV) of neutronic doses. The hydrogen does not cause embrittlement in the reactor operation temperatures, and helium contributes to embrittlement if the material is warmed upon 6500C. (author)

  16. Structural design of a shipping container of fuel elements, non-irradiated, for research reactors

    International Nuclear Information System (INIS)

    This work is part of a project whose ultimate goal the creation and subsequent discharge of a transport container fuel assemblies for use by the Chilean Commission for Energy Nuclear. In principle it is covered in the design stage, considering the materials and methods used, to further develop a stage of checking voltages in the container to be manufactured. To achieve the first phase of the study is necessary to understand and warn the importance, geometry and content of the fuel elements to be transported, for which there are standards that provide fundamental material for proper classification of both content and container design. Once approved the design of the structure is critical examine both in normal operation and in the case of accidents that are established by international bodies. for appropriate analytical methods that seek to achieve is use a appropriate representation of the behavior of the structure. in addition to strengthen the theory computer simulations of the tests used applied, where the results will be contrasted with the first method of calculation. Results are obtained for the stress field and displacement total delivering the information necessary to approve the container

  17. On-Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Very High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (∼ 1-mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4%-10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

  18. 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

  19. Status and development of instrumented fuel rod testing simulating the power reactor operating conditions in the research reactor MIR

    International Nuclear Information System (INIS)

    The paper includes the analysis of parameters for instrumented refabricated fuel rods having high burnup under power cycling as an example. The SSC RF RIAR carries out all activities, which are required for testing. The whole cycle of such activities, with the experiment with irradiated fuel rods of WWER fuel assembly has been carried out in the MIR reactor, is shown. The zero power reactor MIR (critical assembly) is used for experimental support and safety analysis. The main characteristics of the loop facilities, which are used for testing and the designs of several devices are presented. The location of fuel rods and sensors throughout the height and over the cross-section of the experimental device is shown. The changes of fuel rod (A) length and diameter as a function of average linear power throughout the fuel rod height and the relation between the length and diameter changes are given. The experiments with WWER-440 and WWER-1000 refabricated fuel rods having a burnup of 50 to 60 MWd/kgU approximately were carried out under repetitive power cycling conditions. The curves of linear power change in these experiments are represented schematically. Some results of the data processing, which are logged under the transient conditions during the first experiment with WWER instrumented refabricated fuel rods having a burnup of 50 MWd/kgU approximately are given. The change of the logged data on fuel temperature and neutron flux and changes in the length of fuel rods in the course of power change after different number of cycles are demonstrated. The experimental capabilities improvements like: 1) Development of the technique and experimental device for testing of WWER fuel rods under conditions simulating some design-basis accidents accompanied with pulse power change; 2) Improvement of dynamic techniques used for analysis of thermal and physical parameters of the WWER fuel rods, calculation of local power and heat-transfer factor with the help of correlation and

  20. 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

  1. 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

  2. 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)

  3. Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element

    International Nuclear Information System (INIS)

    The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results are related to the existing models and a mechanistic model for the ''annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs

  4. Nuclear physics calculation for IMF-MOX fuel irradiation test in the halden boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Noh, Jae Man; Kim, Ha Yong [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    As a part of activity for future fuel development project in KAERI, test MOX fuel rods are going to be loaded and irradiated in Halden reactor under a KAERI's joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device developed by KAERI. The test fuel assembly rig contains total six test fuel rods: three MOX rods and three inert matrix fuel rods. For comparison purposes, one of three MOX rods will be fabricated by BNFL and the other two MOX rods will be manufactured jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with (Zr-Y-Er-Pu) oxide by PSI. One of the three IMF rods will be fabricated with wet process and the other two rods will be fabricated with dry process. Neutronic calculation was preliminarily performed for the test fuel assembly to generate the rod power histories of test rods which will be use to evaluate the irradiation performance of the test fuel rods. A power scenario is proposed to meet the experiment conditions, target linear power and burnup. In accordance with this power scenario, the maximum linear power rate is about 350 W/cm, considering the axial power peaking factor of 1.05, for both IMF and MOX. After five years irradiation, the achieved burnup will be about 49 MWD/kgHM for MOX and 490 MWD/kgHM for IMF. To get the proposed power scenario, the expected irradiation position may be in the ring 6 of the HBWR core the position of test fuel to be loaded in HBWR will also depend on the environment. A surrounding of these spikes, as considered, in a symmetric configuration would give a flat power distribution in the IMF/MOX-IFA and hence the most successful results. (author). 21 figs., 4 tabs.

  5. The BOR-60 loop-channel design for testing the BREST reactor fuel

    International Nuclear Information System (INIS)

    The paper is devoted to development and calculated substantiation of the design of the autonomous lead-cooled loop (AILCL) intended for testing the fuel pins prototypes in the BOR-60 reactor for the BREST-OD-300 reactor. The design features of the loop, its characteristics, instrumentation are considered. The auxiliary systems required to provide the loop operation are described. The main neutron-physical and thermohydraulic characteristics of the loop are presented. The basic operating conditions of the loop are shown. Great consideration is given to analysis of faults and failures affecting the loop serviceability and reactor safety. It is shown that the BOR-60 reactor safety is provided under all normal modes and postulated failures. (author)

  6. Residual stress and strain examination in peach bottom fuel test elements

    International Nuclear Information System (INIS)

    An examination of residual stresses and strains has been carried out experimentally on structural graphite components removed from the Peach Bottom high temperatures gas cooled reactor (HTGR) fuel elements. The purpose is to confirm predicted stress distributions. Twenty-nine teledial fuel elements (of six and eight hole design) were irradiated in the Peach Bottom HTGR to fluences of 25 (E>29fJ) and time average temperatures up to 10000C. The irradiation history was modeled with HTGR design codes. Performance predictions were verified by in-pile thermocouples and fuel burnup measurements. The predicted irradiation induced strains were found to be in quantitative agreement with post irradiation measurements (<=0.5% shrinkage over 70 mm diameter). As a result of the temperature distribution in the teledial bodies, compressive stresses were predicted at the periphery and tensile stresses in the inner zones. A total of thirteen graphite bodies from eight different fuel elements were examined destructively for structural integrity and residual stresses. For the higher exposure elements, local stresses at shutdown in excess of the uniaxial strength of the material were predicted, yet no structural damage was found. This may be because uniaxial failure criteria are conservative. Axial strips of 460 mm length were cut and measured for changes in curvature which indicate release of residual axial stresses. Release of compressive residual stresses were observed from slight extension of the strips, when compared with length measurements before cutting

  7. 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.)

  8. Study on usage of low enriched uranium Russian type fuel elements for design of an experimental ADS research reactor

    International Nuclear Information System (INIS)

    Conceptual design of an accelerator driven sub-critical experimental research reactor (ADSRR) was initiated in 1999 at the Vinca Institute of Nuclear Sciences, Serbia and Montenegro. Initial results of neutronic analyses of the proposed ADSRR-H were carried out by Monte Carlo based codes and available high-enriched uranium dioxide (HEU) dispersed Russian type TVR-S fuel elements (FE) placed in a lead matrix. Beam of charged particles (proton or deuteron) would be extracted from the high-energy channel H5B of the VINCY cyclotron of the TESLA Accelerator Installation. In 2002, the Vinca Institute has, in compliance with the Reduced Enrichment for Research and Test Reactors (RERTR) Program, returned fresh HEU TVR-S type FEs back to the Russian Federation. Since usage of HEU FEs in research reactors is not further recommended, a new study of an ADSRR-L conceptual design has initiated in Vinca Institute in last two years, based on assumed availability of low-enriched uranium (LEU) dispersed type TVR-S FEs. Initial results of numerical simulations of this new ADSRR-L, published for the first time in this paper, shows that such a small low neutron flux system can be used as an experimental - 'demonstration' - ADS with neutron characteristics similar to proposed well-known lead moderated and cooled power sub-critical ADS with intermediate neutron spectrum. Neutron spectrum characteristics of the ADSRR-L are compared to ones of the ADSRR-H with the same mass (7.7 g) of 235U nuclide per TVR-S FE. (author)

  9. 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

  10. Direct ultimate disposal of spent fuel elements. Mechanical equipment tests

    International Nuclear Information System (INIS)

    Simulation of the shaft transport of waste forms is described. Proceeding from a concept of a shaft hoist with a payload of 85 t, the applicability of the state of the art of essential components, such as hoisting machine, cage and hoisting cables, to such payloads is described. For these components a test stand has been planned which meets safety-related regulations. (DG)

  11. Production of Iodine-132 from Irradiated Test Fuel Elements

    International Nuclear Information System (INIS)

    The preparation of iodine-132 from irradiated uranium is described. After irradiation and cooling, tellurium-132 is isolated from other fission products and the uranium recovered. Extraction of iodine-132 from tellurium-132 is performed either by an ion-exchange method or by adsorption on platinum. Purity tests for the product iodine-132 are described and examples of applications are reviewed. (author)

  12. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    The description is given of a nuclear reactor fuel assembly comprising fuel elements arranged in a supporting frame composed of two end pieces, one at the top and the other at the bottom, on which are secured the ends of a number of vertical tubes, each end piece comprising a plane bottom on which two series of holes are made for holding the tubes and for the passage of the coolant. According to the invention, the bottom of each end piece is fixed to an internal plate fitted with the same series of holes for holding the tubes and for the fluid to pass through. These holes are of oblong section and are fitted with fixing elements cooperating with corresponding elements for securing these tubes by transversal movement of the inside plate

  13. 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)

  14. 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)

  15. 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)

  16. 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)

  17. Comparison of some measurements in the Greek Research Reactor-1 (GRR-1) with 20% and 93% enriched uranium fuel elements

    International Nuclear Information System (INIS)

    Experimental results of the core performance of the GRR-1 with 20% and 93% enriched 235U are compared. In particular, the rod worth, the flux distribution (axial and radial), the temperature, and the void effects are analyzed and compared for the two cores. The experimental results are in good agreement with benchmark calculations for this reactor and for other similar reactors. Although the fuel elements of the two cores do not have the same number of plates, the results are of special significance, since they present the only experimental comparative work known so far on the core conversion problem from the use of highly enriched uranium to the use of low enriched uranium fuels

  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. 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

  20. State of fuel elements in BN-600 reactor fuel assembly 917137489 when reaching maximal damaging dose of 93.7 dpa

    International Nuclear Information System (INIS)

    A complex of post-irradiated studies was accomplished for BN-600 reactor fuel cans of steel ChS-68. The studies were carried out by methods of profilometry, metallography short-term mechanical tests after irradiation up to damaging dose of 93.7 dpa. High volumetric changes, zero plasticity, corrosion defects testify to the fact that further operation of the steel under conditions of BN-600 reactor up to damaging doses more than 91-93 dpa is inadmissible

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

    International Nuclear Information System (INIS)

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

  2. Residual stress and strain examination in Peach Bottom fuel test elements

    International Nuclear Information System (INIS)

    An examination of residual stresses and strains has been carried out experimentally on structural graphite components removed from Peach Bottom High-Temperature Gas-Cooled Reactor (HTGR) fuel elements. The purpose of this work is to confirm predicted stress distributions. Twenty-nine teledial fuel elements (of six- and eight-hole design) were irradiated in the Peach Bottom HTGR to fluences of less than or equal to 4 x 1025 n/m2 (E greater than 29 fJ) and time-average temperatures up to 11000C. The irradiation history was modeled with HTGR design codes. Performance predictions were verified by in-pile thermocouples and fuel burnup measurements. The predicted irradiation-induced strains were found to be in quantitative agreement with postirradiation measurements (less than or equal to 0.5 percent shrinkage over a 70-mm diameter). As a result of the temperature distribution in the teledial bodies, compressive stresses were predicted at the periphery and tensile stresses in the inner zones. A total of 19 graphite bodies from 8 different fuel elements were examined destructively for structural integrity and residual stresses

  3. 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)

  4. 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)

  5. 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

  6. Preliminary results of the BTF-104 experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

    International Nuclear Information System (INIS)

    The BTF-104 experiment is one of a series of in-reactor tests being performed to measure fuel behaviour and fission-product release from nuclear fuel subjected to accident conditions. The primary objective of the BTF-104 experiment was to measure fission-product releases from a CANDU-sized fuel element under combined Loss-of-Coolant Accident (LOCA) and Loss-of-Emergency-Core-Cooling (LOECC) conditions at an average fuel temperature of about 1550 deg C. The preliminary results of the BTF-104 experiment are presented in this paper. (author). 6 refs., 12 figs

  7. 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.)

  8. 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)

  9. The interactions of the nuclear reactor fuel element components during LOCA

    International Nuclear Information System (INIS)

    The objective of the study has been to investigate the phase interactions and melting phenomena that result when the LWR core temperature is increased due to the loss of coolant accident. Diffusin couple composed of UO2-Zircaloy and steel as well as models of fuel elements were used in experiments. In the inert atmosphere low-melting metallic eutectics appear. In oxidizing atmosphere the melting temperature is pushed towards higher values due to the oxidation of metallic components. The attendant embrittlement of fuel cladding can lead to simultaneous collapse of fuel rods. The intensive melting is observed above 1773 deg K, steel spacers melting first. The melt dissolves the parts of Zircaloy cladding and penetrates between UO2 pellets. (author)

  10. Modernization of OIK unit for fabrication of mixed-oxide fuel, vibrocompacted fuel elements and fuel-containing assemblies of BN-600 reactor using plutonium of weapon quality

    International Nuclear Information System (INIS)

    In the framework of participation in international project of weapon plutonium utilization modernization of the technological complex for fabrication of granulated fuel, vibrocompacted fuel elements and fuel-containing assemblies is realizing. Taking into account domestic and foreign experience of MOX-fuel fabrication different versions of equipment are examined

  11. Fabrication of Confinement Facility of Failed Fuel Elements

    International Nuclear Information System (INIS)

    The confinement facility of failed fuel elements is provide for isolating the elements so that their fission product could not contaminate reactor pool. Since RSG-GAS does not have such facility yet, the fabrication of the confinement is compulsory needed. The fabrication of confinement was initialized by providing technical drawing, materials procurement, fabricating and testing, each confinement capacity is 2 elements. The test result showed that the facility can be used to store the two failed fuel elements safely. (author)

  12. Development of IRT-type fuel assembly with pin-type fuel elements for LEU conversion of WWR-SM research reactor in Uzbekistan

    International Nuclear Information System (INIS)

    The principle design of an IRT-type FA with pin-type fuel elements has been developed. This work is being conducted in two main directions: 1. technological research on fabrication of FA as a whole and of its constituent parts; 2. calculational research on substantiation of IRT-type FE and FA designs with regard to the WWR-SM reactor (Uzbekistan). At present, large set of technological investigations of FE and FA fabrication processes and large set of neutronics, thermal hydraulic and strength calculations has been performed. These research of FE and FA design are being performed in close cooperation with NZChK, ANL and RRC KI within Russian and international RERTR programs. The dispersion fuel U-9%Mo in Al matrix is chosen. The optimum U-235/FA loading is de-fined. Two full-scaled dummy of FA for hydraulic tests have been manufactured jointly by Bochvar Institute and Novosibirsk plant (NZChK). Preliminary results of joint studies are presented in paper. (author)

  13. Safety of operations in the manufacture of driver fuel for the first charge of the Dragon Reactor and modifications to the safety document for the Dragon Fuel Element Production Building

    International Nuclear Information System (INIS)

    The manufacture of the zirconium containing 'driver' fuel and fuel elements for the First Charge of the Dragon Reactor Experiment has been completed without incident. This is a report on the safety of operations in the Dragon Fuel Element Production Building during an approximately six month period when the 'driver' fuel was manufactured and 25 elements containing this fuel were assembled and exported to the Reactor Building. The opportunity is taken to bring the Safety Document up-to-date and to report on any significant operational failures of equipment. (author)

  14. Preliminary results of the BTD-105B experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

    International Nuclear Information System (INIS)

    The BTF-105B test is one of a series of in-reactor, all-effects, tests performed to measure fuel behaviour, fission-product release, and transport from nuclear fuel subjected to accident conditions. The primary objective of the BTF-105B test was to determine the timing, amount, and transport characteristics of fission products released from a previously irradiated CANDU-sized fuel element subjected to a high-temperature transient, representative of a loss-of-coolant accident with loss-of-emergency-core-cooling conditions, at an average fuel temperature of 1800 deg C. The preliminary results of the BTF-105B test are presented in this paper. The results include process parameters, from which boundary conditions could be derived for simulating the test; fuel parameters, such as sheath temperatures and hydrogen production rate; and fission-product release and transport data. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

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

  16. 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

  17. 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)

  18. Contribution to the elaboration of the neutronic calculation scheme for the light water reactors with slab fuel elements

    International Nuclear Information System (INIS)

    The use of slab fuel elements in the light water reactors is interesting for the thermohydraulic aspects but, of course, induces some difficulties for the neutronic calculations. This thesis is devoted to the discussion of some of them: comparison with the pin lattices, resonance capture by 238U, effective temperature for the Doppler effect, reduction of the number of energy groups, adjustment of the fission cross section of 235U in order to improve the temperature coefficient calculation. A complete calculation scheme is proposed and checked against experiment

  19. 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)

  20. 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

  1. Penetration depth testing the fuel element cladding and end element joint by eddy currents

    International Nuclear Information System (INIS)

    Advantage of the eddy method of testing the fuel cladding penetration depth as compared with ultrasonic and X-ray ones presupposes increase of sensitivity to low penetration and an opportunity of testing the penetration near the edge. Increase of the sensitivity is attained by means of an aerostatic bearing, on the working edge of which induction coils are dislocated, and increase of edge effect- by means of a superposed differential converter with sector windings, the adjacent linearly stretched sections of which are located at the right angle. At the diameter of the converter of 12 mm, nozzle chamber diameter of 5 mm, diffuser diameter of 7 mm the measuring range of the penetration depth makes up 0-1.1 mm

  2. 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

  3. 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

  4. 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

  5. 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

  6. BN-600 fuel elements and fuel assemblies operating experience

    International Nuclear Information System (INIS)

    Consideration is given to the data on fuel burnup of standard fuel assemblies of the BN-600 reactor first core charge and that for modified core; data on operation ability of fuel assemblies of the first charge type are given. Data on main results of primary post-irradiation examination of fuel assemblies and fuel elements and maximal values of fuel burnup, achieved in particular fuel assemblies of BN-600 reactor are presented. 4 figs.; 1 tab

  7. Reprocessing ability of high density fuels for research and test reactors

    International Nuclear Information System (INIS)

    The development of a new high density fuel is becoming a key issue for Research Reactors operators. Such a new fuel should be a Low Enrichment Uranium (LEU) fuel with a high density, to improve present in core performances. It must be compatible with the reprocessing in an industrial plant to provide a steady back-end solution. Within the framework of a work group CEA/CERCA/COGEMA on new fuel development for Research Reactors, COGEMA has performed an evaluation of the reprocessing ability of some fuel dispersants selected as good candidates. The results will allow US to classify these fuel dispersants from a reprocessing ability point of view. (author)

  8. 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

  9. 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.

  10. In-pile tests and post-reactor investigations of fuel pins with U-Pu Fuel

    International Nuclear Information System (INIS)

    Resulting of testing experimental fuel subassemblies with oxide uranium-plutonium fuel and data of fuel pin investigations at the IPPE hot laboratory have been presented. MOX and UOX fuels behaviour under irradiation are compared. (author)

  11. Study on fracture of fuel element cladding for naval reactor during typical accidents

    International Nuclear Information System (INIS)

    Aiming at defining the grade of nuclear emergency response, the best estimate model has been adopted; the simulation of large break loss of coolant accident (LBLOCA) has been carried out by the radioactive analysis software coupled with relap5/mod 3.2 and core physics model. First, the peak clad temperature of the critical failure channel is calculated in relap5 code, and simultaneously its power factor is obtained. Second, pin power distribution of the fuel assemblies has been calculated in coarse-mesh nodal method. According to the pin power distribution in the whole core and the result gained above, the fraction of fuel element fracture is calculated. Finally, the radioactive analysis has been carried out and the reasonable source term is gotten, which can offer reference for the nuclear emergency decision making. (authors)

  12. MTR fuel plate qualification in OSIRIS reactor

    International Nuclear Information System (INIS)

    Qualification of new MTR fuel needs the irradiation in research reactors under representative neutronic, heat flux and thermohydraulic conditions. The experiments are performed in France in the OSIRIS reactor by irradiating MTR full size fuel plates in the IRIS device located in the reactor core. The fuel plates are easily removed from the device during the shutdown of the reactor for performing thickness measurements along the plates by means of a swelling measurement device. Beside the calculation capabilities, the experimental platform includes: the ISIS neutron mock-up for the measurement of neutron flux distribution along the plates; the γ spectrometry for the purpose of measuring the activities of the radionuclides representative of the power and the burnup and to compare with the neutronic calculation. Owing to the experience feedback, a good agreement is observed between calculation and measurement; destructive post irradiation examinations in the LECA facility (Cadarache). New irradiations with the IRIS device and at higher heat flux are under preparation for qualification of MTR fuels. (author)

  13. 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.

  14. Thermochemistry of nuclear fuels in advanced reactors

    International Nuclear Information System (INIS)

    The presence of a large number of elements, accompanied with steep temperature gradient results in dynamic chemistry during nuclear fuel burn-up. Understanding this chemistry is very important for efficient and safe usage of nuclear fuels. The radioactive nature of these fuels puts lot of constraint on regulatory bodies to ensure their accident free operation in the reactors. One of the common aims of advanced fuels is to achieve high burn-up. As burn-up of the fuel increases, chemistry of fission-products becomes increasingly more important. To understand different phenomenon taking place in-pile, many out of-pile experiments are carried out. Extensive studies of thermodynamic properties, phase analysis, thermophysical property evaluation, fuel-fission product clad compatibility are carried out with relevant compounds and simulated fuels (SIMFUEL). All these data are compiled and jointly evaluated using different computational methods to predict fuel behaviour during burn-up. Only when this combined experimental and theoretical information confirms safe operation of the pin, a test pin is prepared and burnt in a test reactor. Every fuel has a different chemistry and different constraints associated with it. In this talk, various thermo-chemical aspects of some of the advanced fuels, mixed carbide, mixed nitride, 'Pu' rich MOX, 'Th' based AHWR fuels and metallic fuels will be discussed. (author)

  15. Startup testing of Romania dual-core test reactor

    International Nuclear Information System (INIS)

    Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

  16. 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)

  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. LEU fuel element produced by the Egyptian fuel manufacturing pilot plant

    International Nuclear Information System (INIS)

    The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a Material Testing Reactor type (MTR) fuel element facility, for producing the specified fuel elements required for the Egyptian Second Research Reactor, ETRR-2. The plant uses uranium hexafluoride (UF6, 19.75% U235 by wt) as a raw material which is processed through a series of the manufacturing, inspection and test plan to produce the final specified fuel elements. Radiological safety aspects during design, construction, operation, and all reasonably accepted steps should be taken to prevent or reduce the chance of accidents occurrence. (author)

  19. Spent fuel management for research reactors

    International Nuclear Information System (INIS)

    There are six research reactors in Argentina using fuel elements uranium enriched from 20 to 90%. Spent fuel elements management is limited to RA-1 (Argonaut type used for training) and RA-3 (for experimentation and radioisotope production), as for the others no changes have been carried out over the cores. The first core of RA-1 was reprocessed and the second core was manufactured with U3O8 obtained from reprocessing, and once spent was transferred to a dry storage. RA-3 is pool type and fuel elements are MTR enriched at 90% 24 fuel elements in the core. Up to now 238 fuel elements have been used with burn-up from 16 to 40% and about 8 gU per plate. At present RA-3 is being remodelled in order to be able to use fuel elements enriched to 20%. After decay in a pool at reactor building, they are transported to a wet storage facility for spent fuel elements. Control rods are treated in the same way. From 15 years of storage facility operation, it follows as a consequence that the proposed objectives have been fulfilled sufficiently well, as shown by radioactivity measurements on the water hole lines (5 x 10-4 Ci/m3 of Cs-137) and the neutron interrogation method on the spent fuel elements tested presented values less than 10-8 gU/cm3, that was considered acceptable). Nevertheless, some corrosion areas were observed. Program activities involve on one hand mainly new chemical and physical controls in order to optimize spent fuel storage and, on the other hand, the study of the corrosion observed with the aim to minimize it and to ensure safe storage. (author). 6 figs, 2 tabs

  20. Criticality analysis for storage and shipping of nuclear fuel elements outside of reactors

    International Nuclear Information System (INIS)

    A calculation method for criticality of PWR elements in storage pool, shipping cask, reprocessing plant... was developed and tested. This method is based on use of computer codes Apollo and Dot 3.5, solving neutron propagation equation in transport theory. Its validity was checked by comparing with results of some criticality experiences and other Benchmarks obtained by computer Monte Carlo code Tripoli