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Sample records for booster fuel assembly

  1. Yalina booster subcritical assembly performance with low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)

    2011-07-01

    The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

  2. 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.

  3. Fuel gas boosters in developing nations

    Energy Technology Data Exchange (ETDEWEB)

    Sumbles, Billy [VALERUS, Ahmadabad (India). South East Asia Regional; Paris, Mike [VALERUS, Houston, TX (United States)

    2008-07-01

    Regulations requiring the use of natural gas, as opposed to flaring, and the need for clean generation solutions are driving global demand for gas-fired generation and corresponding fuel gas booster stations. Gas-producing nations are looking for viable solutions to be able to capture and use natural gas. And nations in the Middle East, Asia and Europe are developing a gas infrastructure similar to the U.S. Gas transported across long distances in pipelines often requires compression to boost pressure and meet industrial needs. In many cases, the gas may need to be processed before it can be used. However, end-users cannot just order gas booster stations out of a catalogue. Several factors - such as turbine operation, gas quality, compression pressure requirements, and the extent of processing required - need to be considered before designing and installing such systems. (author)

  4. Fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To reconstruct a BWR type reactor into a high conversion reactor with no substantial changes for the reactor inner structure such as control rod structure. Constitution: The horizontal cross sectional shape of a channel box is reformed into a square configuration and the arrangement of fuel rods is formed as a trigonal lattice-like configuration. As a method of improving the conversion ratio, there is considered to use a dense lattice by narrowing the distance between fuel rods and trigonal lattice arrangement for fuel rod is advantageous therefor. A square shape cross sectional configuration having equal length both in the lateral and longitudinal directions is suitable for the channel box as a guide upon movement of the control rod. Fuel rods can be arranged with no loss by the trigonal lattice configuration, by which it is possible to improve the neutron moderation, increase the reactor core reactivity and conduct effective fuel combustion. In this way, it is possible to attain the object by inserting the follower portion of the control rod at the earier half and extracting the same at the latter half during the operation period in the reactor core comprising fuel assemblies suitable to a high conversion BWR type reactor having average conversion ratio of about 0.8. (Kamimura, M.)

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Since the neutron flux distribution and the power distribution of a fuel assembly in which short fuel rods vary greatly in the vicinity of a boundary where the distribution of uranium amount is different, the reading value of local power range monitors, having the detectors positioned in the vicinity of the boundary is varied. Then in the present invention, the upper end of the effective axial length of fuel rod is so made as not approaching with the detection position of the local power range monitor in a reactor core. Further, the upper end of the effective axial length of fuel rods in a 4 x 4 fuel rod lattice positioned at the corner on the side of the local power range monitor is so made as not approaching the detection position of the local power range monitor. As a result, the change of the neutron flux distribution and power distribution in the vicinity of the position where the detector of the local power range monitor is situated can be extremely reduced. Accordingly, there is no scattering and fluctuation for the reading value by the local power range monitor, to improve the monitoring performance for thermal characteristics in the reactor core. (N.H.)

  6. YALINA-booster subcritical assembly pulsed-neutron experiments : data processing and spatial corrections.

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Y.; Gohar, Y.; Nuclear Engineering Division

    2010-10-11

    The YALINA-Booster experiments and analyses are part of the collaboration between Argonne National Laboratory of USA and the Joint Institute for Power & Nuclear Research - SOSNY of Belarus for studying the physics of accelerator driven systems for nuclear energy applications using low enriched uranium. The YALINA-Booster subcritical assembly is utilized for studying the kinetics of accelerator driven systems with its highly intensive D-T or D-D pulsed neutron source. In particular, the pulsed neutron methods are used to determine the reactivity of the subcritical system. This report examines the pulsed-neutron experiments performed in the YALINA-Booster facility with different configurations for the subcritical assembly. The 1141 configuration with 90% U-235 fuel and the 1185 configuration with 36% or 21% U-235 fuel are examined. The Sjoestrand area-ratio method is utilized to determine the reactivities of the different configurations. The linear regression method is applied to obtain the prompt neutron decay constants from the pulsed-neutron experimental data. The reactivity values obtained from the experimental data are shown to be dependent on the detector locations inside the subcritical assembly and the types of detector used for the measurements. In this report, Bell's spatial correction factors are calculated based on a Monte Carlo model to remove the detector dependences. The large differences between the reactivity values given by the detectors in the fast neutron zone of the YALINA-Booster are reduced after applying the spatial corrections. In addition, the estimated reactivity values after the spatial corrections are much less spatially dependent.

  7. Fuel Assembly Damping Summary

    International Nuclear Information System (INIS)

    This paper summary the fuel assembly damping data in air/in still water/under flow, released from foreign fuel vendors, compared our data with the published data. Some technical issues in fuel assembly damping measurement testing are also briefly discussed. Understanding of each fuel assembly damping mechanisms according to the surrounding medium and flow velocity can support the fuel design improvement in fuel assembly dynamics and structural integrity aspect. Because the upgraded requirements of the newly-developed advanced reactor system will demands to minimize fuel design margin in integrity evaluation, reduction in conservatism of fuel assembly damping can contribute to alleviate the fuel design margin for sure. Damping is an energy dissipation mechanism in a vibrating mechanical structure and prevents a resonant structure from having infinite vibration amplitudes. The sources of fuel assembly damping are various from support friction to flow contribution, and it can be increased by the viscosity or drag of surrounding fluid medium or the average velocity of water flowing. Fuel licensing requires fuel design evaluation in transient or accidental condition. Dynamic response analysis of fuel assembly is to show fuel integrity and requires information on assembly-wise damping in dry condition and under wet or water flowing condition. However, damping measurement test for the full-scale fuel assembly prototype is not easy to carry out because of the scale (fuel prototype, test facility), unsteadiness of test data (scattering, random sampling and processing), instrumentation under water flowing (water-proof response measurement), and noise. LWR fuel technology division in KAERI is preparing the infra structure for damping measurement test of full-scale fuel assembly, to support fuel industries and related research activities. Here is a preliminary summary of fuel assembly damping, published in the literature. Some technical issues in fuel assembly damping

  8. Spent fuel assembly hardware

    International Nuclear Information System (INIS)

    When spent nuclear fuel is disposed of in a repository, the waste package will include the spent fuel assembly hardware, the structural portion of the fuel assembly, and the fuel pins. The spent fuel assembly hardware is the subject of this paper. The basic constituent parts of the fuel assembly will be described with particular attention on the materials used in their construction. The results of laboratory analyses performed to determine radionuclide inventories and trace impurities also will be described. Much of this work has been incorporated into a US Department of Energy (DOE) database maintained by Oak Ridge National Laboratory (ORNL). This database is documented in DOE/RW-0184 and can be obtained from Karl Notz at ORNL. The database provides a single source for information regarding wastes that may be sent to the repository

  9. Yalina-Booster Assembly: from HEU to LEU

    International Nuclear Information System (INIS)

    The YALINA facility is a unique facility which was designed as a zero power model of real ADS (Accelerator Driven System). It is intended to study ADS neutronics and kinetics of the subcritical reactors driven by external neutron sources. Accelerator-driven systems may play an important role in future nuclear fuel cycles to reduce the long-term radiotoxicity and volume of spent nuclear fuel. Successful operation of this facility is a scientific contribution from the Republic of Belarus, as well as the international community. The experimental data are used to benchmark and validate methods and computer codes for designing and licensing ADS. In this paper the investigation of spatial kinetics of the sub-critical systems with external neutron sources, validation of the experimental techniques for sub-criticality monitoring and estimation of probability of minor actinides and fission products transmutation is made for different configurations of Yalina-Booster during conversion from HEU to LEU: - 1st configuration – with HEU fuel in the fast zone (metallic uranium of 90% enrichment by 235U and UO2 of 36% enrichment by 235U) and uranium dioxide of 10% enrichment by 235U in thermal zone; - 2nd one – with 36% UO2 in fast zone and 10% in thermal zone; - 3rd one – with 21% UO2 in fast zone and 10% in thermal zone; - 4th one – with 21% UO2 in fast zone and 10% in thermal zone, differing from the 3rd configuration by rounded shape of fast zone and annular shape of the absorber zone with permanent number of the absorbing rods. (author)

  10. Fuel nozzle assembly

    Science.gov (United States)

    Johnson, Thomas Edward; Ziminsky, Willy Steve; Lacey, Benjamin Paul; York, William David; Stevenson, Christian Xavier

    2011-08-30

    A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

  11. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To increase the fuel assembly rigidity while making balance in view of the dimension thereby improving the earthquake proofness. Constitution: In a nuclear fuel assembly having a control rod guide thimble tube, the gap between the thimble tube and fuel insert (inner diameter of the guiding thimble tube-outer diameter of the fuel insert) is made greater than 1.0 mm. Further, the wall thickness of the thimble tube is made to about 4 - 5 % of the outer diameter, while the flowing fluid pore cross section S in the thimble tube is set as: S = S0 x A0/A where S0: cross section of the present flowing fluid pore, A: effective cross section after improvement, = Π/4(d2 - D2) in which d is the thimble tube inner diameter and the D is the fuel insert outer diameter. A0: present effective cross section. (Seki, T.)

  12. Transfer of fuel assemblies

    International Nuclear Information System (INIS)

    Fuel assemblies of a nuclear reactor are transferred during fueling or refueling or the like by a crane. The work-engaging fixture of the crane picks up an assembly, removes it from this slot, transfers it to the deposit site and deposits it in its slot at the deposit site. The control for the crane includes a strain gauge connected to the crane line which raises and lowers the load. The strain gauge senses the load on the crane. The signal from the strain gauge is compared with setpoints; a high-level setpoint, a low-level setpoint and a slack-line setpoint. If the strain gauge signal exceeds the high-level setpoint, the line drive is disabled. This event may occur during raising of a fuel assembly which encounters resistance. The high-level setpoint may be overridden under proper precautions. The line drive is also disabled if the strain gauge signal is less than the low-level setpoint. This event occurs when a fuel assembly being deposited contacts the bottom of its slot or an obstruction in, or at the entry to the slot. To preclude lateral movement and possible damage to a fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge exceets the lov-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than the slack-line set-point. This occurs when the lines has been set in slack-line setting. When the line is tensioned after slack-li ne setting, the traverse drive remains enabled only if the line has been disconnected from the fuel assembly

  13. Fuel assembly supporting structure

    International Nuclear Information System (INIS)

    For use in forming the core of a pressurized-water reactor, a fuel assembly supporting structure for holding a bundle of interspaced fuel rods, is formed by interspaced end pieces having holes in which the end portions of control rod guide tubes are inserted, fuel rod spacer grids being positioned by these guide tubes between the end pieces. The end pieces are fastened to the end portions of the guide tubes, to integrate the supporting structure, and in the case of at least one of the end pieces, this is done by means which releases that end piece from the guide tubes when the end pieces receive an abnormal thrust force directed towards each other and which would otherwise place the guide tubes under a compressive stress that would cause them to buckle. The spacer grids normally hold the fuel rods interspaced by distances determined by nuclear physics, and buckling of the control rod guide tubes can distort the fuel rod spacer grids with consequent dearrangement of the fuel rod interspacing. A sudden loss of pressure in a pressurized-water reactor pressure vessel can result in the pressurized coolant in the vessel discharging from the vessel at such high velocity as to result in the abnormal thrust force on the end pieces of each fuel assembly, which could cause buckling of the control rod guide tubes when the end pieces are fixed to them in the normal rigid and unyielding manner

  14. Fuel assemblies chemical cleaning

    International Nuclear Information System (INIS)

    NPP Paks found a thermal-hydraulic anomaly in the reactor core during cycle 14 that was caused by corrosion product deposits on fuel assemblies (FAs) that increased the hydraulic resistance of the FAs. Consequently, the coolant flow through the FAs was insufficient resulting in a temperature asymmetry inside the reactor core. Based on this fact NPP Paks performed differential pressure measurements of all fuel assemblies in order to determine the hydraulic resistance and subsequently the limit values for the hydraulic acceptance of FAs to be used. Based on the hydraulic investigations a total number of 170 FAs was selected for cleaning. The necessity for cleaning the FAs was explained by the fact that the FAs were subjected to a short term usage in the reactor core only maximum of 1,5 years and had still a capacity for additional 2 fuel cycles. (authors)

  15. Seismic behaviour of fuel assembly

    International Nuclear Information System (INIS)

    A general approach for the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced from earthquake. The dynamic responses such as fuel assembly shear force, bending moment and displacement, and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed. (Author) 9 refs., 24 figs., 1 tab

  16. Seismic behaviour of fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Song, Heuy Gap; Jhung, Myung Jo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-11-01

    A general approach for the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced from earthquake. The dynamic responses such as fuel assembly shear force, bending moment and displacement, and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed. (Author) 9 refs., 24 figs., 1 tab.

  17. YALINA Booster conversion project

    International Nuclear Information System (INIS)

    The YALINA Booster subcritical assembly was constructed at the Joint Institute for Power and Nuclear Research (JIPNR)-SOSNY, Belarus to examine the physics of Accelerator Driven Systems (ADS). The assembly has fast and thermal zones to study the coupling between the two zones, the transuranics transmutation, and the ADS kinetics. It is driven by external neutron source located at the assembly center. The central fast zone (the booster zone) consists of high enriched uranium (HEU) fuel rods loaded in a lead matrix and it is surrounded by thermal zone. The thermal zone has low enriched uranium (LEU) fuel rods loaded in polyethylene moderator. Between the two zones, there is a thermal neutron absorber zone. (JIPNR)-SOSNY has an International Science and Technology Center project in collaboration with Argonne National Laboratory of USA to convert the HEU fuel of YALINA Booster to LEU fuel without penalizing its performance. The first step of this project is to characterize and define the performance of the YALINA Booster subcritical assembly with HEU fuel by performing detailed analytical and experimental studies. The second step is to convert the booster zone to use uranium fuel rods with 21% enrichment. The YALINA Booster configuration is modified to reach the original subcriticality level. The analytical analyses have developed accurate calculational models without geometrical approximations for performing Monte Carlo and Deterministic calculations. MCNP, MCNPX, MCB, MONK, ERANOS, and PARTISN computer codes with different nuclear data libraries based on ENDF/VI, JEF2.2, and JEF3.1 have been used for static and kinetic analyses. The geometrical details are included explicitly without approximation or homogenization. In the experimental program, the subcriticality has been measured as a function of the number of the fuel rods loaded in the subcritical assembly. Different methods have been used to measure the assembly subcriticality during the fuel loading process. In

  18. FUEL ASSEMBLY SHAKER TEST SIMULATION

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.; Hanson, Brady D.

    2013-05-30

    This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through

  19. Monte Carlo modeling and analyses of YALINA- booster subcritical assembly Part II : pulsed neutron source.

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, A.; Gohar, M. Y. A.; Rabiti, C.; Nuclear Engineering Division

    2008-10-22

    One of the most reliable experimental methods for measuring the kinetic parameters of a subcritical assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology for characterizing the kinetic parameters of a subcritical assembly using the Sjoestrand method, which allows comparing the analytical and experimental time dependent reaction rates and the reactivity measurements. In this methodology, the reaction rate, detector response, is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the fission delayed neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction is vanished. The obtained reaction rate is superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The new calculation methodology has shown an excellent agreement with the experimental results available from the YALINA-Booster facility of Belarus. The facility has been driven by a Deuterium-Deuterium or Deuterium-Tritium pulsed neutron source and the (n,p) reaction rate has been experimentally measured by a {sup 3}He detector. The MCNP calculation has utilized the weight window and delayed neutron biasing variance reduction techniques since the detector volume is small compared to the assembly volume. Finally, this methodology was used to calculate the IAEA benchmark of the YALINA-Booster experiment.

  20. Alternative systems for fuel gas boosters for small gas turbine engines

    Science.gov (United States)

    Faulkner, Henry B.

    1992-04-01

    The study was done to investigate alternative technologies for fuel gas boosters for gas turbine engines under 5 MW output. The goal was to identify concepts which would significantly reduce the overall life cycle cost of these boosters. In a broad review of alternative systems, centrifugal compressors were found to be most promising. Electrically driven centrifugals, either direct drive or gear driven, were found to be quite limited in speed. Therefore they require many stages for these applications, and no cost advantage was indicated. Considerable promise was indicated for centrifugals driven by bleed air from the engine compressor, using turbocompressor units which are conversions of existing turbochargers for internal combustion engines. A first cost advantage of 30 to 80 percent was indicated for applications with an annual market size of at least ten units. Considerable savings in installation and maintenance costs are expected in addition.

  1. Monte Carlo modeling and analyses of YALINA-booster subcritical assembly part 1: analytical models and main neutronics parameters.

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, A.; Gohar, M. Y. A.; Nuclear Engineering Division

    2008-09-11

    This study was carried out to model and analyze the YALINA-Booster facility, of the Joint Institute for Power and Nuclear Research of Belarus, with the long term objective of advancing the utilization of accelerator driven systems for the incineration of nuclear waste. The YALINA-Booster facility is a subcritical assembly, driven by an external neutron source, which has been constructed to study the neutron physics and to develop and refine methodologies to control the operation of accelerator driven systems. The external neutron source consists of Californium-252 spontaneous fission neutrons, 2.45 MeV neutrons from Deuterium-Deuterium reactions, or 14.1 MeV neutrons from Deuterium-Tritium reactions. In the latter two cases a deuteron beam is used to generate the neutrons. This study is a part of the collaborative activity between Argonne National Laboratory (ANL) of USA and the Joint Institute for Power and Nuclear Research of Belarus. In addition, the International Atomic Energy Agency (IAEA) has a coordinated research project benchmarking and comparing the results of different numerical codes with the experimental data available from the YALINA-Booster facility and ANL has a leading role coordinating the IAEA activity. The YALINA-Booster facility has been modeled according to the benchmark specifications defined for the IAEA activity without any geometrical homogenization using the Monte Carlo codes MONK and MCNP/MCNPX/MCB. The MONK model perfectly matches the MCNP one. The computational analyses have been extended through the MCB code, which is an extension of the MCNP code with burnup capability because of its additional feature for analyzing source driven multiplying assemblies. The main neutronics parameters of the YALINA-Booster facility were calculated using these computer codes with different nuclear data libraries based on ENDF/B-VI-0, -6, JEF-2.2, and JEF-3.1.

  2. Impact Analysis for Fuel Assemblies in Spent Fuel Storage Rack

    International Nuclear Information System (INIS)

    The design and structural integrity evaluation of a spent fuel storage rack (SFSR) utilized for storing and protecting the spent fuel assemblies generated during the operation of a reactor are very important in terms of nuclear safety and waste management. The objective of this study is to show the validity of the SFSR design as well as fuel assembly through a structural integrity evaluation based on a numerical analysis. In particular, a dynamic time history analysis considering the gaps between the fuel assemblies and the walls of the storage cell pipes in the SFSR was performed to check the structural integrity of the fuel assembly and storage cell pipe

  3. Nuclear fuel assembly identification using computer vision

    International Nuclear Information System (INIS)

    This report describes an improved method of remotely identifying irradiated nuclear fuel assemblies. The method uses existing in-cell TV cameras to input an image of the notch-coded top of the fuel assemblies into a computer vision system, which then produces the identifying number for that assembly. This system replaces systems that use either a mechanical mechanism to feel the notches or use human operators to locate notches visually. The system was developed for identifying fuel assemblies from the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor, but could be used for other reactor assembly identification, as appropriate

  4. Fuel fire tests of selected assemblies

    Science.gov (United States)

    Kydd, G.; Spindola, K.; Askew, G. K.

    1982-04-01

    A varing assortment of clothing assemblies was tested in the Fuel Fire Test Facility at the Naval Air Development Center. Included was a Nomex-Kevlar Cloque Coverall which had relatively good protection from fuel flames.

  5. Evaluation of reactivity monitoring techniques in experiments with pulsed neutron source in the Yalina-Booster subcritical assembly; Evaluacion de tecnicas de monitorizacion de la reactividad en experimentos con fuente de neutrones pulsada en el conjunto subcritico Yalina-Booster

    Energy Technology Data Exchange (ETDEWEB)

    Becares, V.; Villamarin, D.; Fernandez-Ordonez, M.; Gonzalez-Romero, E. M.

    2010-07-01

    As a part of EUROTRANS program, it has carried out an experimental campaign focused in the validation of reactivity monitoring techniques in the Yalina-Booster subcritical assembly. The aim of this paper is to present the analysis of part of the experiments results, in particular those carried out with a pulsed neutron source.

  6. Thermal Analysis of a TREAT Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Papadias, Dionissios [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, Arthur E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-07-09

    The objective of this study was to explore options as to reduce peak cladding temperatures despite an increase in peak fuel temperatures. A 3D thermal-hydraulic model for a single TREAT fuel assembly was benchmarked to reproduce results obtained with previous thermal models developed for a TREAT HEU fuel assembly. In exercising this model, and variants thereof depending on the scope of analysis, various options were explored to reduce the peak cladding temperatures.

  7. Modular fuel-cell stack assembly

    Science.gov (United States)

    Patel, Pinakin

    2010-07-13

    A fuel cell assembly having a plurality of fuel cells arranged in a stack. An end plate assembly abuts the fuel cell at an end of said stack. The end plate assembly has an inlet area adapted to receive an exhaust gas from the stack, an outlet area and a passage connecting the inlet area and outlet area and adapted to carry the exhaust gas received at the inlet area from the inlet area to the outlet area. A further end plate assembly abuts the fuel cell at a further opposing end of the stack. The further end plate assembly has a further inlet area adapted to receive a further exhaust gas from the stack, a further outlet area and a further passage connecting the further inlet area and further outlet area and adapted to carry the further exhaust gas received at the further inlet area from the further inlet area to the further outlet area.

  8. Seismic response of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Hlaváč Z.

    2014-06-01

    Full Text Available The paper deals with mathematical modelling and computer simulation of the seismic response of fuel assembly components. The seismic response is investigated by numerical integration method in time domain. The seismic excitation is given by two horizontal and one vertical synthetic accelerograms at the level of the pressure vessel seating. Dynamic response of the hexagonal type nuclear fuel assembly is caused by spatial motion of the support plates in the reactor core investigated on the reactor global model. The modal synthesis method with condensation is used for calculation of the fuel assembly component displacements and speeds on the level of the spacer grid cells.

  9. Establishment of China Nuclear Fuel Assembly Database

    Institute of Scientific and Technical Information of China (English)

    CHENPeng; ZHANGYing-chao; LIUTing-jin; JINYong-li

    2003-01-01

    During researching, designing, manufacturing and post irradiation, a large amount of data on fuel assembly of China nuclear power plants has been accumulated. It is necessary to collect the data together,so that the researchers, designers, manufactures and managers could use the data conveniently. It was proposed to establish a China Nuclear Fuel Assembly Database through the Internet on workstations during the year of 2003 to 2006, so the data would be shared in China nuclear industry.

  10. Nuclear reactor fuel assembly with fuel rod removal means

    International Nuclear Information System (INIS)

    A fuel assembly is described for a nuclear reactor. The assembly has a bottom nozzle, at least one longitudinally extending control rod guide thimble attached to and projecting upwardly from the bottom nozzle and transverse grids spaced along the thimble. An organized array of elongated fuel rods are transversely spaced and supported by the grids and axially captured between the bottom nozzle and a top nozzle. The assembly comprises: (a) a transversely extending adapter plate formed by an arrangement of integral cross-laced ligaments defining a plurality of coolant flow openings; (b) means for mounting the adapter plate on an upper end portion of the thimble and spaced axially above and disposed transversely over the upper ends of all of the fuel rods present in the fuel assembly such that ones of the ligaments overlie corresponding ones of the fuel rods so as to prevent the fuel rods from moving upwardly through the coolant flow openings; and (c) removable plug means confined within the adapter plate and positioned over and spaced axially above selected ones of the fuel rods in providing access to at least one fuel rod for removal thereof upwardly through the axially spaced adapter plate without removing the top nozzle from the fuel assembly

  11. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    The fuel assembly of a boiling water reactor contains a number of vertical fuel rods with their lower ends against a bottom tie plate. The rods are positioned by spacers, which are fixed to the canning. The upward motion is reduced by the top plate of a special design. (G.B.)

  12. Apparatus and method for assembling fuel elements

    International Nuclear Information System (INIS)

    A nuclear fuel element assembling method and apparatus is preferably operable under programmed control unit to receive fuel rods from storage, arrange them into axially aligned stacks of closely monitored length, and transfer the stacks of fuel rods to a loading device for insertion into longitudinal passages in the fuel elements. In order to handle large numbers of one or more classifications of fuel rods or other cylindrical parts, the assembling apparatus includes at least two feed troughs each formed by a pair of screw members with a movable table having a plurality of stacking troughs for alignment with the feed troughs and with a conveyor for delivering the stacks to the loading device, the fuel rods being moved along the stacking troughs upon a fluid cushion. 23 claims, 6 figures

  13. SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

    Science.gov (United States)

    Schluderberg, D.C.; Ryon, J.W.

    1962-05-01

    A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

  14. Appearance detection device for fuel assembly

    International Nuclear Information System (INIS)

    The present invention concerns a device for detecting the appearance of fuel assemblies for a power plant, in which the device photographs corners of fuel assemblies by a TV-camera to perform detection with higher reliability. Namely, heretofore, fuel assembly to substantially square pillar shape for a BWR and a PWR has been rotated and one or two faces have been detected from the front by the TV-camera. In the present invention, a TV-camera used exclusively for corners is additionally disposed on or near the diagonal line of the corners. With such a constitution, corners of the fuel assemblies can be photographed simultaneously with the conventional appearance test. As a result, since appearance test for the front and the corners can be conducted at the same time, extremely effective detection can be conducted in terms of detection of a rupture of grids and prevention of dead angle. The corners of assemblies which tend to undergo damages upon charge/discharge of fuels can be detected carefully. Accordingly, a highly reliable detection can be conducted. (I.S.)

  15. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    In the fuel assembly of the present invention, a means for mounting and securing short fuel rods is improved. Not only long fuel rods but also short fuel rods are disposed in channel of the fuel assembly to improve reactor safety. The short fuel rods are supported by a screw means only at the lower end plug. The present invention prevents the support for the short fuel rod from being unreliable due to the slack of the screw by the pressure of inflowing coolants. That is, coolant abutting portions such as protrusions or concave grooves are disposed at a portion in the channel box where coolants flowing from the lower tie plate, as an uprising stream, cause collision. With such a constitution, a component caused by the pressure of the flowing coolants is formed. The component acts as a rotational moment in the direction of screwing the male threads of the short fuel rod into the end plug screw hole. Accordingly, the screw is not slackened, and the short fuel rods are mounted and secured certainly. (I.S.)

  16. Nuclear fuel assembly identification using computer vision

    International Nuclear Information System (INIS)

    A new method of identifying fuel assemblies has been developed. The method uses existing in-cell TV cameras to read the notch-coded handling sockets of Fast Flux Test Facility (FFTF) assemblies. A computer looks at the TV image, locates the notches, decodes the notch pattern, and produces the identification number. A TV camera is the only in-cell equipment required, thus avoiding complex mechanisms in the hot cell. Assemblies can be identified in any location where the handling socket is visible from the camera. Other advantages include low cost, rapid identification, low maintenance, and ease of use

  17. RBMK fuel assemblies: Current status and perspectives

    International Nuclear Information System (INIS)

    The safety enhancement measures implemented since 1986 have led to substantial burnup reduction in the RBMK fuel assemblies and consequently to economical losses. With the purpose to compensate the losses, computer analysis and experiments were performed during the last decade. The works were aimed at the RBMK fuel charge perfection to reduce void reactivity effect and to increase fuel burnup. The paper presents principle results of the studies which are currently under implementation or are supposed to be implemented in the nearest future. (author)

  18. Calibration of spent fuel measurement assembly

    International Nuclear Information System (INIS)

    The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110mAg isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110mAg isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system. - Highlights: • Calibration of research reactor spent fuel measurement assembly. • On-site prepared 110mAg isotope used for the measurement. • Calculated self-shielding factor for the IRT-2M fuel. • Applicable to other research reactor fuel geometries

  19. Optimization of fuel rod enrichment distribution for BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-09-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. the combinatorial optimization problem of grouping fuel rods into a given number of rod groups with the same enrichment, and the problem of determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by the linear combination C{sub 1}X + C{sub 2}X{sub G}, where X and X{sub G} stand, respectively, for control variables giving constraint to the local power peaking factor and the gadolinium rod power. C{sub 1} and C{sub 2} are user-definable weighting factors to accommodate design preferences. The algorithm for solving this combinatorial optimization problem starts by finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering. This latter problem is solved using the method of approximation programming. A practical application is shown for a contemporary 8 x 8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  20. FUEL ASSEMBLY SHAKER AND TRUCK TEST SIMULATION

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Jensen, Philip J.; Sanborn, Scott E.; Hanson, Brady D.

    2014-09-25

    This study continues the modeling support of the SNL shaker table task from 2013 and includes analysis of the SNL 2014 truck test campaign. Detailed finite element models of the fuel assembly surrogate used by SNL during testing form the basis of the modeling effort. Additional analysis was performed to characterize and filter the accelerometer data collected during the SNL testing. The detailed fuel assembly finite element model was modified to improve the performance and accuracy of the original surrogate fuel assembly model in an attempt to achieve a closer agreement with the low strains measured during testing. The revised model was used to recalculate the shaker table load response from the 2013 test campaign. As it happened, the results remained comparable to the values calculated with the original fuel assembly model. From this it is concluded that the original model was suitable for the task and the improvements to the model were not able to bring the calculated strain values down to the extremely low level recorded during testing. The model needs more precision to calculate strains that are so close to zero. The truck test load case had an even lower magnitude than the shaker table case. Strain gage data from the test was compared directly to locations on the model. Truck test strains were lower than the shaker table case, but the model achieved a better relative agreement of 100-200 microstrains (or 0.0001-0.0002 mm/mm). The truck test data included a number of accelerometers at various locations on the truck bed, surrogate basket, and surrogate fuel assembly. This set of accelerometers allowed an evaluation of the dynamics of the conveyance system used in testing. It was discovered that the dynamic load transference through the conveyance has a strong frequency-range dependency. This suggests that different conveyance configurations could behave differently and transmit different magnitudes of loads to the fuel even when travelling down the same road at

  1. Method for the detection of defective nuclear fuel assemblies

    International Nuclear Information System (INIS)

    There is applied an ultrasonic transmitter on a tape carrier by means of which the ultrasonic transmitter can be guided underwater between the fuel assemblies. If a fuel assembly is defective, i.e. filled with water, the reflection coefficient at the front interface between cladding and inner space of the fuel assembly will decrease. Essential parts of the ultrasonic signal will move through the liquid and will not be reflected until the backward liquid/metal interface of the fuel assembly. This impulse echo is different from that of the gas-filled fuel assembly. (DG)

  2. Advanced membrane electrode assemblies for fuel cells

    Science.gov (United States)

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  3. Spent fuel assembly source term parameters

    International Nuclear Information System (INIS)

    Containment of cask contents by a transport cask is a function of the cask body, one or more closure lids, and various bolting hardware, and seals associated with the cavity closure and other containment penetrations. In addition, characteristics of cask contents that impede the ability of radionuclides to move from an origin to the external environment also provide containment. In essence, multiple release barriers exist in series in transport casks, and the magnitude of the releasable activity in the cask is considerably lower than the total activity of its contents. A source term approach accounts for the magnitude of the releasable activity available in the cask by assessing the degree of barrier resistance to release provided by material characteristics and inherent barriers that impede the release of radioactive contents. Standardized methodologies for defining the spent-fuel transport packages with specified regulations have recently been developed. An essential part of applying the source term methodology involves characterizing the response of the spent fuel under regulatory conditions of transport. Thermal and structural models of the cask and fuel are analyzed and used to predict fuel rod failure probabilities. Input to these analyses and failure evaluations cover a wide range of geometrical and material properties. An important issue in the development of these models is the sensitivity of the radioactive source term generated during transport to individual parameters such as temperature and fluence level. This paper provides a summary of sensitivity analyses concentrating on the structural response and failure predictions of the spent fuel assemblies

  4. Interface ring for gas turbine fuel nozzle assemblies

    Science.gov (United States)

    Fox, Timothy A.; Schilp, Reinhard

    2016-03-22

    A gas turbine combustor assembly including a combustor liner and a plurality of fuel nozzle assemblies arranged in an annular array extending within the combustor liner. The fuel nozzle assemblies each include fuel nozzle body integral with a swirler assembly, and the swirler assemblies each include a bellmouth structure to turn air radially inwardly for passage into the swirler assemblies. A radially outer removed portion of each of the bellmouth structures defines a periphery diameter spaced from an inner surface of the combustor liner, and an interface ring is provided extending between the combustor liner and the removed portions of the bellmouth structures at the periphery diameter.

  5. Fuel assembly with a flute for water distribution

    International Nuclear Information System (INIS)

    The fuel assembly is arranged so that groups of fuel rods are enclosed into walls. The top end of the assembly has a peripherical distribution channel which recieves water for emergency cooling and distributes it evenly over the fuel rods. (G.B.)

  6. Optical matrix for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    In order to detect the presence of fuel rods, it was selected a reflection optical transducer, which provides a measurable electrical signal when the beam at a certain distance is interrupted then there is a reflection causing a excitation to the sensor that provides a change of state at the output of transducer. This step is coupled through an operational amplifier which drives the opto coupler circuit isolating this step of the interface and a personal computer. This work presents the description of components, designs, signal coupler and opto isolater circuit, interface circuit and tutorial assemble program. (Author)

  7. Central position detection method for fuel assembly and device therefor

    International Nuclear Information System (INIS)

    The present invention provides a method for detecting a central position of a fuel assembly by an image processing technique without influenced by a deviation of the central position of the fuel assembly depending on the accuracy for the stoppage of an underwater vehicle and rattling of fuels in a fuel basket. Namely, a characteristic amount comparing method and a linear detecting method are utilized by image processing techniques. Images are taken by a camera disposed at a predetermined position, and common characteristically shaped portions of each of the top portions of fuel assemblies are detected based on the photographed images. The central position at the top of the fuel assembly is detected based on the characteristic. In a case of a BWR fuel assembly, a channel fastener screw portion and a handle at the top of the fuel constitute the characteristic portions. The longitudinal component of the handle is detected by the linear method, and the aperture like circular portion of the channel fastener screw portion is detected by the characteristic amount comparing method. In a case of a PWR type fuel assembly, two positioning pin holes at a fuel top corner portion are detected using the characteristic amount comparing method. The central position of the fuel assembly is detected based on each of the results. (I.S.)

  8. Zirconium fuel cladding corrosion prediction in fuel assembly operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kritsky, V.G.; Berezina, I.G., E-mail: kritsky@givnipiet.spb.ru, E-mail: alemaskina@givnipiet.ru [Leading Inst. ' VNIPIET' , Saint Petersburg (Russian Federation)

    2010-07-01

    At present, the work to extend fuel cycles is carried out at NPP with VVER reactors. With the increase of fuel assembly burn-up to 70-100 MWd/kg U and linear power, the local coolant «nucleate boiling» is inevitable which in combination with coolant «acidification» alongside with the existing water chemistry norms will increase zirconium alloy corrosion. The rate of Zr alloy corrosion under reactor irradiation depends on temperature and heat flux through fuel cladding, coolant chemistry (concentrations of H{sub 2}O{sub 2}, OH{sup -}, O{sub 2}, hydrogen, ammonia, strong alkalis - LiOH, KOH, pH, ets.), steam content, alloy composition and some other parameters. A generalized model for calculating Zr alloys corrosion, which take into account the above-mentioned factors, was developed: K = k{sub 1}e {sup -}ΣvQ{sub 1}/R(T+ΔT) + k{sub 2} 1/1 - α + β Φ{sup n} where K{sub 1}, K{sub 2} are the coefficients depending on the water chemistry conditions and composition of Zr alloys; α is the value of steam content; Φ is a neutron flux; n is the coefficient depending on the fuel assembly type; β is the coefficient considering the impact of impurities suppressing the radiolysis, Q{sub 1} is energy contributions of alloying components and water impurities to oxide formation, v{sub i} - stehiometry coefficient. This model allows to predict a fuel cladding corrosion taking into account the alloys composition, water chemistry and fuel burn-up. The model was verified with the help of autoclave and reactor tests for commercial and modified Zr alloys. The activation energy of oxidation process is calculating on the base of ideal mixed oxide formation model. The success of such approach makes possible to propose a generalized model for calculating the corrosion of different Zr alloys in all types of water chemistry environments of old and new LWRs. (author)

  9. Separator assembly for use in spent nuclear fuel shipping cask

    Science.gov (United States)

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  10. Precharacterization Report for Instrumented Fuel Assembly (IFA)-527

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, M. E.; Bradley, E. R.; Daniel, J. L.; Davis, N. C.; Lanning, D. D.; Williford, R. E.

    1981-07-01

    This report is a resource document covering the rationale, design, fabrication, and preirradiation characterization of instrumented fuel assembly (IFA)-527. This assembly is being irradiated in the Halden Boiling Water Reactor (HBWR) in Norway as part of the Experimental Support and Development of Single-Rod Fuel Codes Program conducted by Pacific Northwest laboratory (PNL) and sponsored by the Fuel Behavior Research Branch of the U.S. Nuclear Regulatory Commission (NRC). Data from this assembly will be used to better understand light water reactor (LWR) fuel behavior under normal operating conditions.

  11. Dynamic modelling of PWR fuel assembly for seismic behaviour

    International Nuclear Information System (INIS)

    Vibration and snap back tests have shown that the behaviour of PWR fuel assemblies was non linear : the fuel assembly eigenfrequencies decrease with the excitation level or with the motion amplitude, which was supposed to be due to the slippage of the fuel rods through the grids. Up to now the fuel assembly models were linear and composed by one beam alone representing both the guide thimbles and the fuel rods or by two beams (one for the guide thimbles and one for the fuel rods). The stiffness of such models' were adjusted to fit with the measured eigenfrequency corresponding to a given amplitude. The aim of this paper is to identify the influence of the slippage between grids and fuel rods on the dynamic behaviour of the fuel assembly. For that purpose a non linear fuel assembly model is proposed representing explicitly the slippage phenomenon and is applied to the reduced scale fuel assemblies which have been tested in the framework of a collaboration between FRAMATOME and CEA-DMT. Comparisons between calculations and experiments will be presented and the limitation of this model will be also discussed

  12. Modelling and modal properties of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2011-12-01

    Full Text Available The paper deals with the modelling and modal analysis of the hexagonal type nuclear fuel assembly. This very complicated mechanical system is created from the many beam type components shaped into spacer grids. The cyclic and central symmetry of the fuel rod package and load-bearing skeleton is advantageous for the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and skeleton linked by several spacer grids in horizontal planes. The derived mathematical model is used for the modal analysis of the Russian TVSA-T fuel assembly and validated in terms of experimentally determined natural frequencies, modes and static deformations caused by lateral force and torsional couple of forces. The presented model is the first necessary step for modelling of the nuclear fuel assembly vibration caused by different sources of excitation during the nuclear reactor VVER type operation.

  13. Method of handling and/or storing a nuclear fuel assembly consisting of an elongated frame that contains fuel rods and fuel assembly designed specially for this method

    International Nuclear Information System (INIS)

    In order to assure subcriticality during handling, transport and/ore storage of nuclear reactor fuel assemblies and additional body containing a neutron absorbing material and touching beside the fuel rods is fixed to the frame of the fuel assembly. This body has a handle with an adapted coupling element mounted on a holding device for handling and/or storage of the fuel assembly. (orig./RW)

  14. Dynamic response of nuclear fuel assembly excited by pressure pulsations

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2012-12-01

    Full Text Available The paper deals with dynamic load calculation of the hexagonal type nuclear fuel assembly caused by spatial motion of the support plates in the reactor core. The support plate motion is excited by pressure pulsations generated by main circulation pumps in the coolant loops of the primary circuit of the nuclear power plant. Slightly different pumps revolutions generate the beat vibrations which causes an amplification of fuel assembly component dynamic deformations and fuel rods coating abrasion. The cyclic and central symmetry of the fuel assembly makes it possible the system decomposition into six identical revolved fuel rod segments which are linked with central tube and skeleton by several spacer grids in horizontal planes.The modal synthesis method with condensation of the fuel rod segments is used for calculation of the normal and friction forces transmitted between fuel rods and spacer grids cells.

  15. The influence of fuel assembly characteristics on reactor safety

    International Nuclear Information System (INIS)

    To improve fuel utilization and nuclear plant economy, most nuclear plants of China adopt increased fuel enrichment and long cycle analysis. Core power distribution will be worse with these advanced items. Radial and axial peak increase too. This is a challenge to reactor safety. Since the fuel assembly is the most important part of a reactor core, fuel assembly characteristics affect reactor safety a lot. A few aspects of influence on reactor safety are discussed in this paper as a reference for fuel assembly design. A better fuel assembly design can increase heat exchange ability, especially in cold wall cells. The grids nearby core outlet can efficiently mix the flow of hot channel and average channel to decrease DNBR. In safety analysis, we always suppose the center of center assembly is the hot channel, but sometimes based on actual power distribution the hot channel occurs at side cell or corner cell. So the distribution of grids pressure drop coefficients can affect the minimum DNBR. A better fuel assembly design can help to spread core power distribution, decrease radial and axial peak efficiently. To spread core power distribution, different neutronic poisons are added into fuel pellet by different ways, and then the relative effects on reactor safety are different. At the same time, better fuel assembly design should leave enough margins for reactor safety to handle high burnup condition and so on. Fuel pellet and clad capabilities are getting worse versus increasing fuel burnup. This is a challenge to reactor safety, so more attentions should be paid to fuel burnup characteristics. (author)

  16. Optical fiber scope for inspecting fuel assembly

    International Nuclear Information System (INIS)

    Since a fiber scope has only one objective section, it has to observe a plurality of places successively. Then, if the time for the observation is long, the objective section is deteriorated by radiation rays, which causes a problem of interrupting the observation and increasing operator's radiation dose. In view of the above, one or two light guides are combined with an image guide to form one objective section, and a plurality of them are formed in parallel and gathered as a comb-like shape. A prism is put into a window of the objective section and resins are filled or a glass cover is attached, to make the objective section smooth and flat. Compared with the case of using only one objective section, it is no more necessary for successive observation, and objection can be conducted at one time. For example, if a fiber scope having nine objective sections is used for observing 8 x 8 arrangement fuel assembly, the observation time is shortened to 1/9. Since the prism, the glass cover, and the resins are used for making the window flat, cruds deposited between the optical fiber and a reflection mirror are easily removed, to obtain clear images. (N.H.)

  17. Effects of radial void distribution within fuel assembly on assembly neutronic characteristics

    International Nuclear Information System (INIS)

    The effect of radial subchannel-wise void distribution in a fuel assembly on assembly neutronic characteristics has been investigated using the assembly calculation code SRAC95 and the subchannel analysis code THERMIT2. With the iterative calculation of assembly calculation and the subchannel analysis (Method 1), subchannel-wise void fraction distribution, pin-power distribution and the infinite multiplication factor of the assembly are calculated. The results are compared with the result of the assembly calculation using uniform void distribution as input (Method 2). The calculation is performed for two assembly configurations in the present study: one is a fuel assembly that does not include a water rod (Case 1) and the other is the assembly that includes a water rod (Case 2). The differences in the infinite multiplication factor and pin-power peaking factor between the two methods are small in both cases. In typical BWR fuel assemblies that are investigated in the present study, the method that does not consider the radial subchannel-wise void fraction distribution within a fuel assembly (Method 2) is accurate enough for practical applications. (author)

  18. GRYPHON: Air launched space booster

    Science.gov (United States)

    1993-06-01

    The project chosen for the winter semester Aero 483 class was the design of a next generation Air Launched Space Booster. Based on Orbital Sciences Corporation's Pegasus concept, the goal of Aero 483 was to design a 500,000 pound air launched space booster capable of delivering 17,000 pounds of payload to Low Earth Orbit and 8,000 pounds of payload to Geosynchronous Earth Orbit. The resulting launch vehicle was named the Gryphon. The class of forty senior aerospace engineering students was broken down into eight interdependent groups. Each group was assigned a subsystem or responsibility which then became their field of specialization. Spacecraft Integration was responsible for ensuring compatibility between subsystems. This group kept up to date on subsystem redesigns and informed those parties affected by the changes, monitored the vehicle's overall weight and dimensions, and calculated the mass properties of the booster. This group also performed the cost/profitability analysis of the Gryphon and obtained cost data for competing launch systems. The Mission Analysis Group was assigned the task of determining proper orbits, calculating the vehicle's flight trajectory for those orbits, and determining the aerodynamic characteristics of the vehicle. The Propulsion Group chose the engines that were best suited to the mission. This group also set the staging configurations for those engines and designed the tanks and fuel feed system. The commercial satellite market, dimensions and weights of typical satellites, and method of deploying satellites was determined by the Payloads Group. In addition, Payloads identified possible resupply packages for Space Station Freedom and identified those packages that were compatible with the Gryphon. The guidance, navigation, and control subsystems were designed by the Mission Control Group. This group identified required tracking hardware, communications hardware telemetry systems, and ground sites for the location of the Gryphon

  19. The Booster

    CERN Multimedia

    1972-01-01

    Where the beams from the Booster's four rings begin to recombine, before transfer to the PS. On the left are dipoles for vertical steering, and on the right is the tank containing two septum magnets which form the first combining element.

  20. The development of flow test technology for PWR fuel assembly

    International Nuclear Information System (INIS)

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

  1. Temperature measuring analysis of the nuclear reactor fuel assembly

    Science.gov (United States)

    F., Urban; Ľ., Kučák; Bereznai, J.; Závodný, Z.; Muškát, P.

    2014-08-01

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  2. Parallel processing of neutron transport in fuel assembly calculation

    International Nuclear Information System (INIS)

    Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's

  3. Immunity booster

    International Nuclear Information System (INIS)

    The immunity booster is, according to its patent description, microbiologically pure water with an D/(D+H) isotopic concentration of 100 ppm, with physical-chemical characteristics similar to those of distilled water. It is obtained by sterilization of a mixture of deuterium depleted water, with a 25 ppm isotopic concentration, with distilled water in a volume ratio of 4:6. Unlike natural immunity boosters (bacterial agents as Bacillus Chalmette-Guerin, Corynebacterium parvum; lipopolysaccharides; human immunoglobulin) or synthetical products (levamysol; isoprinosyne with immunostimulating action), which cause hypersensitivity and shocks, thrill, fever, sickness and the immunity complex disease, the water of 100 ppm D/(D + H) isotopic concentration is a toxicity free product. The testing for immune reaction of the immunity booster led to the following results: - an increase of cell action capacity in the first immunity shielding stage (macrophages), as evidenced by stimulation of a number of essential characterizing parameters, as well as of the phagocytosis capacity, bactericide capacity, and opsonic capacity of serum; - an increase of the number of leucocyte particularly of the granulocyte in peripheral blood, produced especially when medullar toxic agents like caryolysine are used; - it hinders the effect of lowering the number of erythrocytes in peripheral blood produced by experimentally induced chronic inflammation; - an increase of nonspecific immunity defence capacity against specific bacterial aggression of both Gram-positive bacteria (Streptococcus pneumoniae558) and of the Gram-negative ones (Klebsiella pneumoniae 507); - an increase of immunity - stimulating activity (proinflamatory), like that of levamisole as evidenced by the test of stimulation of experimentally induced inflammation by means of carrageenan. The following advantages of the immunity booster are stressed: - it is toxicity free and side effect free; - can be orally administrated as food

  4. ABB. CASE's GUARDIANTM Debris Resistant Fuel Assembly Design

    International Nuclear Information System (INIS)

    ABB CE's experience, that 72% of all recent fuel-rod failures are caused by debris fretting, is typical. In response to this problem, ABB Combustion Engineering began supplying in the late 1980s fuel assemblies with a variety of debris resistant features, including both long-end caps and small flow holes. Now ABB CAE has developed an advanced debris resistant design concept, GUARDIANTM, which has the advantage of capturing and retaining more debris than other designs, while displacing less plenum or active fuel volume than the long end-cap design. GUARDIANTM design features have now been implemented into four different assembly designs. ABB CASE's GUARDIANTM fuel assembly is an advanced debris-resistant design which has both superior filtering performance and uniquely, excellent debris retention, Retention effectively removes the debris from circulation in the coolant so that it is not able to threaten the fuel again. GUARDIANTM features have been incorporated into four ABB. CAE fuel assembly designs. These assemblies are all fully compatible with the NSLS, and full-batch operation with GUARDIANTM began in 1992. The number of plants of both CAE and non-CAE design which accept GUARDIANTM for debris protection is expected to grow significantly during the next few years

  5. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  6. Review of qualifications for fuel assembly fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Slabu, Dan; Zemek, Martin; Hellwig, Christian [Axpo AG, Baden (Switzerland)

    2013-02-15

    The required quality of nuclear fuel in industrial production can only be assured by applying processes in fabrication and inspection, which are well mastered and have been proven by an appropriate qualification. The present contribution shows the understanding and experiences of Axpo with respect to qualifications in the frame of nuclear fuel manufacturing and reflects some related expectations of the operator. (orig.)

  7. Thermomechanical evaluation of the fuel assemblies fabricated in the ININ

    International Nuclear Information System (INIS)

    The pilot plant of fuel production of the National Institute of Nuclear Research (ININ) provided to the Laguna Verde Nuclear Power Plant (CNLV) four fuel assemblies type GE9B. The fuel irradiation was carried out in the unit 1 of the CNLV during four operation cycles, highlighting the fact that in their third cycle the four assemblies were placed in the center of the reactor core. In the Nuclear Systems Department (DSN) of the ININ it has been carried out studies to evaluate their neutron performance and to be able to determine the exposure levels of this fuels. Its also outlines the necessity to carry out a study of the thermomechanical behavior of the fuel rods that compose the assemblies, through computational codes that simulate their performance so much thermal as mechanical. For such purpose has been developing in the DSN the FETMA code, together with the codes that compose the system Fuel Management System (FMS), which evaluates the thermomechanical performance of fuel elements. In this work were used the FETMA and FEMAXI codes (developed by JAERI) to study the thermomechanical performance of the fuel elements manufactured in the ININ. (Author)

  8. Quantity Distance for the Kennedy Space Center Vehicle Assembly Building for Solid Propellant Fueled Launchers

    Science.gov (United States)

    Stover, Steven; Diebler, Corey; Frazier, Wayne

    2006-01-01

    The NASA KSC VAB was built to process Apollo launchers in the 1960's, and later adapted to process Space Shuttles. The VAB has served as a place to assemble solid rocket motors (5RM) and mate them to the vehicle's external fuel tank and Orbiter before rollout to the launch pad. As Space Shuttle is phased out, and new launchers are developed, the VAB may again be adapted to process these new launchers. Current launch vehicle designs call for continued and perhaps increased use of SRM segments; hence, the safe separation distances are in the process of being re-calculated. Cognizant NASA personnel and the solid rocket contractor have revisited the above VAB QD considerations and suggest that it may be revised to allow a greater number of motor segments within the VAB. This revision assumes that an inadvertent ignition of one SRM stack in its High Bay need not cause immediate and complete involvement of boosters that are part of a vehicle in adjacent High Bay. To support this assumption, NASA and contractor personnel proposed a strawman test approach for obtaining subscale data that may be used to develop phenomenological insight and to develop confidence in an analysis model for later use on full-scale situations. A team of subject matter experts in safety and siting of propellants and explosives were assembled to review the subscale test approach and provide options to NASA. Upon deliberations regarding the various options, the team arrived at some preliminary recommendations for NASA.

  9. Hydrogen storage and integrated fuel cell assembly

    Science.gov (United States)

    Gross, Karl J.

    2010-08-24

    Hydrogen is stored in materials that absorb and desorb hydrogen with temperature dependent rates. A housing is provided that allows for the storage of one or more types of hydrogen-storage materials in close thermal proximity to a fuel cell stack. This arrangement, which includes alternating fuel cell stack and hydrogen-storage units, allows for close thermal matching of the hydrogen storage material and the fuel cell stack. Also, the present invention allows for tailoring of the hydrogen delivery by mixing different materials in one unit. Thermal insulation alternatively allows for a highly efficient unit. Individual power modules including one fuel cell stack surrounded by a pair of hydrogen-storage units allows for distribution of power throughout a vehicle or other electric power consuming devices.

  10. Ram booster

    Science.gov (United States)

    Brand, Vance D. (Inventor); Morgan, Walter Ray (Inventor)

    2011-01-01

    The present invention is a space launch system and method to propel a payload bearing craft into earth orbit. The invention has two, or preferably, three stages. The upper stage has rocket engines capable of carrying a payload to orbit and provides the capability of releasably attaching to the lower, or preferably, middle stage. Similar to the lower stage, the middle stage is a reusable booster stage that employs all air breathing engines, is recoverable, and can be turned-around in a short time between missions.

  11. Pressure drop evaluation in fuel assembly bottom nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Sydney da Silva; Brittes, Luiz Henrique A. [Industrias Nucleares do Brasil S.A. (INB), Resende, RJ (Brazil)]. E-mails: sydney@inb.gov.br; brittes@inb.gov.br; Navarro, Moyses A. [Centro de Desenvolvimento de Energia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mail: navarro@cdtn.br

    2007-07-01

    Experiments were conducted in order to assess the pressure drop through an anti-debris bottom nozzle relatively to a standard bottom nozzle of a nuclear fuel assembly. Two kinds of experiments have been performed: one using bottom nozzles connected with the lower part of a Fuel Assembly (containing two spacer grids) and another one using the bottom nozzle alone. Reynolds numbers ranging from 10500 . 95000 have been employed, temperatures ranging from 40 . 55 deg C and pressures up to 4 bar. Results have shown that the pressure drop coefficients of the anti-debris nozzle referring to the whole lower region of the Fuel Assembly were {approx} 13% (for Re {approx_equal} 95000) till {approx} 17% (for Re {approx_equal} 10500) higher than the coefficients for standard bottom nozzle. This difference increases up to 118% when the pressure drop coefficients of the bottom nozzle alone are considered. (author)

  12. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.

    2010-06-17

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

  13. Measuring device for effective multiplication factors of a fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ueda, Makoto

    1988-11-14

    Purpose: To measure the effective multiplication factor of a fuel assembly without using an external neutron source. Constitution: Neutron absorbers disposed on the surface of a fuel assembly incorporating a spontaneous neutron source so as to put the surface of the assembly therebetween is moved. As the neutron absorber, a cadmium plate is most suitable, but boron, gadolinium or disprosium may also be used. Neutron counting rate phi upon setting the distance between the neutron absorbers and the surface of the fuel assembly to greater than 2 cm and neutron counting rate phi' upon setting it to less than 2 cm are measured by neutron detectors. The effective multiplication factor of the fuel assembly is calculated based on the results of both of the measurements according to the following equation: K = (A(phi/phi')-1)/(AB(phi/phi'-1)) According to this method, exchange for the external neutron source is no more required, and the maintenance is easier and working efficiency is higher as compared with the prior art. Further, since phi/phi' can be determined in one identical detector in a short period of time, the measuring error can be reduced. (Horiuchi, T.).

  14. A spent fuel assemblies monitoring device by nondestructive analysis 'PYTHON'

    International Nuclear Information System (INIS)

    The monitoring of spent fuel assemblies (16 x 16 UOX) in KWG-reactor pool with the use of non-destructive methods (total Gamma and neutron counting) allow the control of average burn-up and the extremity burn-up. The measurements allow a safety-criticality control before loading the fuel assemblies into the transport casks. A device called PYTHON has been tested and qualified in France. This paper presents a description of the industrial PYTHON device and the results of the measurements. (orig.)

  15. Nuclear imaging of the fuel assembly in ignition experiments

    International Nuclear Information System (INIS)

    First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium–tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate that the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models' prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%–25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface

  16. Nuclear imaging of the fuel assembly in ignition experiments

    Energy Technology Data Exchange (ETDEWEB)

    Grim, G. P.; Guler, N.; Merrill, F. E.; Morgan, G. L.; Danly, C. R.; Volegov, P. L.; Wilde, C. H.; Wilson, D. C.; Clark, D. S.; Hinkel, D. E.; Jones, O. S.; Raman, K. S.; Izumi, N.; Fittinghoff, D. N.; Drury, O. B.; Alger, E. T.; Arnold, P. A.; Ashabranner, R. C.; Atherton, L. J.; Barrios, M. A.; Batha, S.; Bell, P. M.; Benedetti, L. R.; Berger, R. L.; Bernstein, L. A.; Berzins, L. V.; Betti, R.; Bhandarkar, S. D.; Bionta, R. M.; Bleuel, D. L.; Boehly, T. R.; Bond, E. J.; Bowers, M. W.; Bradley, D. K.; Brunton, G. K.; Buckles, R. A.; Burkhart, S. C.; Burr, R. F.; Caggiano, J. A.; Callahan, D. A.; Casey, D. T.; Castro, C.; Celliers, P. M.; Cerjan, C. J.; Chandler, G. A.; Choate, C.; Cohen, S. J.; Collins, G. W.; Cooper, G. W.; Cox, J. R.; Cradick, J. R.; Datte, P. S.; Dewald, E. L.; Di Nicola, P.; Di Nicola, J. M.; Divol, L.; Dixit, S. N.; Dylla-Spears, R.; Dzenitis, E. G.; Eckart, M. J.; Eder, D. C.; Edgell, D. H.; Edwards, M. J.; Eggert, J. H.; Ehrlich, R. B.; Erbert, G. V.; Fair, J.; Farley, D. R.; Felker, B.; Fortner, R. J.; Frenje, J. A.; Frieders, G.; Friedrich, S.; Gatu-Johnson, M.; Gibson, C. R.; Giraldez, E.; Glebov, V. Y.; Glenn, S. M.; Glenzer, S. H.; Gururangan, G.; Haan, S. W.; Hahn, K. D.; Hammel, B. A.; Hamza, A. V.; Hartouni, E. P.; Hatarik, R.; Hatchett, S. P.; Haynam, C.; Hermann, M. R.; Herrmann, H. W.; Hicks, D. G.; Holder, J. P.; Holunga, D. M.; Horner, J. B.; Hsing, W. W.; Huang, H.; Jackson, M. C.; Jancaitis, K. S.; Kalantar, D. H.; Kauffman, R. L.; Kauffman, M. I.; Khan, S. F.; Kilkenny, J. D.; Kimbrough, J. R.; Kirkwood, R.; Kline, J. L.; Knauer, J. P.; Knittel, K. M.; Koch, J. A.; Kohut, T. R.; Kozioziemski, B. J.; Krauter, K.; Krauter, G. W.; Kritcher, A. L.; Kroll, J.; Kyrala, G. A.; Fortune, K. N. La; LaCaille, G.; Lagin, L. J.; Land, T. A.; Landen, O. L.; Larson, D. W.; Latray, D. A.; Leeper, R. J.; Lewis, T. L.; LePape, S.; Lindl, J. D.; Lowe-Webb, R. R.; Ma, T.; MacGowan, B. J.; MacKinnon, A. J.; MacPhee, A. G.; Malone, R. M.; Malsbury, T. N.; Mapoles, E.; Marshall, C. D.; Mathisen, D. G.; McKenty, P.; McNaney, J. M.; Meezan, N. B.; Michel, P.; Milovich, J. L.; Moody, J. D.; Moore, A. S.; Moran, M. J.; Moreno, K.; Moses, E. I.; Munro, D. H.; Nathan, B. R.; Nelson, A. J.; Nikroo, A.; Olson, R. E.; Orth, C.; Pak, A. E.; Palma, E. S.; Parham, T. G.; Patel, P. K.; Patterson, R. W.; Petrasso, R. D.; Prasad, R.; Ralph, J. E.; Regan, S. P.; Rinderknecht, H.; Robey, H. F.; Ross, G. F.; Ruiz, C. L.; Seguin, F. H.; Salmonson, J. D.; Sangster, T. C.; Sater, J. D.; Saunders, R. L.; Schneider, M. B.; Schneider, D. H.; Shaw, M. J.; Simanovskaia, N.; Spears, B. K.; Springer, P. T.; Stoeckl, C.; Stoeffl, W.; Suter, L. J.; Thomas, C. A.; Tommasini, R.; Town, R. P.; Traille, A. J.; Wonterghem, B. Van; Wallace, R. J.; Weaver, S.; Weber, S. V.; Wegner, P. J.; Whitman, P. K.; Widmann, K.; Widmayer, C. C.; Wood, R. D.; Young, B. K.; Zacharias, R. A.; Zylstra, A.

    2013-05-01

    First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium–tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate that the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models’ prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%–25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface.

  17. Advanced PWR fuel assembly development programs in Korea

    International Nuclear Information System (INIS)

    Both KNFC and Westinghouse have continued to focus on developing products that will meet the challenge of increasing fuel duty requirements in Korea. These higher duty conditions include higher energy core designs through improved plant capacity factors, power uprate, extended fuel burnup, peaking factor increases, and more severe coolant chemistry (including high lithium concentration). Recent advanced fuel development activities in Korea include implementation of the 17x17 Robust Fuel Assembly (RFA), which is currently in operation with excellent performance in the United States and Europe, as well as the 16x16 PLUS7TM fuel assembly for use in KSNP plants. KNFC and Westinghouse are jointly developing advanced fuel that will meet future fuel duty challenges of 17x17 and 16x16 Westinghouse type plants. This paper focuses on advanced fuel assembly development programs that are underway and how these designs demonstrate improved margins under high duty plant operating conditions. In designing for these high duty conditions key design considerations for the various operational modes (i.e. power uprating, high burnup, long cycles, etc.) must be identified. These design considerations will include the traditional factors such as safety margin (DNB and LOCA), fuel rod design margin (e.g. corrosion, internal pressure, etc.) and mechanical design margins, among others. In addressing these design considerations, the fundamental approach is to provide additional design margin through materials, mechanical, and thermal performance enhancements, to assure flawless fuel performance. The foundation of all fuel designs is the product development process used to meet the demands of modern high duty operation including power uprating, high burnup, longer cycles, and high-lithium coolant chemistries. These advanced fuel assembly designs incorporate features that provide improved mechanical design margin, as well as thermal performance margin (DNB). Enhanced grid designs result in a

  18. Reassembling Procedure of the Fuel Assemblies for the Nuclear Power Ship ''Mutsu''

    International Nuclear Information System (INIS)

    Japan's first voyage utilized by nuclear power was made by the nuclear powered ship ''Mutsu'' in 1990. After a research voyage in 1992, decommissioning work of the nuclear reactor for ''Mutsu'' was started to change it from the nuclear power ship to an ordinary power ship. Thirty-four irradiated fuel assemblies of ''Mutsu'' were removed from the reactor and transported to the Reactor Fuel Examination Facility (RFEF) in Nuclear Science Research Institute (NSRI) of Japan Atomic Energy Agency (JAEA). ''Mutsu'' fuel assemblies were loaded into a hot cell of RFEF using the roof gate as the top loading procedure. After the reliability confirmation tests, fuel assemblies were reassembled for reprocessing. To perform the reliability confirmation tests and reassembling, new devices were developed and installed in the hot cells, ''Fuel assembly transportation device'' for transporting the fuel assemblies between the hot cells, ''Upper nozzle cutting device'' for removing the upper nozzle from the fuel assembly, ''Fuel rod drawing device'' for drawing a fuel rod from the fuel assembly and so on. Thirty-four fuel assemblies were reassembled as six PWR type fuel assemblies in order to adjust the acceptable specifications of the reprocessing plant in JAEA: the shape of fuel assembly is the same as the PWR type commercial reactor fuel and the average enrichment of uranium in the assembly is under 4.0%. This paper reports the reassembling techniques of the ''Mutsu'' irradiated fuel assemblies for reprocessing. (author)

  19. Post-irradiation examination of Fugen reactor fuel assembly at reactor fuel examination facility

    International Nuclear Information System (INIS)

    Post-irradiation examination of the first assembly of a monitoring program for Heavy Water Reactor ''Fugen'' of PNC (Power Reactor and Nuclear Fuel Development Corporation) has been executed since Oct. 1983 at the Reactor Fuel Examination Facility, JAERI Tokai (Japan Atomic Energy Research Institute, Tokai Research Establishment). The fuel assembly is a cylindrical cluster, with 4,400mm length, composed of 28 rods in 3 concentric circles, 12 spring-grid spacers and the upper and lower tie plates. The fuel is plutonium-uranium mixed oxide (0.8 w/o), and the material of cladding tube is Zry-2. The average burnup of the fuel assembly is about 13,600 MWd/t. This paper describes the methods and some results on the post irradiation examination items as follows: 1. Radioactive measurement of water in transportation cask; 2. Visual inspection of the fuel assembly in dry cell, before and after removing the crud, by ultrasonic vibration method; 3. Chemical analyses and radioactive measurement of the crud materials; 4. Dimensional measurement of assembly length and rod-rod gaps, before and after removing the crud; 5. Disassembly and dimensional measurement of rod-rod gaps in the inner circles; 6. Several nondestructive testing techniques of fuel rods. (author)

  20. A new SCWR fuel assembly with two-row fuel rods between the hexagonal moderator channels

    International Nuclear Information System (INIS)

    Highlights: • We propose a SCWR fuel assembly with two-row fuel rods between the hexagonal moderator channels. • The new concept can resolve the contradiction between uniform and sufficient moderation. • Structural size and thermal–hydraulic performance are taken account of in the fuel assembly. • Larger infinite multiplication factor and smaller local power peaking factor could be obtained. • Two two-row hexagonal fuel assembly concepts are proposed for the engineering application. - Abstract: A new hexagonal fuel assembly (FA) design which has two rows of fuel rods between the hexagonal moderator channels is proposed for the thermal supercritical water cooled reactor (SCWR). The new concept is well considered for the performance of uniform moderation and sufficient moderation, and with respect to structural size and thermal–hydraulic performance. The neutron physical performance of the two-row hexagonal FA with acceptable configuration is discussed. The results show clearly that a better balance between uniform moderation and sufficient moderation can be obtained in the two-row hexagonal fuel assembly

  1. Behaviour of spent fuel assemblies during extended storage

    International Nuclear Information System (INIS)

    This report is the final report of the IAEA Co-ordinated Research Programme on Behaviour of Spent Fuel Assemblies During Extended Storage (BEFAST, Phase I, 1981-86). It contains the results on wet and dry spent fuel storage technologies obtained from 11 institutes (10 countries: Austria, Canada, Czechoslovakia, Finland, German Democratic Republic, Hungary, Japan, Sweden, USA and USSR) participating in the BEFAST CRP during the time period 1981-86. Names of participating institutes and chief investigators are given. The interim spent fuel storage has been recognized as an important independent step in the nuclear fuel cycle. Due to the delay in commercial reprocessing of spent fuel in some cases it should be stored up to 30-50 years or more before reprocessing or final disposal. This programme was evaluated by all its participants and observers as very important and helpful for the nuclear community and it was decided to continue it further (1986-91) as BEFAST, Phase II

  2. Nuclear fuel assembly top nozzle with improved arrangement of hold-down leaf spring assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lawson, C.N.

    1991-06-26

    A fuel assembly has a top nozzle which includes a lower adapter plate and a plurality of guide structures which are attached to an extend along the periphery of the lower plate and upwardly therefrom. The top nozzle also includes an upper hold-down plate supported by a plurality of leaf spring assemblies. The upper plate is mounted to the guide structures for vertical slidable movement relative thereto. The leaf spring assemblies are provided in a non-peripheral arrangement relative to the periphery of the lower plate in which the assemblies are provided in a non-peripheral arrangement relative to the periphery of the lower plate in which the assemblies cross the interior of the lower plate in a diagonal fashion between adjacent ones of the guide structures. (author).

  3. Determination of mixing factors for VVER-440 fuel assembly head

    International Nuclear Information System (INIS)

    CFD models have been developed for the heads of the old, the present and new type VVER-440 fuel assemblies using the experiences of former validation process. With these models, temperature distributions were investigated in typical assemblies and in-core thermocouple signals were calculated. The analyses show that coolant mixing is intensive but not-perfect in the assembly heads. Difference between the thermocouple signal and cross-sectional average temperature at measurement level depends on the assembly type. Using the models, weight factors of the rod bundle regions for the in-core thermocouple have been determined. With these factors, the thermocouple signals were estimated and results were statistically tested using the registered data of the Hungarian nuclear power plant. This test shows that deviations between measured and calculated temperatures can be significantly decreased and consequently monitoring uncertainties can be reduced with using the weight factors. (author)

  4. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  5. Calculation of the BN-600 fuel assemblies mode in a gas medium

    International Nuclear Information System (INIS)

    Potentiality of calculated modeling of temperature conditions of warming up elements of spent fuel assemblies of the BN-600 reactor during their transportation within gaseous medium is shown. The calculated modeling of spent fuel assemblies warming up in gaseous medium, their residual heat release values being different, permits substantiating and optimizing safe conditions of post-reactor handling of the fuel assemblies

  6. Design of Neptunium-bearing Fuel Assembly for Transmutation Research in CEFR

    International Nuclear Information System (INIS)

    In order to have a better understanding of irradiation performance of the fuel containing neptunium, an experimental assembly is designed for future irradiation in CEFR. There is only one fuel pin in the assembly with neptunium content of 5%. Temperature monitors and neutron fluence detectors are attached. The report presents the basic structure of the fuel pin and the assembly. (author)

  7. The Welding Process of the Small In-pile Testing Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The small in-pile testing fuel assembly is designed for high performance fuel assembly study. It has two parts of which are four fuel element with double layer cladding and a detect system for measurement of testing pressure and temperature. The fuel element is composed of UO2 pellets, the stainless steel cladding and end caps. The detect system is direct contact with the fuel element by electron beam welding. In the fabrication of the assembly, some special welding technologies are

  8. Vibration Characteristics of a Plate Type Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Yim, J.S.; Kim, H.J.; Tahk, Y.W.; Oh, J.Y.; Lee, B.H. [Nuclear Fuel Design for Research Reactor, KAERI, 305-353 Daeduck Daero 1045, Yuseong-Ku, Taejon (Korea, Republic of)

    2011-07-01

    A flat fuel plate and a box type fuel assembly for a research reactor were modeled to be finite element meshes of the ANSYS to predict dynamic characteristics, such as natural frequencies and mode shapes. These characteristics provide the basic information about their vibrations. If the model is properly prepared, it can be used for further calculations of the dynamic behaviors under the SSE or even in the static stress calculation. With the FE Model, the natural frequencies and the mode shapes of a fuel plate and a FA were obtained in air and in water environments. The effects of fluid surrounding the fuel plate and the FA as well as the combs on the natural vibration of the FA are discussed. (author)

  9. Neutron resonance transmission analysis of reactor spent fuel assemblies

    International Nuclear Information System (INIS)

    A method called Neutron Resonance Transmission Analysis (NRTA) is under study which would use a pulsed neutron beam for nondestructive isotopic assay of a complete spent fuel assembly. Neutrons removed from the collimated beam by absorption or scattering in the resonances of the various isotopes in the spent fuel appear as dips in the neutron transmission. The method is completely insensitive to matrix materials such as oxide, fuel cladding, and other structural members. Measurements on spent fuel buttons using the NBS linac as a pulsed neutron source demonstrate a high accuracy capability for the isotopes 234235236238U, 239240241242Pu, 241Am, 243Am, and several fission products. The NRTA method offers high speed and modest operational cost, and it can be implemented with commercially available medical or radiographic γ-ray generators adapted for neutron production. (Auth.)

  10. Analysis of the sub-channel of SCWR two-row fuel assembly

    International Nuclear Information System (INIS)

    Based on the COBRA-Ⅳ code, a new sub-channel code system developed for the supercritical water cooled reactor (SCWR) fuel assembly is analyzed. In order to optimize the SCWR fuel assembly design, a sub-channel analysis of two rows SCWR fuel assembly is performed, including steady-state and transient calculation. For the steady-state calculation, several channel's parameters are selected to evaluate the thermal-hydraulic performance of the fuel assemblies. Based on the steady-state results, two transient calculations (change of fuel rod power and change of coolant flow) are carried out to estimate the dynamic behavior of the fuel assemblies. The results achieved so far indicate a good applicability of the sub-channel code for the SCWR fuel assembly analysis, which is good for the future optimization of SCWR fuel assembly design. (authors)

  11. The Technology Trend of Japanese Patent for the Nuclear Fuel Assembly Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo; Lee, Nam Ho; Jeong, Kyung Min; Suh, Yong Chil; Kim, Chang Hoi; Shin, Jung Cheol

    2008-06-15

    Japanese technology patents for the nuclear fuel assembly inspection unit, from the year 1993 to the year 2006, were investigated. The fuel rods which contain fissile material are grouped together in a closely-spaced array within the fuel assembly. Various kinds of reactor including the PWR reactor are being operated in Japan. There are many kinds of nuclear fuel assemblies in Japan, and the shape and the size of these nuclear fuel assemblies are various also. As the structure of these various fuel assemblies is a regular square as the same as the Korean one, the inspection method described in Japanese technology patent can be applied to the inspection of the nuclear fuel assembly of the Korea. This report focuses on advances in VIT(visual inspection test) of nuclear fuel assembly using the state-of-the-art CCD camera system.

  12. Micro-Pocket Fission Detectors (MPFD) For Fuel Assembly Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Troy Unruh; Michael Reichenberger; Phillip Ugorowski

    2013-09-01

    Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutron detectors or “Micro-Pocket Fission Detectors” (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

  13. Process and device for fabricating nuclear fuel assembly grids

    International Nuclear Information System (INIS)

    The method for fabricating PWR fuel assembly grids consists to place the grid of which the constituent parts are held firmly in place within a frame into a sealed chamber full of inert gas. This chamber can rotate about an axis. The welding on one face at a time is carried out with a laser beam orthogonal to the axis orientation of the device. The laser source is outside of the chamber and the beam penetrates via a transparent view port

  14. Dynamics of nuclear fuel assemblies in vertical flow channels

    International Nuclear Information System (INIS)

    DYNMOD is a computer program designed to predict the dynamic behaviour of nuclear fuel assemblies in axial flow. The calculations performed by DYNMOD and the input data required by the program are described in this report. Examples of DYNMOD usage and a brief assessment of the accuracy of the dynamic model are also presented. It is intended that the report will be used as a reference manual by users of DYNMOD

  15. A CFD Simulation Process for Fast Reactor Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Kurt D. Hamman; Ray A. Berry

    2010-09-01

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k–e and SST (Menter) k–? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  16. A CFD simulation process for fast reactor fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hamman, Kurt D., E-mail: Kurt.Hamman@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Berry, Ray A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States)

    2010-09-15

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly 'benchmark' geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-{epsilon} and SST (Menter) k-{omega} were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  17. Fuel assembly simulations using LRGR-CFD and CGCFD

    International Nuclear Information System (INIS)

    In addition to the traditional fuel assembly simulation approaches using system codes, subchannel codes or porous medium approaches, as well as detailed CFD simulations to analyze single sub channels, a Low Resolution Geometry Resolving (LRGR) CFD approach and a Coarse-Grid-CFD (CGCFD) approach is taken. Both methods are based on a low resolution mesh that allows the capture of large and medium scale flow features such as recirculation zones, which cannot be reproduced by the system codes, subchannel codes and porous media approaches. The LRGR approach allows for instance fine-tuning the porous parameters which are important input for a porous medium approach. However, it should be noted that the prediction of detailed flow features such as secondary flows is not feasible. Using this approach, the consequences of flow blockages for detection possibilities and cladding temperatures can be discussed. Within the Coarse-Grid CFD approach a subgrid model (SGM) accounts for sub grid volumetric forces which are derived from validated CFD simulations. The volumetric forces take account of the non resolved physics due to the coarse mesh. The CGCFD approach with SGM can be applied to simulate complete fuel assemblies or even complete cores capturing the unique features of the complex flow induced by the fuel assembly geometry and its spacers. In such a case, grids with a very low grid resolution are employed. The current paper discusses and presents both, the CGCFD and the LRGR approaches. (author)

  18. The Dit nuclear fuel assembly physics design code

    International Nuclear Information System (INIS)

    DIT is the Combustion Engineering, Inc. (C-E) nuclear fuel assembly design code. It belongs to a class of codes, all similar in structure and strategy, which may be characterized by the spectrum and spatial calculations being performed in 2D and in a single job step for the entire assembly. The forerunner of this class of codes is the U.K.A.E.A. WIMS code, the first version of which was completed 25 years ago. The structure and strategy of assembly spectrum codes have remained remarkably similar to the original concept thus proving its usefulness. As other organizations, including C-E, have developed their own versions of the concept, many important variations have been added which significantly influence the accuracy and performance of the resulting computational tool. This paper describes and discusses those features which are unique to the DIT code and which might be of interest to the community of fuel assembly physics design code users and developers

  19. The DIT nuclear fuel assembly physics design code

    International Nuclear Information System (INIS)

    The DIT code is the Combustion Engineering, Inc. (C-E) nuclear fuel assembly design code. It belongs to a class of codes, all similar in structure and strategy, that may be characterized by the spectrum and spatial calculations being performed in two dimensions and in a single job step for the entire assembly. The forerunner of this class of codes is the United Kingdom Atomic Energy Authority WIMS code, the first version of which was completed 25 yr ago. The structure and strategy of assembly spectrum codes have remained remarkably similar to the original concept thus proving its usefulness. As other organizations, including C-E, have developed their own versions of the concept, many important variations have been added that significantly influence the accuracy and performance of the resulting computational tool. Those features, which are unique to the DIT code and which might be of interest to the community of fuel assembly physics design code users and developers, are described and discussed

  20. Calibration of the TVO spent BWR reference fuel assembly

    International Nuclear Information System (INIS)

    In 1989 the Support Programmes of Finland (FSP) and Sweden (SSP) initiated a joint task to cross calibrate the burnup of the IAEA spent BWR reference fuel assembly at the TVO AFR storage facility (TVO KPA-STORE) in Finland. The reference assembly, kept separately under the IAEA seal, is used for verification measurements of spent fuel by GBUV method (SG-NDA-38). The cross calibration was performed by establishing a calibration curve, 244Cm neutron rate versus burnup, using passive neutron assay (PNA) measurements. The declared burnup of the reference assembly was compared with the burnup value deduced from the calibration curve. A calibration line was also established by using the GBUV method with the aid of high resolution gamma ray spectrometry (HRGS). Normalization between the two different facilities was performed using sealed neutron and gamma calibration sources. The results of the passive neutron assay show consistency, better than 1 %, between the declared mean burnup of the reference assembly and the burnup deduced from the calibration curve. The corresponding consistency is within +-2 % for the HRGS measurements

  1. Fuel assembly design for APR1400 with low CBC

    Science.gov (United States)

    Hah, Chang Joo

    2015-04-01

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to ΔkTARGET. A set of new designed fuel assembly satisfies an objective function, min [f =∑i (ΔkF A-Δki ) ] , and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to ΔkTARGET as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

  2. Fuel assembly design for APR1400 with low CBC

    Energy Technology Data Exchange (ETDEWEB)

    Hah, Chang Joo, E-mail: changhah@kings.ac.kr [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to Δk{sub TARGET}. A set of new designed fuel assembly satisfies an objective function, min [f=∑{sub i}(Δk{sub FA}−Δk{sub i})], and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to Δk{sub TARGET} as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

  3. Development of an ultrasonic cleaning method for fuel assemblies

    International Nuclear Information System (INIS)

    Almost all radiation buildup in light water reactors is the result of the deposition of activated corrosion and wear products in out-of-core areas. After operation, a significant quantity of corrosion and wear products is deposited on the fuel rods as crud. At refueling shutdowns, these activation products are available for removal. If they can be quickly and easily removed, buildup of radioactivity on out-of-core surfaces and individual exposure dose can be greatly reduced. After studying various physical cleaning methods (e.g., water jet and ultrasonic), the ultrasonic cleaning method was selected as the most effective for fuel assembly cleaning. The ultrasonic cleaning method is especially able to efficiently clean the fuel without removing the channel box. The removed crud in the channel box would be swept out to the filtration unit. Parameter survey tests were carried out to evaluate the optimum conditions for ultrasonic cleaning using a mock-up of a short section of fuel assembly with the channel box. The ultrasonic device used was a 600-W ultrasonic transducer operating at 26-kHz ultrasonic frequency

  4. Development of an ultrasonic cleaning method for fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Heki, H.; Komura, S.; Kato, H.; Sakai, H. (Toshiba Corp., Kawasaki City (Japan)); Hattori, T. (Tokyo Electric Power Co., Kashiwazaki-shi (Japan))

    1991-01-01

    Almost all radiation buildup in light water reactors is the result of the deposition of activated corrosion and wear products in out-of-core areas. After operation, a significant quantity of corrosion and wear products is deposited on the fuel rods as crud. At refueling shutdowns, these activation products are available for removal. If they can be quickly and easily removed, buildup of radioactivity on out-of-core surfaces and individual exposure dose can be greatly reduced. After studying various physical cleaning methods (e.g., water jet and ultrasonic), the ultrasonic cleaning method was selected as the most effective for fuel assembly cleaning. The ultrasonic cleaning method is especially able to efficiently clean the fuel without removing the channel box. The removed crud in the channel box would be swept out to the filtration unit. Parameter survey tests were carried out to evaluate the optimum conditions for ultrasonic cleaning using a mock-up of a short section of fuel assembly with the channel box. The ultrasonic device used was a 600-W ultrasonic transducer operating at 26-kHz ultrasonic frequency.

  5. Tomographic techniques for safeguards measurements of nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lundqvist Saleh, Tobias

    2007-10-15

    Nuclear power is currently experiencing increased interest over the world. New nuclear reactors are being built and techniques for taking care of the nuclear waste are being developed. This development puts new demands and standards to safeguards, i.e. the international efforts for ensuring the non-proliferation of nuclear weapons. New measuring techniques and devices are continuously being developed for enhancing the ability to detect diversion of fissile material. In this thesis, tomographic techniques for application in safeguards are presented. Tomographic techniques can non-destructively provide information of the inner parts of an object and may thus be used to control that no material is missing from a nuclear fuel assembly. When using the tomographic technique described in this thesis, the radiation field around a fuel assembly is first recorded. In a second step, the internal source distribution is mathematically reconstructed based on the recorded data. In this work, a procedure for tomographic safeguards measurements is suggested and the design of a tomographic measuring device is presented. Two reconstruction algorithms have been specially developed and evaluated for the application on nuclear fuel; one algorithm for image reconstruction and one for reconstructing conclusive data on the individual fuel rod level. The combined use of the two algorithms is suggested. The applicability for detecting individual removed or replaced rods has been demonstrated, based on experimental data

  6. Tomographic techniques for safeguards measurements of nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Nuclear power is currently experiencing increased interest over the world. New nuclear reactors are being built and techniques for taking care of the nuclear waste are being developed. This development puts new demands and standards to safeguards, i.e. the international efforts for ensuring the non-proliferation of nuclear weapons. New measuring techniques and devices are continuously being developed for enhancing the ability to detect diversion of fissile material. In this thesis, tomographic techniques for application in safeguards are presented. Tomographic techniques can non-destructively provide information of the inner parts of an object and may thus be used to control that no material is missing from a nuclear fuel assembly. When using the tomographic technique described in this thesis, the radiation field around a fuel assembly is first recorded. In a second step, the internal source distribution is mathematically reconstructed based on the recorded data. In this work, a procedure for tomographic safeguards measurements is suggested and the design of a tomographic measuring device is presented. Two reconstruction algorithms have been specially developed and evaluated for the application on nuclear fuel; one algorithm for image reconstruction and one for reconstructing conclusive data on the individual fuel rod level. The combined use of the two algorithms is suggested. The applicability for detecting individual removed or replaced rods has been demonstrated, based on experimental data

  7. Operation experience of WWER-440 fuel assemblies and measures to increase fuel reliability

    International Nuclear Information System (INIS)

    The paper presents technical data for the fuel cycles used in 14 WWER-440 reactors of B-213 type situated outside CIS-territory on the basis of the 2001 operational results. The paper reflects the dynamics of average and maximum fuel burnup as well as information on the annual rate of the leaking fuel rods for the above reactor group identified during the 1997-2001 discharge period. As an example of work performed by RIAR in 2001 the paper brings forth the PIE-results of a leaking WWER-440 fuel assemblies (FAs). It is reported that the reason behind the leaking and failed fuel rods of the FA was interaction with a foreign object being in the coolant flow. The paper describes the measures taken by the NPPs together with the Supplier (JSC TVEL) and Manufacturer (JSC MSZ) to enhance the fuel operational safety. (author)

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

    International Nuclear Information System (INIS)

    Experimental data are summarized on irradiation-induced dimensional changes of fuel compacts and graphite sleeves of the first to ninth OGL-1 fuel assemblies. The range of fast-neutron fluence is up to 4 x 1024 n/m2 (E > 0.18 MeV); and that of irradiation temperature is 900 - 1400 deg C for fuel compacts and 800 - 1050 deg C for graphite sleeves. The dimensional change of the fuel compacts was shrinkage under these test conditions, and the shrinkage fraction increased almost linearly with fast-neutron fluence. The shrinkage fraction of the fuel compacts was larger by 20 % in the axial direction than in the radial direction. Influence of the irradiation temperature on the dimensional-change behavior of the fuel compacts was not observed clearly; presumably the influence was hidden by scatter of the data because of low level of the fast-neutron fluence and the resultant small dimensional changes. (author)

  9. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 1, Activation measurements and comparison with calculations for spent fuel assembly hardware

    Energy Technology Data Exchange (ETDEWEB)

    Luksic, A.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1. 5 refs., 4 figs., 21 tabs.

  10. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 2, Calculated activity profiles of spent nuclear fuel assembly hardware for pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  11. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  12. Revaluation on measured burnup values of fuel assemblies by post-irradiation experiments at BWR plants

    International Nuclear Information System (INIS)

    Fuel composition data for 8x8 UO2, Tsuruga MOX and 9x9-A type UO2 fuel assemblies irradiated in BWR plants were measured. Burnup values for measured fuels based on Nd-148 method were revaluated. In this report, Nd-148 fission yield and energy per fission obtained by burnup analyses for measured fuels were applied and fuel composition data for the measured fuel assemblies were revised. Furthermore, the adequacies of revaluated burnup values were verified through the comparison with burnup values calculated by the burnup analyses for the measured fuel assemblies. (author)

  13. Control assembly for controlling a fuel cell system during shutdown and restart

    Science.gov (United States)

    Venkataraman, Ramki; Berntsen, George; Carlson, Glenn L.; Farooque, Mohammad; Beachy, Dan; Peterhans, Stefan; Bischoff, Manfred

    2010-06-15

    A fuel cell system and method in which the fuel cell system receives and an input oxidant gas and an input fuel gas, and in which a fuel processing assembly is provided and is adapted to at least humidify the input fuel gas which is to be supplied to the anode of the fuel cell of the system whose cathode receives the oxidant input gas via an anode oxidizing assembly which is adapted to couple the output of the anode of the fuel cell to the inlet of the cathode of the fuel cell during normal operation, shutdown and restart of the fuel cell system, and in which a control assembly is further provided and is adapted to respond to shutdown of the fuel cell system during which input fuel gas and input oxidant gas cease to be received by the fuel cell system, the control assembly being further adapted to, when the fuel cell system is shut down: control the fuel cell system so as to enable a purging gas to be able to flow through the fuel processing assembly to remove humidified fuel gas from the processing assembly and to enable a purging gas to be able to flow through the anode of the fuel cell.

  14. Nuclear material attractiveness: an assessment of used-fuel assemblies

    International Nuclear Information System (INIS)

    This paper examines the material attractiveness of used-fuel assemblies in a hypothetical scenario in which terrorists steal one or more assemblies in order to use the special nuclear materials (SNM) within an assembly in a nuclear explosive device. For assessing material attractiveness, this paper uses the Figure of Merit (FOM) that was used in earlier studies to examine the attractiveness of the SNM associated with the reprocessing of used light water reactor (LWR) fuel by various reprocessing schemes. However, for a theft scenario the mass used in the Acquisition Factor of the FOM is the mass of the stolen object conta ining SNM ; whereas the mass used for analyzing the material attractiveness of the products of various reprocessing schemes in the earlier studies was a fraction of the bare critical mass in recognition that a successful proliferator must avoid a criticality accident. This paper will indicate how long after discharge the radiation emanating from a cooling assembly is no longer self-protecting. Additionally, this paper will give the time scale for the SNM within the assembly to become more attractive. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of ''attractiveness levels'' that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented. Additionally, this paper discusses how the results presented herein impact the application of safeguards and the securitization of SNM, and how they could be used to help inform policy makers.

  15. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  16. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array

  17. Criteria for removal of defective fuel rod from fuel assembly under repair without cladding rupture

    International Nuclear Information System (INIS)

    During repair of a failed fuel assembly (FA) there is a risk of cladding rupture while a defective fuel rod is forced out of the assembly skeleton. To reduce the corresponding risks, a program of experimental and analytical studies for WWER fuel was performed. It resulted in formulation of criteria for successful removal of the defective fuel rod from the FA under repair. 'Successful' means that no cladding rupture occurs. The paper summarizes the available data of post-irradiation examinations of WWER FAs with leaking fuel rods. A technique for express estimation of hydrogen content in cladding of a defective fuel rod is presented. The degradation of cladding mechanical properties can be estimated with this technique as well. A criterion of severe secondary hydriding involved in the risk analysis is also discussed. Finally, it is shown how the information on operation conditions may be used for prompt evaluation of the limiting force for successful removal of a defective fuel rod during FA repair in the inspection stand. (author)

  18. Manufacture of a Dual-Cooled Fuel Assembly Mockup for Mechanical Characterization Tests

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jaeyong; Kim, Hyungkyu; Yoon, Kyungho; Lee, Kanghee; Kang, Heungseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    All components were made of stainless steel 304 for research. A DUO fuel assembly mockup was assembled by mechanical fastening and laser welding methods with them. The conceptual feasibility of each component was checked through it. In this paper, manufactured items for a DUO fuel and a DUO fuel assembly are briefly described. Although the research of a DUO fuel has been done by USA, they have just focused on pellets, not mechanical parts such as TEP/BEP, GTs, and SGs. We designed and manufactured them and assembled a DUO fuel assembly. The realizable possibility of a DUO fuel assembly was checked. Mechanical characterization tests will be performed to measure the DUO fuel's mechanical properties such as bending rigidity, modal characteristics, impact durability, etc.

  19. The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology

    Directory of Open Access Journals (Sweden)

    Bobrov Evgenii

    2016-01-01

    Full Text Available This paper shows basic features of different fuel assembly (FA application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water–fuel ratio in the VVER FA affects on the fuel characteristics produced by REMIX technology during multiple recycling.

  20. Design report of the disposal canister for twelve fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.) 35 refs.

  1. Study on pressure drop prediction of Tight Lattice Fuel Assembly using CFD

    International Nuclear Information System (INIS)

    This paper presents a prediction method about pressure drop of the tight lattice fuel assembly. To evaluate the core performance, the pressure drop of fuel assembly is important parameter for the reliability and safety. The shape of fuel assembly is complicated and the shape has a strong effect on pressure drop. Therefore, to obtain the pressure drop of fuel assembly, the experiments are needed. However the experiments need a lot of time and money. The purpose of this study is to predict the pressure drop of the tight lattice fuel assembly using CFD (computational fluid dynamics) and law of two phase flow similarity without experiments. Prediction method for pressure drop is, first the shape of fuel assembly was reproduced by 3D-CAD, then to evaluate the parameters depending on the shape using CFD analysis under the single phase flow condition. And then, to calculate pressure drop of two phase flow using law of two phase flow similarity. The predicted results by this method were compared with the experimental results of 3 types of the tight lattice fuel assemblies, 7-rod, 14-rod, and 37-rod fuel assembly. Fluid conditions (pressure, mass flux and quality) of test results covered the typical operating range of the LWR fuel assembly. It was found from comparison result that this method can predict the pressure drop of the LWR fuel assembly with sufficient accuracy. The prediction accuracy (predicted value/test value) is about 10%. (author)

  2. Development of a reference spent fuel library of 17x17 PWR fuel assemblies

    International Nuclear Information System (INIS)

    One of the most common ways to investigate new Non-Destructive Assays (NDA) for the spent fuel assemblies are Monte Carlo simulations. In order to build realistic models the user must define in an accurate way the material compositions and the source terms in the system. This information can be obtained using burnup codes such as ORIGEN-ARP and ALEPH2.2, developed at SCK-CEN. These software applications allow the user to select the irradiation history of the fuel assembly and to calculate the corresponding isotopic composition and neutron/gamma emissions as a function of time. In the framework of the development of an innovative NDA for spent fuel verifications, SCK•CEN built an extensive fuel library for 17x17 PWR assemblies, using both ORIGEN-ARP and ALEPH2.2. The parameters considered in the calculations were initial enrichment, discharge burnup, and cooling time. The combination of these variables allows to obtain more than 1500 test cases. Considering the broad range of the parameters, the fuel library can be used for other purposes apart from spent fuel verifications, for instance for the direct disposal in geological repositories. In addition to the isotopic composition of the spent fuel, the neutron and photon emissions were also calculated and compared between the two codes. The comparison of the isotopic composition showed a good agreement between the codes for most of the relevant isotopes in the spent fuel. However, specific isotopes as well as neutron and gamma spectra still need to be investigated in detail.

  3. Out-of-pile Verifying Test for the Hydraulic Stability of the CARR Standard Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The CARR standard fuel element is a flat-plate-type assembly. A fuel plate consists of 0.6 mmthickness layer of uranium- silicon - aluminum fuel (U3Si2-Al) and 0.38 mm thickness of aluminumcladding. The fuel plates are attached to aluminum alloy side plates by a "roll swaging" technique. Thistype of fuel assembly is first used in China. The testing simulates the in-pile thermal-hydraulic operating conditions except for neutron

  4. Experimental and calculational study of temperature distributions in deformed model fuel assemblies of fast reactors

    International Nuclear Information System (INIS)

    Experimental and calculational data tastify to absence of temperature nonuniformity stabilization in fuel assembly peripheral area. The effect of fuel lattice deformation on the fuel assembly temperature field at shroud crushing in the core centre is demonstrated. 17 refs.; 21 figs

  5. Simulating fuel assemblies with low resolution CFD approaches

    Energy Technology Data Exchange (ETDEWEB)

    Roelofs, F., E-mail: roelofs@nrg.eu [NRG, Petten (Netherlands); Gopala, V.R. [NRG, Petten (Netherlands); Chandra, L. [IIT Rajasthan (India); Viellieber, M.; Class, A. [KIT, Karlsruhe (Germany)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Justification for development of low resolution mesh approaches. Black-Right-Pointing-Pointer Mathematical background of the approaches. Black-Right-Pointing-Pointer Meshing considerations for different approaches are presented. Black-Right-Pointing-Pointer Examples of applications are provided. - Abstract: In addition to the traditional fuel assembly simulations using system codes, subchannel codes or porous medium approaches, as well as detailed CFD simulations to analyze single sub channels, a Low Resolution Geometry Resolving (LRGR) CFD approach and a Coarse-Grid-CFD (CGCFD) approach are taken. Both methods are based on a low resolution mesh that allows the capture of large and medium scale flow features such as recirculation zones, which are difficult to be reproduced by the system codes, subchannel codes and porous media approaches. The LRGR approach allows for instance fine-tuning the porous parameters which are important input for a porous medium approach. However, it should be noted that the prediction of detailed flow features such as secondary flows (small flows in the direction perpendicular to the main flow) is not feasible. Using this approach, the consequences of flow blockages for detection possibilities and cladding temperatures can be discussed. The goal of the CGCFD approach with SGM is that it can be applied to simulate complete fuel assemblies or even complete cores capturing the unique features of the complex flow induced by the fuel assembly geometry and its spacers. In such a case, grids with a very low grid resolution are employed. Within the CGCFD a subgrid model (SGM) accounts for sub grid volumetric forces which are derived from validated CFD simulations. The volumetric forces take account of the non resolved physics due to the coarse mesh. The current paper discusses and presents both, the CGCFD and the LRGR approaches.

  6. Welding fixture for nuclear reactor fuel assembly grid

    International Nuclear Information System (INIS)

    A welding fixture for a nuclear fuel assembly grid having rigid top and bottom members and having apparatus for releasably securing them together. The bottom and top members each have an egg-crate configuration of interleaved fixture straps. Top notches on the bottom fixture straps' top edges are placed to engage the grid straps' bottom edges when the grid straps are aligned for welding. Likewise, bottom notches on the top fixture straps' bottom edges are placed to engage the grid straps' top edges when the grid straps are aligned for welding

  7. Critical experiments for BWR fuel assemblies with cluster of gadolinia rods

    International Nuclear Information System (INIS)

    Gadolinia-bearing fuel rods are needed for high-burnup fuels. Strong neutron absorption of gadolinia makes an assembly heterogeneous from the viewpoint of reactor physics. The cluster of gadolinia-bearing fuel rods is useful for higher-burnup fuels than current fuels. Few critical experiments have been reported for fuel assemblies with the cluster of gadolinia-bearing fuel rods. We conducted critical experiments for BWR fuel assemblies with the cluster of gadolinia-bearing fuel rods in the Toshiba Nuclear Critical Assembly (NCA). Critical water level and power distribution were measured. Measurements were compared with analyses by a continuous-energy Monte Carlo code, MCNP, with the JENDL3.3 nuclear data library. (author)

  8. 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)

  9. Membrane electrode assemblies for unitised regenerative polymer electrolyte fuel cells

    Science.gov (United States)

    Wittstadt, U.; Wagner, E.; Jungmann, T.

    Membrane electrode assemblies for regenerative polymer electrolyte fuel cells were made by hot pressing and sputtering. The different MEAs are examined in fuel cell and water electrolysis mode at different pressure and temperature conditions. Polarisation curves and ac impedance spectra are used to investigate the influence of the changes in coating technique. The hydrogen gas permeation through the membrane is determined by analysing the produced oxygen in electrolysis mode. The analysis shows, that better performances in both process directions can be achieved with an additional layer of sputtered platinum on the oxygen electrode. Thus, the electrochemical round-trip efficiency can be improved by more than 4%. Treating the oxygen electrode with PTFE solution shows better performance in fuel cell and less performance in electrolysis mode. The increase of the round-trip efficiency is negligible. A layer sputtered directly on the membrane shows good impermeability, and hence results in high voltages at low current densities. The mass transportation is apparently constricted. The gas diffusion layer on the oxygen electrode, in this case a titanium foam, leads to flooding of the cell in fuel cell mode. Stable operation is achieved after pretreatment of the GDL with a PTFE solution.

  10. Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

    Science.gov (United States)

    Shore, Lawrence; Matlin, Ramail

    2009-12-22

    The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

  11. The Conceptual Design for a Fuel Assembly of a New Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, J-S.; Cho, Y-G.; Yoon, D-B.; Dan, H-J.; Chae, H-T.; Park, C.

    2004-10-06

    A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibration characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.

  12. Nuclear fuel assembly top nozzle with improved arrangement of hold-down leaf spring assemblies

    Energy Technology Data Exchange (ETDEWEB)

    De Mario, E.E.; Lawson, C.N.

    1991-10-15

    This patent describes a top nozzle for use in a fuel assembly having guide thimbles for mounting the top nozzle. It comprises: a lower adapter plate having a periphery bounding an interior thereof mountable to the guide thimbles: guide structures attached to and extending along the periphery of the adapter plate and upwardly therefrom; an upper hold-down plate mounted to the guide structures for slidable movement relative thereto such that the upper plate can move toward and away from the interior of the lower plate within the space bounded by the guide structures as the upper plate slidably moves along the guide structures; and leaf spring assemblies interposed between and engaged with the lower and upper plates so as to yieldably support the upper plate in spaced relation above the lower plate and bias the upper plate for movement away from the lower plate; the leaf spring assemblies being provided in a non-peripheral arrangement relative to the periphery of the lower plate in which the assemblies cross the interior of the lower plate in a diagonal fashion between adjacent ones of the guide structures.

  13. Irradiation experiments of the 13th-15th OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    The Oarai Gas Loop-1, OGL-1, which had been installed in the Japan Materials Testing Reactor (JMTR) of the Oarai Research Establishment of the Japan Atomic Energy Research Institute (JAERI), was an in-pile helium loop for irradiation of high-temperature gas-cooled reactor fuels at high pressure and temperature. This report describes results of fabrication, irradiation and post-irradiation examinations (PIEs) of the 13th-15th OGL-1 fuel assemblies. The 13th and 15th fuel assemblies employed the first-charge fuel of the High Temperature Engineering Test Reactor (HTTR). The 13th assembly was loaded with a high quality fuel, whose as-produced failure fraction had been drastically decreased, compared with that for fuels before that time. The 15th assembly was loaded with a fuel, which had been produced by the same apparatus that was used afterwards for the first charge fuel of the HTTR. Both of these fuel assemblies gave good results in PIEs as well as in the fission-gas release rates during irradiation. The 14th fuel assembly used a trial product of an advanced fuel for high burnup utilization, which employed coated fuel particles (CFPs) with thicker coating layers than those for the first charge fuel. This fuel assembly indicated a spike release of fission gas during irradiation at 1500degC after a transient temperature increase up to this value. As a whole, all of the 13th - 15th assemblies demonstrated good performance of the loaded fuels, giving significantly lower values in fission-gas release rates during irradiation and in failure fractions of CFPs after irradiation, than the corresponding design limit values for the first charge fuel of the HTTR. (author)

  14. Irradiation experiments of the 13th-15th OGL-1 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio; Sawa, Kazuhiro; Kitajima, Toshio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Shiratori, Tetsuo; Kikuchi, Hironobu; Fukuda, Kousaku; Itoh, Tadaharu; Waragai, Heita [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-01-01

    The Oarai Gas Loop-1, OGL-1, which had been installed in the Japan Materials Testing Reactor (JMTR) of the Oarai Research Establishment of the Japan Atomic Energy Research Institute (JAERI), was an in-pile helium loop for irradiation of high-temperature gas-cooled reactor fuels at high pressure and temperature. This report describes results of fabrication, irradiation and post-irradiation examinations (PIEs) of the 13th-15th OGL-1 fuel assemblies. The 13th and 15th fuel assemblies employed the first-charge fuel of the High Temperature Engineering Test Reactor (HTTR). The 13th assembly was loaded with a high quality fuel, whose as-produced failure fraction had been drastically decreased, compared with that for fuels before that time. The 15th assembly was loaded with a fuel, which had been produced by the same apparatus that was used afterwards for the first charge fuel of the HTTR. Both of these fuel assemblies gave good results in PIEs as well as in the fission-gas release rates during irradiation. The 14th fuel assembly used a trial product of an advanced fuel for high burnup utilization, which employed coated fuel particles (CFPs) with thicker coating layers than those for the first charge fuel. This fuel assembly indicated a spike release of fission gas during irradiation at 1500degC after a transient temperature increase up to this value. As a whole, all of the 13th - 15th assemblies demonstrated good performance of the loaded fuels, giving significantly lower values in fission-gas release rates during irradiation and in failure fractions of CFPs after irradiation, than the corresponding design limit values for the first charge fuel of the HTTR. (author)

  15. Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

    International Nuclear Information System (INIS)

    A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception

  16. The PS booster

    CERN Multimedia

    1972-01-01

    The PS booster which accelerates protons from the linac at an energy of 50 MeV to an energy of 800 MeV before injecting them into the main magnet ring of the synchrotron. The booster consists of four superposed rings. In the photograph can be seen the input beam line from the linac and the output beam lines, where beams from the four booster levels have been combined into two beams before final recombination.

  17. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Science.gov (United States)

    Panferov, Pavel; Kochkin, Viacheslav; Erak, Dmitry; Makhotin, Denis; Reshetnikov, Alexandr; Timofeev, Andrey

    2016-02-01

    The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  18. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Directory of Open Access Journals (Sweden)

    Panferov Pavel

    2016-01-01

    Full Text Available The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  19. Cap assembly for a bundled tube fuel injector

    Energy Technology Data Exchange (ETDEWEB)

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-04-26

    A cap assembly for a bundled tube fuel injector includes an impingement plate and an aft plate that is disposed downstream from the impingement plate. The aft plate includes a forward side that is axially separated from an aft side. A tube passage extends through the impingement plate and the aft plate. A tube sleeve extends through the impingement plate within the tube passage towards the aft plate. The tube sleeve includes a flange at a forward end and an aft end that is axially separated from the forward end. A retention plate is positioned upstream from the impingement plate. A spring is disposed between the retention plate and the flange. The spring provides a force so as to maintain contact between at least a portion of the aft end of the tube sleeve and the forward side of the aft plate.

  20. Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

    Energy Technology Data Exchange (ETDEWEB)

    Viererbl, L., E-mail: vie@ujv.cz [Nuclear Research Institute Rez plc, 250 68 Husinec-Rez 130 (Czech Republic); Research Centre Rez Ltd. (Czech Republic); Lahodova, Z.; Voljanskij, A.; Klupak, V.; Koleska, M. [Nuclear Research Institute Rez plc, 250 68 Husinec-Rez 130 (Czech Republic); Research Centre Rez Ltd. (Czech Republic); Cabalka, M. [Nuclear Research Institute Rez plc, 250 68 Husinec-Rez 130 (Czech Republic); Turek, K. [Nuclear Physics Institute, Academy of Sciences of the Czech Republic (Czech Republic)

    2011-10-01

    During operation of a fission nuclear reactor, many radionuclides are generated in fuel by fission and activation of {sup 235}U, {sup 238}U and other nuclides present in the assembly. After removal of a fuel assembly from the core, these radionuclides are sources of different types of radiation. Gamma and neutron radiation emitted from an assembly can be non-destructively detected with different types of detectors. In this paper, a new method of measurement of radiation from a spent fuel assembly is presented. It is based on usage of passive detectors, such as alanine dosimeters for gamma radiation and track detectors for neutron radiation. Measurements are made on the IRT-2M spent fuel assemblies used in the LVR-15 research reactor. During irradiation of detectors, the fuel assembly is located in a water storage pool at a depth of 6 m. Detectors are inserted into central hole of the assembly, irradiated for a defined time interval, and after the detectors removed from the assembly, gamma dose or neutron fluence are evaluated. Measured profiles of gamma dose rate and neutron fluence rate inside of the spent fuel assembly are presented. This measurement can be used to evaluate relative fuel burn-up.

  1. Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

    Science.gov (United States)

    Viererbl, L.; Lahodová, Z.; Voljanskij, A.; Klupák, V.; Koleška, M.; Cabalka, M.; Turek, K.

    2011-10-01

    During operation of a fission nuclear reactor, many radionuclides are generated in fuel by fission and activation of 235U, 238U and other nuclides present in the assembly. After removal of a fuel assembly from the core, these radionuclides are sources of different types of radiation. Gamma and neutron radiation emitted from an assembly can be non-destructively detected with different types of detectors. In this paper, a new method of measurement of radiation from a spent fuel assembly is presented. It is based on usage of passive detectors, such as alanine dosimeters for gamma radiation and track detectors for neutron radiation. Measurements are made on the IRT-2M spent fuel assemblies used in the LVR-15 research reactor. During irradiation of detectors, the fuel assembly is located in a water storage pool at a depth of 6 m. Detectors are inserted into central hole of the assembly, irradiated for a defined time interval, and after the detectors removed from the assembly, gamma dose or neutron fluence are evaluated. Measured profiles of gamma dose rate and neutron fluence rate inside of the spent fuel assembly are presented. This measurement can be used to evaluate relative fuel burn-up.

  2. A spray cooling technique for spent fuel assembly stored in pool

    International Nuclear Information System (INIS)

    For the safety of spent nuclear fuel assemblies stored in storage pool in the extreme condition where the water is lost completely, a passive spray cooling technique was designed, and its effectiveness has been validated by a functional experiment. The spray cooling characteristics of the spent fuel assembly have also been investigated by the experiment.

  3. A spray cooling technique for spent fuel assembly stored in pool

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Dao-Gang; Cao, Q. [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Wang, Y.; Zhong, Hao-Liang; Duan, Xiao-Han

    2016-05-15

    For the safety of spent nuclear fuel assemblies stored in storage pool in the extreme condition where the water is lost completely, a passive spray cooling technique was designed, and its effectiveness has been validated by a functional experiment. The spray cooling characteristics of the spent fuel assembly have also been investigated by the experiment.

  4. AGS Booster prototype magnets

    Energy Technology Data Exchange (ETDEWEB)

    Danby, G.; Jackson, J.; Lee, Y.Y.; Phillips, R.; Brodowski, J.; Jablonski, E.; Keohane, G.; McDowell, B.; Rodger, E.

    1987-03-19

    Prototype magnets have been designed and constructed for two half cells of the AGS Booster. The lattice requires 2.4m long dipoles, each curved by 10/sup 0/. The multi-use Booster injector requires several very different standard magnet cycles, capable of instantaneous interchange using computer control from dc up to 10 Hz.

  5. Modeling the effect of engine assembly mass on engine friction and vehicle fuel economy

    Science.gov (United States)

    An, Feng; Stodolsky, Frank

    An analytical model is developed to estimate the impact of reducing engine assembly mass (the term engine assembly refers to the moving components of the engine system, including crankshafts, valve train, pistons, and connecting rods) on engine friction and vehicle fuel economy. The relative changes in frictional mean effective pressure and fuel economy are proportional to the relative change in assembly mass. These changes increase rapidly as engine speed increases. Based on the model, a 25% reduction in engine assembly mass results in a 2% fuel economy improvement for a typical mid-size passenger car over the EPA Urban and Highway Driving Cycles.

  6. Buoyancy-driven flow excursions in fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, J.E.; Paul, P.K.; Menna, J.D.

    1995-12-31

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating moderator downward through channels in cylindrical fuel tubes. Powers were limited to prevent a flow excursion from occurring in one or more of these parallel channels. During full-power operation, limits prevented a boiling flow excursion from taking place. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increases beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of historical levels.

  7. Buoyancy-driven flow excursions in fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, J.E.; Paul, P.K.; Menna, J.D. [Westinghouse Savannah River Company, Aiken, SC (United States)

    1995-09-01

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating heavy water moderator downward through channels in cylindrical fuel tubes. Powers were limited to safeguard against a flow excursion in one of more of these parallel channels. During-full-power operation, limits safeguarded against a boiling flow excursion. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increased beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of the limiting power for previous long-term reactor operations.

  8. Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor

    Directory of Open Access Journals (Sweden)

    Daxin Gong

    2015-01-01

    Full Text Available Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved.

  9. Sensitivity and System Response of Pin Power Peaking in VVER-1000 Fuel Assembly Using TSUNAMI-2D

    Science.gov (United States)

    Frybort, J.

    2014-04-01

    Pin power peaking in a VVER-1000 fuel assembly and its sensitivity and uncertainty was analyzed by TSUNAMI-2D code. Several types of fuel assemblies were considered. They differ in number and position of gadolinium fuel pins. The calculations were repeated for several fuel compositions obtained by fuel depletion calculation. The results are quantified sensitivity data, which can be used for enrichment profiling.

  10. ROSA-IV Large Scale Test Facility (LSTF) system description for second simulated fuel assembly

    International Nuclear Information System (INIS)

    The ROSA-IV Program's Large Scale Test Facility (LSTF) is a test facility for integral simulation of thermal-hydraulic response of a pressurized water reactor (PWR) during small break loss-of-coolant accidents (LOCAs) and transients. In this facility, the PWR core nuclear fuel rods are simulated using electric heater rods. The simulated fuel assembly which was installed during the facility construction was replaced with a new one in 1988. The first test with this second simulated fuel assembly was conducted in December 1988. This report describes the facility configuration and characteristics as of this date (December 1988) including the new simulated fuel assembly design and the facility changes which were made during the testing with the first assembly as well as during the renewal of the simulated fuel assembly. (author)

  11. Influence of Spacer Grid Outer Strap on Fuel Assembly Thermal Hydraulic Performance

    Directory of Open Access Journals (Sweden)

    Jingwen Yan

    2014-01-01

    Full Text Available The outer strap as a typical structure of a spacer grid enhances the mechanical strength, decreases hang-up susceptibility, and also influences thermal hydraulic performance, for example, pressure loss, mixing performance, and flow distribution. In the present study, a typical grid spacer with different outer strap designs is adopted to investigate the influence of outer strap design on fuel assembly thermal hydraulic performance by using a commercial computational fluid dynamics (CFD code, ANSYS CFX, and a subchannel analysis code, FLICA. To simulate the outer straps’ influence between fuel assemblies downstream, four quarter-bundles from neighboring fuel assemblies are constructed to form the computational domain. The results show that the outer strap design has a major impact on cross-flow between fuel assemblies and temperature distribution within the fuel assembly.

  12. Measurement of uranium and plutonium content in a fuel assembly using the RPI spent fuel assay device

    International Nuclear Information System (INIS)

    In this paper we report measurements of the significant parameters, the sensitivities of the slowing-down-time assay device to the fissile contents of a boiling water reactor (BWR) assembly mock-up of fresh fuel

  13. Neutronic and thermohydraulic characteristics of a new breeding thorium–uranium mixed SCWR fuel assembly

    International Nuclear Information System (INIS)

    Highlights: • A new Th–U mixed fuel assembly for SCWR has been introduced and investigated. • Neutronic and thermohydraulic characteristics of the new assembly have been studied. • The new fuel assembly satisfies design rules of SCWR. • The introduced fuel assembly can fulfill the sustainable breeding Th–U cycle. • The new fuel assembly also has advantages with respect to lower generation of minor actinides and reactor safety. - Abstract: The exploitation of thorium fuel is a promising way to overcome the pressing problems of nuclear fuel supply, nuclear waste and nuclear proliferation. In this paper, a novel conceptual design of a breeding thorium–uranium (Th–U) mixed fuel assembly in SCWR is proposed, which is aimed to achieve the breeding ratio bigger than 1.0, so as to fulfill the sustainable breeding thorium–uranium cycle. Through the calculations of neutronics and neutronic/thermohydraulic (N–T) coupling, the results indicate that the introduced conceptual design of a breeding Th–U mixed fuel assembly in SCWR satisfies design rules of SCWR, with considerable advantages with respect to breeding performance, lower minor actinide generation and reactor safety

  14. Estimation of the nuclear fuel assembly eigenfrequencies in the probability sense

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2014-12-01

    Full Text Available The paper deals with upper and lower limits estimation of the nuclear fuel assembly eigenfrequencies, whose design and operation parameters are random variables. Each parameter is defined by its mean value and standard deviation or by a range of values. The gradient and three sigma criterion approach is applied to the calculation of the upper and lower limits of fuel assembly eigenfrequencies in the probability sense. Presented analytical approach used for the calculation of eigenfrequencies sensitivity is based on the modal synthesis method and the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and load-bearing skeleton linked by spacer grids. The method is applied for the Russian TVSA-T fuel assembly in the WWER1000/320 type reactor core in the Czech nuclear power plant Temelín.

  15. Effects of the chemical decontamination on the component parts of the ATR fuel assembly

    International Nuclear Information System (INIS)

    The chemical decontamination technique has been developed in order to remove the crud adhering to the surface of the components constructing the primary coolant system, as a part of the measure to decrease the exposure in the annual inspection. The technique has been already applied to the prototype reactor 'Fugen', in the core of which the fuel assemblies were not loaded. The chemical decontamination, for the core in which the fuel assemblies are loaded, has been planned for the purpose of improving the utilization factor. It is necessary to confirm, through the test before putting the plan into practice, that the decontamination reagent does not exert a bad influence upon the components constructing the fuel assembly. This report describes the test results which have been carried out so as to investigate the influence of the reagent on the components constructing the fuel assembly. The outline of the results is as follows: (1) The susceptibility to stress corrosion cracking of the chemical decontamination treatment and the residual decontamination reagent on the components constructing the fuel assembly is low enough. (2) The chemical decontamination treatment and the residual decontamination reagent do not exert a bad influence upon the integrity of the fuel assembly concerning the fuel rod holding function of the spacer and the characteristics of the fretting wear caused on the fuel claddings. (author)

  16. Preliminary Design of U-Mo Alloy Dispersion Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>As a kind of new type fuel for research reactor, high density U-Mo alloy dispersion fuel which will substitute current fuel in the future is being studied and developed by RERTR. There are two characteristics

  17. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

  18. Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

    Energy Technology Data Exchange (ETDEWEB)

    Goodsell, Alison Victoria [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzl, Vladimir [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rael, Carlos D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-09-18

    Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performed in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.

  19. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    OpenAIRE

    Waseem; Murtaza Ghulam; Siddiqui Ashfaq Ahmad; Akhtar Syed Waseem

    2016-01-01

    Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA) of Chashma Nuclear Power Plant-1 (CHASNUPP-1) at room temperature in air. The non-linear contact and structural tensile analysis have been performed using ANSYS 13.0, in order to determine the fuel assembly (FA) elongation behaviour as well as the location and values of the stress intensity and stresses developed in ax...

  20. Numerical simulation of gas dynamics and heat exchange tasks in fuel assemblies of the nuclear reactors

    International Nuclear Information System (INIS)

    This report presents a PC-based program for solution gas dynamics and heat exchange mathematical tasks in fuel assemblies of the fast-neutron nuclear reactors. A fuel assembly consisting of bulk heat-generating elements, which are integrated together by the system of supply and pressure manifolds, is examined. Spherical heat-generating microelements, which contain nuclear fuel, are pulled into the heat-generating elements. Gaseous coolant proceed from supply manifolds to heat-generating elements, where it withdraws the nuclear reaction heat and assembles in pressure manifolds

  1. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    Directory of Open Access Journals (Sweden)

    Waseem

    2016-01-01

    Full Text Available Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA of Chashma Nuclear Power Plant-1 (CHASNUPP-1 at room temperature in air. The non-linear contact and structural tensile analysis have been performed using ANSYS 13.0, in order to determine the fuel assembly (FA elongation behaviour as well as the location and values of the stress intensity and stresses developed in axial direction under applied tensile load of 9800 N or 2 g being the fuel assembly handling or lifting load [Y. Zhang et al., Fuel assembly design report, SNERDI, China, 1994]. The finite element (FE model comprises spacer grids, fuel rods, flexible contacts between the fuel rods and grid's supports system and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behavior of the geometry is non-linear due to its curvature or design tolerance. It has been observed that fuel assembly elongation values obtained through FE analysis and experiment [SNERDI Tech. Doc., Mechanical strength and calculation for fuel assembly, Technical Report, F3.2.1, China, 1994] under applied tensile load are comparable and show approximately linear behaviors. Therefore, it seems that the permanent elongation of fuel assembly may not occur at the specified load. Moreover, the values of stresses obtained at different locations of the fuel assembly are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Since the results of both studies (analytical and experimental are comparable, therefore, validation of the FE methodology is confirmed. The stress intensity of the FE model and maximum stresses developed along the guide thimbles in axial direction are

  2. Single PWR spent fuel assembly heat transfer data for computer code evaluations

    International Nuclear Information System (INIS)

    The descriptions and results of two separate heat transfer tests designed to investigate the dry storage of commercial PWR spent fuel assemblies are presented. Presented first are descriptions and selected results from the Fuel Temperature Test performed at the Engine Maintenance and Disassembly facility on the Nevada Test Site. An actual spent fuel assembly from the Turkey Point Unit Number 3 Reactor with a decay heat level of 1.17 KW, was installed vertically in a test stand mounted canister/liner assembly. The boundary temperatures were controlled and the canister backfill gases were alternated between air, helium and vacuum to investigate the primary heat transfer mechanisms of convection, conduction and radiation. The assembly temperature profiles were experimentally measured using installed thermocouple instrumentation. Also presented are the results from the Single Assembly Heat Transfer Test designed and fabricated by Allied General Nuclear Services, under contract to the Department of Energy, and ultimately conducted by the Pacific Northwest Laboratory. For this test, an electrically heated 15 x 15 rod assembly was used to model a single PWR spent fuel assembly. The electrically heated model fuel assembly permitted various ''decay heat'', levels to be tested; 1.0 KW and 0.5 KW were used for these tests. The model fuel assembly was positioned within a prototypic fuel tube and in turn placed within a double-walled sealed cask. The complete test assembly could be positioned at any desired orientation (horizontal, vertical, and 250 from horizontal for the present work) and backfilled as desired (air, helium, or vacuum). Tests were run for all combinations of ''decay heat,'' backfill, and orientation. Boundary conditions were imposed by temperature controlled guard heaters installed on the cask exterior surface

  3. Fuel nozzle assembly for use as structural support for a duct structure in a combustor of a gas turbine engine

    Science.gov (United States)

    Wiebe, David J; Fox, Timothy A

    2015-03-31

    A fuel nozzle assembly for use in a combustor apparatus of a gas turbine engine. An outer housing of the fuel nozzle assembly includes an inner volume and provides a direct structural connection between a duct structure and a fuel manifold. The duct structure defines a flow passage for combustion gases flowing within the combustor apparatus. The fuel manifold defines a fuel supply channel therein in fluid communication with a source of fuel. A fuel injector of the fuel nozzle assembly is provided in the inner volume of the outer housing and defines a fuel passage therein. The fuel passage is in fluid communication with the fuel supply channel of the fuel manifold for distributing the fuel from the fuel supply channel into the flow passage of the duct structure.

  4. Proving test on thermal-hydraulic performance of BWR fuel assemblies

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation (NUPEC) has conducted a proving test for thermal-hydraulic performance of BWR fuel (high-burnup 8 x 8, 9 x 9) assemblies entrusted by the Ministry of Economy, Trade and Industry (NUPEC-TH-B Project). The high-burnup 8 x 8 fuel (average fuel assembly discharge burnup: about 39.5 GWd/t), has been utilized from 1991. And the 9 x 9 fuel (average fuel assembly discharge burnup: about 45 GWd/t), has started to be used since 1999. There are two types (A-type and B-type) of fuel design in 9 x 9 fuel assembly. Using an electrically heated test assembly which simulated a BWR fuel bundle on full scale, flow induced vibration, pressure drop, critical power under steady state condition and post-boiling transition (post-BT) tests were carried out in an out-of pile test facility that can simulate the high pressure and high temperature conditions of BWRs. This paper completed the results of 9 x 9 fuel combined with the previously reported results of high-burnup 8 x 8 fuel. As a result of NUPEC-TH-B Project, the validity of the current BWR thermal-hydraulic design method was confirmed and the reliability of BWR thermo-hydraulic fuel performance was demonstrated. Based on the test data, a new correlation of the estimation of fuel rod vibration amplitude, new post-BT heat transfer and rewet correlations for the estimation of fuel rod surface temperature were developed. (author)

  5. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    Energy Technology Data Exchange (ETDEWEB)

    G. W. Eccleston; H. O. Menlove; M. Abhold; M. Baker; J. Pecos

    1999-05-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency {sup 3}He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the {sup 240}Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual.

  6. Nondestructive verification with minimal movement of irradiated light-water-reactor fuel assemblies

    International Nuclear Information System (INIS)

    Nondestructive verification of irradiated light-water reactor fuel assemblies can be performed rapidly and precisely by measuring their gross gamma-ray and neutron signatures. A portable system measured fuel assemblies with exposures ranging from 18.4 to 40.6 GWd/tU and with cooling times ranging from 1575 to 2638 days. Differences in the measured results for side or corner measurements are discussed. 25 figures, 20 tables

  7. Positioning of Nuclear Fuel Assemblies by Means of Image Analysis on Tomographic Data

    OpenAIRE

    Troeng, Mats

    2004-01-01

    A tomographic measurement technique for nuclear fuel assemblies has been developed at the Department of Radiation Sciences at Uppsala University [1]. The technique requires highly accurate information about the position of the measured nuclear fuel assembly relative to the measurement equipment. In experimental campaigns performed earlier, separate positioning measurements have therefore been performed in connection to the tomographic measurements. In this work, another positioning approach h...

  8. A CAREM fuel assembly prototype construction in order to verify its mechanical design using hydrodynamic testing

    International Nuclear Information System (INIS)

    The scope of this paper is to describe the activities of several Groups from three Atomic Centers (C. A. Bariloche, C. A. Ezeiza and C. A. Constituyentes), involved in the manufacturing of a CAREM fuel assembly prototype. The Design Group (UAIN-CAB) carried out the fuel assembly engineering. Cladding components were constructed by the Special Alloys Pilot Factory (UAMCN-CAE). Engineering Group (UACN-CAC) manufactured the parts to be processed, resorting to qualified suppliers. Elastic spacers were completely designed and constructed by this Group, and fuel rods, control rods, guide tubes and spacers were also welded here. Research Reactors Fuels Group (UACN-CAC) carried out the dimensional control of the elaborated parts, while Postirradiation Testing Group (UACN-CAC) performed the assembling of the fuel element. This paper also refers to the design and development of special equipment and devices, all of them required for the prototype construction. (author)

  9. Application of PLUTO Test Facility for U. S. NRC Licensing of a Fuel Assembly

    International Nuclear Information System (INIS)

    The fuel assembly of the PLUS-7 loaded in the APR-1400 follows the same schedule. Meanwhile, In July 1998, the U.S. NRC adopted a research plan to address the effects of high burnup from a Loss of Coolant Accident (LOCA). From these programs, several important technical findings for rule revision were obtained. Based on the technical findings, the U. S. NRC has amended the 10 CFR 50.46 which will be proclaimed sooner or later. Through the amendment, a LOCA analysis on the fuel assembly has to show the safety at both a fresh and End of Life (EOL) state. The U. S. NRC has already required EOL effects on seismic/LOCA performance for a fuel assembly since 1998. To obtain U.S NRC licensing of a fuel assembly, based on the amendment of 10CFR50.46, a LOCA analysis of the fuel assembly has to show safety both fresh and EOL states. The proper damping factor of the fuel assembly measured at the hydraulic test loop for a dynamic model in a LOCA and a seismic analysis code are at least required. In this paper, we have examined the damping technologies and compared the test facility of PLUTO with others in terms of performance. PLUTO has a better performance on the operating conditions than any others

  10. Application of PLUTO Test Facility for U. S. NRC Licensing of a Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dongseok; Shin, Changhwan; Lee, Kanghee; Kang, Heungseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The fuel assembly of the PLUS-7 loaded in the APR-1400 follows the same schedule. Meanwhile, In July 1998, the U.S. NRC adopted a research plan to address the effects of high burnup from a Loss of Coolant Accident (LOCA). From these programs, several important technical findings for rule revision were obtained. Based on the technical findings, the U. S. NRC has amended the 10 CFR 50.46 which will be proclaimed sooner or later. Through the amendment, a LOCA analysis on the fuel assembly has to show the safety at both a fresh and End of Life (EOL) state. The U. S. NRC has already required EOL effects on seismic/LOCA performance for a fuel assembly since 1998. To obtain U.S NRC licensing of a fuel assembly, based on the amendment of 10CFR50.46, a LOCA analysis of the fuel assembly has to show safety both fresh and EOL states. The proper damping factor of the fuel assembly measured at the hydraulic test loop for a dynamic model in a LOCA and a seismic analysis code are at least required. In this paper, we have examined the damping technologies and compared the test facility of PLUTO with others in terms of performance. PLUTO has a better performance on the operating conditions than any others.

  11. 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.

  12. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly.

    Science.gov (United States)

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-04-21

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  13. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Ben De Pauw

    2016-04-01

    Full Text Available Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  14. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly.

    Science.gov (United States)

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

  15. AllianceTM, the fuel assembly on the threshold of the third millennium

    International Nuclear Information System (INIS)

    Nuclear energy competitiveness has become a major challenge of the end of this century. In order to respond, Framatome has developed a new nuclear fuel assembly for an enhanced utilisation of existing PWRs, and also for the future European Pressurized Reactor (EPR). This development is the result of a joint Framatome and Framatome Cogema Fuels (FCF) strategy to propose today a fuel product meeting their customer's future needs. ALLIANCE is a fuel assembly designed for assembly burnup of at least 70 GWd/t. This allows the utilities to consider modes of operation of their reactors which were not technically accessible until now. ALLIANCE is based on an in depth analysis of the market needs of Framatome and Fragema, and also of FCF. ALLIANCE benefits from Framatome long term commitment in a large R and D program, which has provided significant outcomes such as new alloys, new components and new fuel assembly concepts. This R and D program allows direct access to all the CEA expertise and facilities. As a result, ALLIANCE is a fuel assembly with unmatched and totally demonstrated performance. ALLIANCE takes also advantage of the extended operational experience available through all the reactors supplied with Framatome fuel: number of irradiated assemblies reactor types operational conditions - sometimes very demanding. Framatome major customers have been directly involved in the development of this assembly. All the available operational feedback has been taken into account at the early stages of the ALLIANCE design. The main features of the ALLIANCE assembly are: cladding tubes and an assembly structure in M5 alloy, a mono-metallic mixing grid with enhanced thermo-hydraulic performance, the possibility to add mid span mixing grids, the Monobloc guide tube, the Trapper bottom nozzle and a new structure designed for assembly burnups of at least 70 GW d/t. The first ALLIANCE assemblies will be loaded in 1999 in an EdF reactor. In 2000, ALLIANCE assemblies will be loaded

  16. Irradiation experiments of the 6th-12th OGL-1 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio; Minato, Kazuo; Kobayashi, Fumiaki; Kikuchi, Hironobu; Fukuda, Kousaku; Kikuchi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saruta, Tohru; Kitajima, Toshio

    1994-10-01

    The Oarai Gas Loop-1, OGL-1, is an in-pile helium gas loop, installed in the Japan Materials Testing Reactor (JMTR), for irradiation of high-temperature gas-cooled reactor fuels at high pressure and temperature. The present report describes results of fabrication, irradiation and post-irradiation examinations (PIE) of the 6th-12th OGL-1 fuel assemblies. The 6th-8th assemblies used coated fuel particles produced by a small-scale fluidized bed. On the other hand, the 9th-12th assemblies used trial manufacturing fuels, produced with a large-scale fluidized bed for mass production of the fuel for the High Temperature Engineering Test Reactor (HTTR) being constructed. For the 9th assembly loaded with the first mass-product fuel, the fission gas release, R/B of {sup 88}Kr, was relatively high, 1.5x10{sup -5}, and various defects were observed in the ceramography of the irradiated coating layers. Afterwards, a decrease was achieved in the through-coating failure fractions at the fabrication. Correspondingly, the R/B of {sup 88}Kr for the 12th assembly was reduced to an excellent value of 2x10{sup -6}. Thus, the production technology and the irradiation performance of the HTTR design fuels were successfully demonstrated. (author).

  17. Irradiation experiments of the 6th-12th OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    The Oarai Gas Loop-1, OGL-1, is an in-pile helium gas loop, installed in the Japan Materials Testing Reactor (JMTR), for irradiation of high-temperature gas-cooled reactor fuels at high pressure and temperature. The present report describes results of fabrication, irradiation and post-irradiation examinations (PIE) of the 6th-12th OGL-1 fuel assemblies. The 6th-8th assemblies used coated fuel particles produced by a small-scale fluidized bed. On the other hand, the 9th-12th assemblies used trial manufacturing fuels, produced with a large-scale fluidized bed for mass production of the fuel for the High Temperature Engineering Test Reactor (HTTR) being constructed. For the 9th assembly loaded with the first mass-product fuel, the fission gas release, R/B of 88Kr, was relatively high, 1.5x10-5, and various defects were observed in the ceramography of the irradiated coating layers. Afterwards, a decrease was achieved in the through-coating failure fractions at the fabrication. Correspondingly, the R/B of 88Kr for the 12th assembly was reduced to an excellent value of 2x10-6. Thus, the production technology and the irradiation performance of the HTTR design fuels were successfully demonstrated. (author)

  18. Physics Design of Criticality Assembly in Experimental Research About Criticality Safety in Spent Fuel Dissolver

    Institute of Scientific and Technical Information of China (English)

    ZHOU; Qi

    2012-01-01

    <正>In order to meet the experimental demand of criticality safety research in the spent fuel dissolver, we need to design a suitable criticality assembly. The key problem of the design work is the core design because there are many limits for it such as the number of fuel rods loaded, fissile materials existed in the solution, reactivity control, core size and etc.

  19. SICOM, An equipment for very accurate dimensional and corrosion inspection of irradiated fuel assemblies

    International Nuclear Information System (INIS)

    Nuclear fuel undergo evolution in its properties and characteristics as result of the operation in the nuclear reactors. Knowledge of this evolution is necessary to guaranty the behaviour and safety in the operation, as well as to optimize the economic performance of the fuel cycle. When, with the objective of improving the performance of fuel, some important modifications are made, it is necessary to prove the effect on the behaviour and evaluate the influence in each of the critical characteristics of the fuel assemblies. In this cases, it could be necessary to burn a reduced number of fuel assemblies of the new design (demonstration assemblies). During the demonstration process, some characterization of the behaviour of the fuel is made, usually at the end of each cycle. An important part of this characterization is made using NDE methods, applied at the nuclear plant during the refuelling outages. Within the Electrotechnical Research and Development Program (PIE), and with participation by IBERDROLA, TECNATOM and ENUSA, an inspection system (SICOM) has been developed for spent fuel assemblies from pressurized water plants, the aim being to check the following features: general condition, apply dimensional controls and measure the oxide layer on the peripheral fuel rods. This equipment was qualified at Tecnatom and Almaraz I NPP during the first quarter of 1995. Subsequently, in September, it was validated for the EDF P'4 plants at C.N.P.E. Belleville

  20. Analysis of Spent Fuel Assembly Thermal Behaviors in Boil-off Accident Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hye-Min; Chun, Tae-Hyun; Kim, Sun-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The spent fuel pool (SFP) accidents would occur due to many different postulated scenarios, for example a SBO (Station Black Out) at SFP storage or an attack from external factor. In this study, we focused on the SFP boil off accident and analyzed the thermal behaviors of spent fuels following this accident, using MELCOR 1.8.6. version. MELCOR, originally the severe accident code, has been developed to also be appropriate to the SFP accident. This paper provides the spent fuel heatup characteristics in terms of decay heat, water level and fuel arrangement. The SFP model is based on 17x17 PWR assembly designed by Westinghouse. Spent fuel coolability has been analyzed with single and 1x4 assembly MELCOR models in the case of boil-off accident. It was shown that the low powered spent fuel assembly could be more vulnerable in the partial loss of coolant inventory because of lack of steam cooling and more fuel being uncovered. In addition, it was found that minimum water level has to be maintained above half of assembly height so as not to experience fuel failure, which depends on decay heat power.

  1. End-of-irradiation data report for the instrumented fuel assembly (IFA)-527

    International Nuclear Information System (INIS)

    This report presents data obtained during the irradiation of the six-rod instrumented fuel assembly (IFA)-527 in the Halden Boiling Water Reactor (HBWR), Halden, Norway. This assembly is the last in a series of US Nuclear Regulatory Commission (NRC)-sponsored tests to obtain data for the development and verification of steady-state fuel performance computer codes. IFA-527 contains five identical rods with high-density stable fuel pellets and 230-μm diametral gaps and one rod with similar fuel pellets but with a 60-μm diametral gap. All six rods were xenon-filled to simulate the effects of fission gas and to enhance the observable effects of fuel cracking and relocation on fuel temperatures. This report presents both pre- and postfailure data for IFA-527

  2. Determination of the Effectiveness of Control Rods in the VVER Reactor Fuel Assemblies

    International Nuclear Information System (INIS)

    The paper describes experiments done in homogeneous mock-ups of the fuel assemblies from the VVER Reactor (at one level of enrichment) to determine the effectiveness of absorbing systems comprising shim fuel assemblies or water cavities and of absorbing rods clad in jackets made of differing materials. The paper also gives data on some experiments that have been done in mock-ups of assemblies with differing levels of enrichment. These experiments make it possible to verify the methods used in calculation and to evaluate the prospects of using them for heterogeneous reactors. (author)

  3. Verification of 235U enrichment of fresh VVER-440 fuel assemblies.

    Science.gov (United States)

    Almási, I; Nguyen, C T; Zsigrai, J; Lakosi, L; Hlavathy, Z; Nagy, P; Buglyó, N

    2012-10-01

    Enrichment of uniformly and non-uniformly enriched ("profiled") fuel assemblies in a range of 1.6-4.4% was verified by gamma-ray spectrometry at a nuclear power plant (NPP). HPGe detectors and a CdZnTe (CZT) detector, the latter fitting into the central tube of the assemblies, were used for obtaining information from outer and inner fuel rods. A procedure which has minimal impact on the NPP work was developed for verifying freshly arrived assemblies under normal operational conditions, and is now in routine use.

  4. Welding device for nuclear fuel assembly structure elements

    International Nuclear Information System (INIS)

    The device has two parallel assembling positions next to each other. The welding robot is carried by a carriage with displacement parallel to the guide tubes and has enough degrees of freedom to move from one assembling position to the other and have access to the structural elements

  5. Fuel assembly mechanical behavior with respect to center guide tube welding points

    International Nuclear Information System (INIS)

    A fuel assembly with modified welding procedure between the center guide tube and sleeves on grid structures is proposed in the work, and some mechanical behavior of the fuel assembly is presented. While all sleeves on grid assemblies are welded to center guide tube in the present design, the modified fuel assembly with reduced welding is proposed. To evaluate mechanical performance of the design, lateral bending deflection test and lateral vibration test were performed in a test facility. Applying a lateral static load at 6-th grid, the fuel assembly deflection at each grid was measured in the lateral static test. In the vibration test, displacement signal at every grid is measured while the fuel assembly is excited by a dynamic shaker at the 6-th grid. Processing the measured signals, modal properties such as natural frequency, damping, and mode shapes can be found based on frequency response functions. The test results, comparing to a previous design with all girds welding to the center guide tube, do not show any notable deviation in the structural behaviors

  6. Transport of fresh MOX fuel assemblies for the MONJU initial core

    Energy Technology Data Exchange (ETDEWEB)

    Miura, Y.; Ouchi, Y.; Kurakami, J. [Power Reactor and Nuclear Fuel Development Corp., Tokai Works, Naka, Ibaraki (Japan); Usami, M. [Power Reactor and Nuclear Fuel Development Corp., Head Office Sankaido Building, Minato, Tokyo (Japan)

    1997-12-31

    Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies (109 assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this packaging design features such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (Author).

  7. A US perspective on fast reactor fuel fabrication technology and experience part I: metal fuels and assembly design

    International Nuclear Information System (INIS)

    This paper is part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF). Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated in a considerable amount of research that resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

  8. Apparatus for removing and/or positioning fuel assemblies of a nuclear reactor

    International Nuclear Information System (INIS)

    Apparatus for positioning fuel assemblies of a nuclear reactor includes a control for a crane comprising a strain gauge connected to the crane line which raises and lowers the load. The signal from the strain gauge is compared with setpoints; which if the strain gauge signal exceeds a high-level setpoint, indicating that the movement of a fuel assembly is obstructed, the line drive is disabled. The line drive is also disabled if the strain gauge signal is less than a low-level setpoint, indicating that a fuel being deposited contacts the bottom of its slot or an obstruction. To preclude lateral movement of the fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge signal exceeds the low-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than a slack-line setpoint. (author)

  9. Application of the ballooning analysis code MATARE on a generic PWR fuel assembly

    International Nuclear Information System (INIS)

    The MATARE (MAbel-TAlink-RElap) code is a new multi-pin deformation analysis code created through the dynamic coupling between the thermal-hydraulic code RELAP5 and multiple instances of the single-pin thermal-mechanics code MABEL. A multi-pin representation of different zones of a typical PWR fuel assembly under post-LOCA reflooding conditions was analysed including some of the most relevant features that characterise a typical nuclear reactor fuel assembly and evaluate their effect on the behaviour of the fuel rods under conditions leading to clad ballooning. The code was able to simulate the deformation of wide regions of a fuel assembly under reflood conditions and has shown how differences in pin pressure and the presence of rod with burnable poisons and control rod guide thimbles also contribute to a substantial incoherent ballooning in agreement with the experimental data. (author)

  10. A computational technique to identify the optimal stiffness matrix for a discrete nuclear fuel assembly model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Nam-Gyu, E-mail: nkpark@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Kyoung-Joo, E-mail: kyoungjoo@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Kyoung-Hong, E-mail: kyounghong@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Suh, Jung-Min, E-mail: jmsuh@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2013-02-15

    Highlights: ► An identification method of the optimal stiffness matrix for a fuel assembly structure is discussed. ► The least squares optimization method is introduced, and a closed form solution of the problem is derived. ► The method can be expanded to the system with the limited number of modes. ► Identification error due to the perturbed mode shape matrix is analyzed. ► Verification examples show that the proposed procedure leads to a reliable solution. -- Abstract: A reactor core structural model which is used to evaluate the structural integrity of the core contains nuclear fuel assembly models. Since the reactor core consists of many nuclear fuel assemblies, the use of a refined fuel assembly model leads to a considerable amount of computing time for performing nonlinear analyses such as the prediction of seismic induced vibration behaviors. The computational time could be reduced by replacing the detailed fuel assembly model with a simplified model that has fewer degrees of freedom, but the dynamic characteristics of the detailed model must be maintained in the simplified model. Such a model based on an optimal design method is proposed in this paper. That is, when a mass matrix and a mode shape matrix are given, the optimal stiffness matrix of a discrete fuel assembly model can be estimated by applying the least squares minimization method. The verification of the method is completed by comparing test results and simulation results. This paper shows that the simplified model's dynamic behaviors are quite similar to experimental results and that the suggested method is suitable for identifying reliable mathematical model for fuel assemblies.

  11. High-level neutron-coincidence-counter (HLNCC) implementation: assay of the plutonium content of mixed-oxide fuel assemblies

    International Nuclear Information System (INIS)

    The portable High-Level Neutron Coincidence Counter is used to assay the 240Pu-effective loading of a reference mixed-oxide fuel assembly by neutron coincidence counting. We have investigated the effects on the coincidence count rate of the total fuel loading (UO2 + PuO2), the fissile loading, the fuel rod diameter, and the fuel rod pattern. The coincidence count rate per gram of 240Pu-effective per centimeter is primarily dependent on the total fuel loading of the assembly; the higher the loading, the higher the coincidence count rate. Detailed procedures for the assay of mixed-oxide fuel assemblies are developed

  12. Summary of Booster Development and Qualification Report

    Energy Technology Data Exchange (ETDEWEB)

    Francois, Elizabeth G. [Los Alamos National Laboratory; Harry, Herbert H. [Los Alamos National Laboratory; Hartline, Ernest L. [Los Alamos National Laboratory; Hooks, Daniel E. [Los Alamos National Laboratory; Johnson, Carl E. [Los Alamos National Laboratory; Morris, John S. [Los Alamos National Laboratory; Novak, Alan M. [Los Alamos National Laboratory; Ramos, Kyle J. [Los Alamos National Laboratory; Sanders, Victor E. [Los Alamos National Laboratory; Scovel, Christina A. [Los Alamos National Laboratory; Lorenz, Thomas [LLNL; Wright, Mark [AWE; Botcher, Tod [PANTEX; Marx, Erin [NSWC-IHDIV; Gibson, Kevin [NSWC-IHDIV

    2012-06-21

    This report outlines booster development work done at Los Alamos National Laboratory from 2007 to present. The booster is a critical link in the initiation train of explosive assemblies, from complex devices like nuclear weapons to conventional munitions. The booster bridges the gap from a small, relatively sensitive detonator to an insensitive, but massive, main charge. The movement throughout the explosives development community is to use more and more insensitive explosive components. With that, more energy is needed out of the booster. It has to initiate reliably, promptly, powerfully and safely. This report is divided into four sections. The first provides a summary of a collaborative effort between LANL, LLNL, and AWE to identify candidate materials and uniformly develop a testing plan for new boosters. Important parameters and the tests required to measure them were defined. The nature of the collaboration and the specific goals of the participating partners has changed over time, but the booster development plan stands on its own merit as a complete description of the test protocol necessary to compare and qualify booster materials, and is discussed in its entirety in this report. The second section describes a project, which began in 2009 with the Department of Defense to develop replacement booster formulations for PBXN-7. Replacement of PBXN-7 was necessary because it contained Triaminotrinitrobenzene (TATB), which was becoming unavailable to the DoD and because it contained Cyclotrimethylenetrinitramine (RDX), which was sensitive and toxic. A LANL-developed explosive, Diaminoazoxyfurazan (DAAF), was an important candidate. This project required any replacement formulation be a drop-in replacement in existing munitions. This project was timely, in that it made use of the collaborative booster development project, and had the additional constraint of matching shock sensitivity. Additionally it needed to be a safety improvement, and a performance

  13. Validation of PWR core seismic models with shaking table tests on interacting scale 1 fuel assemblies

    International Nuclear Information System (INIS)

    The fuel assembly mechanical strength must be justified with respect to the lateral loads under accident conditions, in particular seismic loads. This justification is performed by means of time-history analyses with dynamic models of an assembly row in the core, allowing for assembly deformations, impacts at grid locations and reactor coolant effects. Due to necessary simplifications, the models include 'equivalent' parameters adjusted with respect to dynamic characterisation tests of the fuel assemblies. Complementing such tests on isolated assemblies by an overall model validation with shaking table tests on interacting assemblies is obviously desirable. Seismic tests have been performed by French CEA (Commissariat a l'Energie Atomique) on a row of six full scale fuel assemblies, including two types of 17 x 17 12ft design. The row models are built according to the usual procedure, with preliminary characterisation tests performed on a single assembly. The test-calculation comparisons are made for two test configurations : in air and in water. The relatively large number of accelerograms (15, used for each configuration) is also favourable to significant comparisons. The results are presented for the impact forces at row ends, displacements at mid assembly, and also 'statistical' parameters. Despite a non-negligible scattering in the results obtained with different accelerograms, the calculations prove realistic, and the modelling process is validated with a good confidence level. This satisfactory validation allows to evaluate precisely the margins in the seismic design methodology of the fuel assemblies, and thus to confirm the safety of the plants in case of seismic event. (author)

  14. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    International Nuclear Information System (INIS)

    A fuel assembly operated in a nuclear power plant typically contains 100–300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative

  15. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Jacobsson Svärd, Staffan, E-mail: staffan.jacobsson_svard@physics.uu.se; Holcombe, Scott; Grape, Sophie

    2015-05-21

    A fuel assembly operated in a nuclear power plant typically contains 100–300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative

  16. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    Science.gov (United States)

    Jacobsson Svärd, Staffan; Holcombe, Scott; Grape, Sophie

    2015-05-01

    A fuel assembly operated in a nuclear power plant typically contains 100-300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative assessment, which

  17. 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

  18. Simulation of Spent PWR Fuel Assembly Behavior Under Normal Conditions of Transport

    International Nuclear Information System (INIS)

    The behavior of a PWR high-burnup spent fuel assembly under normal conditions of transport is simulated in a dynamic analysis of a 0.3-m free drop of a transportation cask unprotected by impact limiters striking a flat rigid surface in the horizontal orientation. The structural analysis employs a finite element numerical model consisting of the cask, the fuel assemblies, the fuel rods, the guide tubes and the cask’s internal structures that hold the fuel assemblies in position. Appropriate mechanical properties for the cask’s structural components, as well as the elastic-plastic properties typical of high-burnup Zircaloy-4 cladding, are utilized. Emphasis is placed on fuel rods responses at locations where maximum forces would be expected, which include end-plate positions and spacer-grid positions at assembly mid-span. Temporal and spatial variations of the forces acting on the fuel rods are calculated and post-processed to obtain frequency distributions, which statistically represent the total fuel rod population in the cask. The results show that the largest pinch force, (ror-to-rod contact force), is 1700 lb, the maximum axial force is 600 lb, and the largest bending moment is 175 in-lb. Failure analysis of fuel rods using these force quantities, and considering the effects of potential hydrides reorientation on cladding failure resistance, indicates, under conservative assumptions, a factor of safety of least 2 against longitudinal tearing, and no failure is predicted for transverse tearing or rod breakage. Fuel reconfiguration is predicted not to occur, and although partial tearing of guide tubes is possible, it is not enough to impair post-accident assembly retrieval. (author)

  19. Transport of fresh MOX fuel assemblies for the Monju initial core

    Energy Technology Data Exchange (ETDEWEB)

    Kurakami, J.; Ouchi, Y. [Power Reactor and Nuclear Fuel Development Corp., Tokai Works (Japan); Usami, M. [Power Reactor and Nuclear Fuel Development Corp., Tokyo Head Office, Power Reactor and Nuclear Fuel Development Corp., Ibaraki (Japan)

    1997-12-01

    Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this package design feature such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying of the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (Author).

  20. Numerical investigation on the characteristics of two-phase flow in fuel assemblies with spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Chen, D.; Yang, Z.; Zhong, Y.; Xiao, Y.; Hu, L. [Chongqing Univ. (China). Key Lab. of Low-grade Energy Utilization Technologies and Systems

    2016-07-15

    In pressurized water reactors (PWRs), the spacer grids of the fuel assembly has significant impact on the thermal-hydraulic performance of the fuel assembly. Particularly, the spacer grids with the mixing vanes can dramatically enhance the secondary flow and have significant effect on the void distribution in the fuel assembly. In this paper, the CFD study has been carried out to analyze the effects of the spacer grid with the steel contacts, dimples and mixing vanes on the boiling two-phase flow characteristics, such as the two-phase flow field, the void distribution, and so on. Considered the influence of the boiling phase change on two-phase flow, a boiling model was proposed and applied in the CFD simulation by using the UDF (User Defined Function) method. Furthermore, in order to analyze the effects of the spacer grid with mixing vanes, the adiabatic (without boiling) two-phase flow has also been investigated as comparison with the boiling two-phase flow in the fuel assembly with spacer grids. The CFD simulation on two-phase flow in the fuel assembly with the proposed boiling model can predict the characteristics of two-phase flow better.

  1. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  2. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO2 and UO2), typically containing 95% or more UO2. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO2 powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO2 powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required

  3. AREVA's fuel assemblies addressing high performance requirements of the worldwide PWR fleet

    International Nuclear Information System (INIS)

    Taking advantage of its presence in the fuel activities since the start of commercial nuclear worldwide operation, AREVA is continuing to support the customers with the priority on reliability, to: >participate in plant operational performance for the in core fuel reliability, the Zero Tolerance for Failure ZTF as a continuous improvement target and the minimisation of manufacturing/quality troubles, >guarantee the supply chain a proven product stability and continuous availability, >support performance improvements with proven design and technology for fuel management updating and cycle cost optimization, >support licensing assessments for fuel assembly and reloads, data/methodologies/services, >meet regulatory challenges regarding new phenomena, addressing emergent performance issues and emerging industry challenges for changing operating regimes. This capacity is based on supplies by AREVA accumulating very large experience both in manufacturing and in plant operation, which is demonstrated by: >manufacturing location in 4 countries including 9 fuel factories in USA, Germany, Belgium and France. Up to now about 120,000 fuel assemblies and 8,000 RCCA have been released to PWR nuclear countries, from AREVA European factories, >irradiation performed or in progress in about half of PWR world wide nuclear plants. Our optimum performances cover rod burn ups of to 82GWD/tU and fuel assemblies successfully operated under various world wide fuel management types. AREVA's experience, which is the largest in the world, has the extensive support of the well known fuel components such as the M5'TM'cladding, the MONOBLOC'TM'guide tube, the HTP'TM' and HMP'TM' structure components and the comprehensive services brought in engineering, irradiation and post irradiation fields. All of AREVA's fuel knowledge is devoted to extend the definition of fuel reliability to cover the whole scope of fuel vendor support. Our Top Reliability and Quality provide customers with continuous

  4. The optimization of an AP1000 fuel assembly for the transmutation of plutonium and minor actinides

    Science.gov (United States)

    Washington, Jeremy A.

    The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. The goal of this thesis is to examine the potential of light water reactors for plutonium and minor actinides transmutation as a near-term solution. This thesis screens the available nuclear isotope database to identify potential absorbers as coatings on a transmutation fuel in a light water reactor. A spectral shift absorber coating tunes the neutron energy spectrum experienced by the underlying target fuel. Eleven different spectral shift absorbers (B4C, CdO, Dy2O3, Er 2O3, Eu2O3, Gd2O3, HfO2, In2O3, Lu2O3, Sm2O3, and TaC) have been selected for further evaluation. A model developed using the NEWT module of SCALE 6.1 code provided performance data for the burnup of the target fuel rods. Irradiation of the target fuels occurs in a Westinghouse 17x17 XL Robust Fuel Assembly over a 1400 Effective Full Power Days (EFPD) interval. The fuels evaluated in this thesis include PuO2, Pu3Si2, PuN, MOX, PuZrH, PuZrHTh, PuZrO 2, and PuUZrH. MOX (5 wt% PuO2), Pu0.31ZrH 1.6Th1.08, and PuZrO2MgO (8 wt%) are selected for detailed analysis in a multi-pin transmutation assembly. A coupled model optimized the resulting transmutation fuel elements. The optimization considered three stages of fuel assemblies containing target fuel pins. The first stage optimized four target fuel pins adjacent to the central instrumentation channel. The second stage evaluated a variety of assemblies with multiple target fuel pins and the third stage re-optimized target fuel pins in the second-stage assembly. A PuZrO2MgO (8 wt%) target fuel with a coating of Lu 2O3 resulted in the greatest reduction in curium-244

  5. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  6. Method and jig for dismantling nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Urata, Megumi; Watahiki, Minoru.

    1989-08-30

    The object of the present inention is to extract a fuel element from a lower tie plate safely and at high efficiency by a remote control operation. That is, a forked top end of a lever of a dismantling jig is inserted between the tapered portion of a lower end plug and a lower tie plate. Then, a load is applied to the counter-lower end side of the lever by a motor. This exerts an elevating force to the fuel elements to easily release fixture between the lower end plug and the lower tie plate. Since the fuel can of fuel elements is not applied with a force by this mehtod, operation safety can be improved. (I.J.).

  7. Integrated Three-Voltage-Booster DC-DC Converter to Achieve High Voltage Gain with Leakage-Energy Recycling for PV or Fuel-Cell Power Systems

    Directory of Open Access Journals (Sweden)

    Chih-Lung Shen

    2015-09-01

    Full Text Available In this paper, an integrated three-voltage-booster DC-DC (direct current to direct current converter is proposed to achieve high voltage gain for renewable-energy generation systems. The proposed converter integrates three voltage-boosters into one power stage, which is composed of an active switch, a coupled-inductor, five diodes, and five capacitors. As compared with conventional high step-up converters, it has a lower component count. In addition, the features of leakage-energy recycling and switching loss reduction can be accomplished for conversion efficiency improvement. While the active switch is turned off, the converter can inherently clamp the voltage across power switch and suppress voltage spikes. Moreover, the reverse-recovery currents of all diodes can be alleviated by leakage inductance. A 200 W prototype operating at 100 kHz switching frequency with 36 V input and 400 V output is implemented to verify the theoretical analysis and to demonstrate the feasibility of the proposed high step-up DC-DC converter.

  8. COXPRO-II: a computer program for calculating radiation and conduction heat transfer in irradiated fuel assemblies

    International Nuclear Information System (INIS)

    This report describes the computer program COXPRO-II, which was written for performing thermal analyses of irradiated fuel assemblies in a gaseous environment with no forced cooling. The heat transfer modes within the fuel pin bundle are radiation exchange among fuel pin surfaces and conduction by the stagnant gas. The array of parallel cylindrical fuel pins may be enclosed by a metal wrapper or shroud. Heat is dissipated from the outer surface of the fuel pin assembly by radiation and convection. Both equilateral triangle and square fuel pin arrays can be analyzed. Steady-state and unsteady-state conditions are included. Temperatures predicted by the COXPRO-II code have been validated by comparing them with experimental measurements. Temperature predictions compare favorably to temperature measurements in pressurized water reactor (PWR) and liquid-metal fast breeder reactor (LMFBR) simulated, electrically heated fuel assemblies. Also, temperature comparisons are made on an actual irradiated Fast-Flux Test Facility (FFTF) LMFBR fuel assembly

  9. Welding machine and welding process for nuclear fuel assembly structures

    International Nuclear Information System (INIS)

    The welding device comprises a mounting jig which receives the guide tubes and the assembly supporting structures in the desired spatial orientation. It also comprises a welding head which can travel on rails along the length of the guide tubes and has at least a welding spring chuck movable in two axes and rotatable relative to the welding machine; the spring chuck can pass between two adjacent tube rows and takes a tubes where a weld is necessary. The welding spring chuck can apply spot-welding pulses. This is used for the assembly of guide tubes and bundles for water-cooled nuclear reactors

  10. Development of prediction method of void fraction distribution in fuel assemblies for use in safety analysis

    International Nuclear Information System (INIS)

    The establishment of code system for BWR safety analysis is now in progress at Institute of Nuclear Safety (INS), in order to predict the onset of boiling transition (BT) in nuclear fuel assemblies in any thermal-hydraulic condition without relying on the thermal-hydraulic characteristic data provided by licensee. The prediction method for void fraction distribution across cross section of BWR fuel assemblies has been developed based on multi-dimensional two-fluid model. Lift forces working on bubbles and void diffusion that can not be handled with one-dimensional analysis were considered. Comparisons between calculated results and experimental data obtained from thermal-hydraulic tests of PWR and BWR mock-up fuel assemblies showed good agreement. Lift force models have been empirical and further studies were needed, but the calculations showed the possibility of applying these models to multi-dimensional gas-liquid two-phase flow analysis. (author)

  11. Use of high energy gamma emission tomography for partial defect verification of spent fuel assemblies

    International Nuclear Information System (INIS)

    The possibility to use passive gamma emission tomography for revealing non-destructively the rod structure of spent BWR fuel assemblies has been studied in cooperation with the Finnish Support Programme to the IAEA Safeguards (task FIN A98) and the Technical University of Budabest in Hungary. The ultimate goal is to develop partial verification methods for verification of spent nuclear fuel. The task included experimental measurements of irradiated BWR assemblies using underwater measurement techniques together with computer analysis of the measured data as well as computer simulation of tomographic measurements. The results obtained show that rod-level partial defect verification of spent LWR fuel assemblies is feasible using computed gamma emission tomography. This report describes the results of this project. (orig.). (7 refs., 29 figs., 2 tabs.)

  12. HEXBU-3D, a three-dimensional PWR-simulator program for hexagonal fuel assemblies

    International Nuclear Information System (INIS)

    HEXBU-3D is a three-dimensional nodal simulator program for PWR reactors. It is designed for a reactor core that consists of hexagonal fuel assemblies and of big follower-type control assemblies. The program solves two-group diffusion equations in homogenized fuel assembly geometry by a sophisticated nodal method. The treatment of feedback effects from xenon-poisoning, fuel temperature, moderator temperature and density and soluble boron concentration are included in the program. The nodal equations are solved by a fast two-level iteration technique and the eigenvalue can be either the effective multiplication factor or the boron concentration of the moderator. Burnup calculations are performed by tabulated sets of burnup-dependent cross sections evaluated by a cell burnup program. HEXBY-3D has been originally programmed in FORTRAN V for the UNIVAC 1108 computer, but there is also another version which is operable on the CDC CYBER 170 computer. (author)

  13. Design and performance verification of fuel assembly and steam generator simulators for SMART reactor

    International Nuclear Information System (INIS)

    The SMART reactor has been developed at KAERI, for the generation of electric power and also for seawater desalination. In order to verify the performance of the SMART design with respect to flow and pressure distribution, an experimental test facility named SCOP has been developed. For the purpose of preserving the flow distribution characteristics, SCOP is linearly reduced with a scaling ratio of 1/5. A CFD analysis was carried out to draw basic design parameters of the venturi tube and the perforated plates in a fuel assembly simulator. A CALIP, which is a flow and pressure drop calibration test facility, has been constructed to evaluate the pressure drop characteristic of fuel assembly and steam generator simulators. This paper shows the results of the actual performance verification and evaluation of fuel assembly and steam generator simulator, were evaluated using a CALIP. (author)

  14. Experience in generalizing the data on the rod fuel assembly burnout by the method of cells

    International Nuclear Information System (INIS)

    The comparative analysis of results of calculations on the rod fuel assembly burnout by the method of cells is conducted. The method of cells is based on determining local parameters of coolant flow in the assembly cross section. The channel cross section is conditionally subdivided into elementary cells communicating with each other. For such system of interacting cells the equations of thermohydraulics are solved. Local values of enthalpy and mass rate obtained are used for fuel burnout calculation. The analysis performed has shown that generalizing the empirical data for a given set of experiments on the basis of local parameters of flow in the cells already nowadays assures in most cases accuracy of burnout conditions prediction which is not less than the traditional dependences based on one-dimensional description of the flow. The conclusion is drawn on the prospects of using the method of cells for calculating the burnout in rod fuel assemblies and necessity of its subsequent development

  15. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    International Nuclear Information System (INIS)

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  16. Development of mechanical test techniques for structural components of irradiated PWR fuel assembly

    International Nuclear Information System (INIS)

    An increase of fuel burnup and duration of fuel life remains one of the main methods for a nuclear power engineering enhancement. Properties of structural materials providing corrosion resistance, mechanical strength, and dimensional instability of the components of a fuel assembly (FA) are of great importance for fuel operational reliability in such fuel life cycles. Generally, PWR fuel assemblies consist of a top nozzle, spacer grid, bottom nozzle, and guide/instrumentation tubes. The top and bottom nozzle are fixed to the guide tubes using a screw or bulge method. The spacer grid fixed to the guide/instrumentation tubes using a spot weld or bulge method. To understand the in-reactor performance of PWR FA, several devices and test techniques have been developed for mechanical property tests. Among the structural components of PWR FA, a spacer grid, a hold down spring of a top nozzle and a connecting part of FA were considered. Experimental works were carried out for the unirradiated and irradiated components of advanced nuclear fuel assemblies for KSNPs and Westinghouse type PWRs at IMEF (Irradiated Materials Examination Facility) at KAERI. The developed techniques were verified through a hot cell tests. (author)

  17. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  18. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO2 powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO2 powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule

  19. Acceptance of failed SNF [spent nuclear fuel] assemblies by the Federal Waste Management System

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. This document discusses acceptance of failed spent fuel assemblies by the Federal Waste Management System. 18 refs., 7 figs., 25 tabs

  20. 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)

  1. Optimization of fuel rod enrichment distribution to minimize rod power peaking throughout life within BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1997-01-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. a combinatorial optimization problem grouping fuel rods into a given number of rod groups with the same enrichment, and a problem determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by a linear combination: C{sub 1}X+C{sub 2}X{sub G}, where X and X{sub G} stand for a control variable to give the constraint respectively for a local power peaking factor and a gadolinium rod power, and C{sub 1} and C{sub 2} are user-definable weighting factor to accommodate the design preference. The algorithm of solving the combinatorial optimization problem starts with finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering, and then the latter problem is solved by using the method of approximation programming. The practical application of the present method is shown for a contemporary 8x8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  2. Process for recycling components of a PEM fuel cell membrane electrode assembly

    Science.gov (United States)

    Shore, Lawrence

    2012-02-28

    The membrane electrode assembly (MEA) of a PEM fuel cell can be recycled by contacting the MEA with a lower alkyl alcohol solvent which separates the membrane from the anode and cathode layers of the assembly. The resulting solution containing both the polymer membrane and supported noble metal catalysts can be heated under mild conditions to disperse the polymer membrane as particles and the supported noble metal catalysts and polymer membrane particles separated by known filtration means.

  3. Study on seismic response characteristics of reactor vessel internals and fuel assembly for OBE elimination

    International Nuclear Information System (INIS)

    To resolve a general argument about OBE elimination for the future nuclear power plant design, seismic responses of reactor vessel internals and fuel assembly for Ulchin nuclear power plant units 3 and 4 in Korea are investigated as an example. Dynamic analyses of the coupled internals and core are performed for the seismic excitations using the reactor vessel motions. By investigating the response relations between OBE and SSE and their response characteristics, the critical components for OBE loading are addressed. Also the fuel assembly responses are calculated using the core plate motions and their behavior is found to be insignificant for OBE elimination. (author)

  4. Dynamics of nuclear fuel assemblies in vertical flow channels: computer modelling and associated studies

    International Nuclear Information System (INIS)

    A computer model, designed to predict the dynamic behaviour of nuclear fuel assemblies in axial flow, is described in this report. The numerical methods used to construct and solve the matrix equations of motion in the model are discussed together with an outline of the method used to interpret the fuel assembly stability data. The mathematics developed for forced response calculations are described in detail. Certain structural and hydrodynamic modelling parameters must be determined by experiment. These parameters are identified and the methods used for their evaluation are briefly described. Examples of typical applications of the dynamic model are presented towards the end of the report. (author)

  5. Heat transfer analysis of fuel assemblies in a heterogeneous gas core nuclear rocket

    Science.gov (United States)

    Watanabe, Yoichi; Appelbaum, Jacob; Diaz, Nils; Maya, Isaac

    1991-01-01

    Heat transfer problems of a heterogeneous gaseous core nuclear rocket were studied. The reactor core consists of 1.5-m long hexagonal fuel assemblies filled with pressurized uranium tetrafluoride (UF4) gas. The fuel gas temperature ranges from 3500 to 7000 K at a nominal operating condition of 40 atm. Each fuel assembly has seven coolant tubes, through which hydrogen propellant flows. The propellant temperature is not constrained by the fuel temperature but by the maximum temperature of the graphite coolant tube. For a core achieving a fission power density of 1000 MW/cu m, the propellant core exit temperature can be as high as 3200 K. The physical size of a 1250 MW gaseous core nuclear rocket is comparable with that of a NERVA-type solid core nuclear rocket. The engine can deliver a specific impulse of 1020 seconds and a thrust of 330 kN.

  6. Booster parameter list

    Energy Technology Data Exchange (ETDEWEB)

    Parsa, Z.

    1986-10-01

    The AGS Booster is designed to be an intermediate synchrotron injector for the AGS, capable of accelerating protons from 200 MeV to 1.5 GeV. The parameters listed include beam and operational parameters and lattice parameters, as well as parameters pertaining to the accelerator's magnets, vacuum system, radio frequency acceleration system, and the tunnel. 60 refs., 41 figs. (LEW)

  7. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, P. III

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  8. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    International Nuclear Information System (INIS)

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% totalPu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products

  9. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.; Bounds, J.A.; Kimpland, R.H.; Damjanovich, R.P.; Jaegers, P.J.

    1997-08-01

    Experiments were performed to measure a variety of parameters for SHEBA: behavior of the facility during transient and steady-state operation; characteristics of the SHEBA fuel; delayed-critical solution height vs solution temperature; initial reactor period and reactivity vs solution height; calibration of power level vs reactor power instrumentation readings; flux profile in SHEBA; radiation levels and neutron spectra outside the assembly for code verification and criticality alarm and dosimetry purposes; and effect on reactivity of voids in the fuel.

  10. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

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

    International Nuclear Information System (INIS)

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

  12. Report of lower endplug welding, and testing and inspecting result for MONJU 1th reload core fuel assembly

    International Nuclear Information System (INIS)

    The procedure and result of lower endplugwelding, Test and Inspection and Shipment of the 1th reload core fuel assembly (80 Fuel Assemblies) for the fast breeder reactor MONJU are reported, which had been examined and inspected in Tamatsukuri Branch, Material Insurance Office, Quality Assurance Section, Technical Administration Division, Plutonium Fuel Center (before: Inspection Section, Plutonium Fuel Division), from June 1994 to January 1996. The number of cladding tubes welded to the endplug were totally 13,804: 7,418 for Core - Inside of 43 fuel Assemblies and 6,836 for Core-Outside of 37 fuel Assemblies. 13,794 of them, 7,414 Core-Inside and 6,379 Core-Outside, were approved by the test and sent to Plutonium Fuel Center. 10 of them weren't approved mainly because of default welding. Disapproval rating was 0.07%. (author)

  13. A Unique Hybrid Propulsion System Design for Large Space Boosters

    Science.gov (United States)

    Rodgers, Frederick C.

    1990-01-01

    A study was made of the application of hybrid rocket propulsion technology to large space boosters. Safety, reliability, cost, and performance comprised the evaluation criteria, in order of relative importance, for this study. The effort considered the so called classic hybrid design approach versus a novel approach which utilizes a fuel-rich gas generator for the fuel source. Other trades included various fuel/oxidizer combinations, pressure-fed versus pump fed oxidizer delivery systems, and reusable versus expandable booster systems. Following this initial trade study, a point design was generated. A gas generated-type fuel grain with pump fed liquid oxygen comprised the basis of this point design. This design study provided a mechanism for considering the means of implementing the gas generator approach for further defining details of the design. Subsequently, a system trade study was performed which determined the sensitivity of the design to various design parameters and predicted optimum values for these same parameters. The study concluded that a gas generator hybrid booster design offers enhanced safety and reliability over current of proposed solid booster designs while providing equal or greater performance levels. These improvements can be accomplished at considerably lower cost than for the liquid booster designs of equivalent capability.

  14. Fuel cell system including a unit for electrical isolation of a fuel cell stack from a manifold assembly and method therefor

    Science.gov (United States)

    Kelley; Dana A. , Farooque; Mohammad , Davis; Keith

    2007-10-02

    A fuel cell system with improved electrical isolation having a fuel cell stack with a positive potential end and a negative potential, a manifold for use in coupling gases to and from a face of the fuel cell stack, an electrical isolating assembly for electrically isolating the manifold from the stack, and a unit for adjusting an electrical potential of the manifold such as to impede the flow of electrolyte from the stack across the isolating assembly.

  15. Assembly of laboratory line for nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    The dismantling of a laboratory line for spent fuel reprocessing after the termination of the research programme and the procedures for hot and semi-hot cell decontamination are described. The equipment was mostly disassembled in smaller parts which were then decontaminated by wiping them with cotton wool soaked in detergent and citric acid, varnished with two-component epoxi varnish, wrapped into multiple polyethylene foils, sealed in PVC bags and thus ready for transport. (B.S.)

  16. Electric power generation by a submersible microbial fuel cell equipped with a membrane electrode assembly

    DEFF Research Database (Denmark)

    Min, Booki; Poulsen, Finn Willy; Thygesen, Anders;

    2012-01-01

    Membrane electrode assemblies (MEAs) were incorporated into the cathode chamber of a submersible microbial fuel cell (SMFC). A close contact of the electrodes could produce high power output from SMFC in which anode and cathode electrodes were connected in parallel. In polarization test, the maxi...

  17. Nuclear fuel assembly with a shock absorber system especially for seismic shocks

    International Nuclear Information System (INIS)

    Hydraulic system for absorbing impacts imparted to fuel assemblies. Internal buffers that brace the fuel assemblies so as to restrain their longitudinal displacement, rest against mobile spring buffers. Some of these spring buffers have plungers sliding in hollow tubular guide uprights provided with longitudinal slots. The end of each upright is closed by a plate fitted with an orifice. When an earthquake forces the plunger and spring buffer assemblies to move, the water under pressure in the upright guide tubes stop this movement. This water, which gushes from the holes in the plates, enables the displacement to take place at a controlled rate at which the forces applied are absorbed in complete safety. The plungers gradually close up the slots in the guide uprights, thereby progressively reducing the section through which the water inside the guide upright can flow out. The resistance increases progressively and protects the structure of the reactor core

  18. Spent fuel dry storage technology development: thermal evaluation of isolated drywell containing spent fuel (1.25 kW PWR spent fuel assembly)

    International Nuclear Information System (INIS)

    A spent fuel Isolated Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site utilizing a pressurized water reactor spent fuel assembly having a decay heat level of approximately 1.25 kW. The fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in an instrumented near-surface drywell is located 50 feet from an adjacent drywell. Instrumentation provided to measure canister, liner and soil temperatures consisted of thermocouples which were inserted into tubes on the outside of the canister and drywell liner and thermocouples which were attached to plastic pipe and grouted into holes in the soil. Temperatures from the isolated drywell and the adjacent soil were recorded throughout the nine month Isolated Drywell Test. Canister and drywell liner temperatures reached their peak values (3230F and 2620F, respectively) during October, 1980. Thereafter, all temperatures began to decrease in response to the decay heat and seasonal atmospheric temperature changes. This thermal response is comparable to that of the approximately 1.0 kW spent fuel assemblies previously tested at E-MAD where peak canister and liner temperatures of 2540F and 2030F were recorded. A previously developed computer model was utilized to predict the thermal response of the surrounding soil are presented and are compared with the test data

  19. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    International Nuclear Information System (INIS)

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system

  20. A parametric study of assembly pressure, thermal expansion, and membrane swelling in PEM fuel cells

    Directory of Open Access Journals (Sweden)

    Maher A.R. Sadiq Al-Baghdadi

    2016-01-01

    Full Text Available Proton Exchange membrane (PEM fuel cells are still undergoing intense development, and the combination of new and optimized materials, improved product development, novel architectures, more efficient transport processes, and design optimization and integration are expected to lead to major gains in performance, efficiency, durability, reliability, manufacturability and cost-effectiveness. PEM fuel cell assembly pressure is known to cause large strains in the cell components. All components compression occurs during the assembly process of the cell, but also during fuel cell operation due to membrane swelling when absorbs water and cell materials expansion due to heat generating in catalyst layers. Additionally, the repetitive channel-rib pattern of the bipolar plates results in a highly inhomogeneous compressive load, so that while large strains are produced under the rib, the region under the channels remains approximately at its initial uncompressed state. This leads to significant spatial variations in GDL thickness and porosity distributions, as well as in electrical and thermal bulk conductivities and contact resistances (both at the ribe-GDL and membrane-GDL interfaces. These changes affect the rates of mass, charge, and heat transport through the GDL, thus impacting fuel cell performance and lifetime. In this paper, computational fluid dynamics (CFD model of a PEM fuel cell has been developed to simulate the pressure distribution inside the cell, which are occurring during fuel cell assembly (bolt assembling, and membrane swelling and cell materials expansion during fuel cell running due to the changes of temperature and relative humidity. The PEM fuel cell model simulated includes the following components; two bi-polar plates, two GDLs, and, an MEA (membrane plus two CLs. This model is used to study and analyses the effect of assembling and operating parameters on the mechanical behaviour of PEM. The analysis helped identifying critical

  1. PEM fuel cell cost minimization using ``Design For Manufacture and Assembly`` techniques

    Energy Technology Data Exchange (ETDEWEB)

    Lomax, F.D. Jr.; James, B.D. [Directed Technologies, Inc., Arlington, VA (United States); Mooradian, R.P. [Ford Motor Co., Dearborn, MI (United States)

    1997-12-31

    Polymer Electrolyte Membrane (PEM) fuel cells fueled with direct hydrogen have demonstrated substantial technical potential to replace Internal Combustion Engines (ICE`s) in light duty vehicles. Such a transition to a hydrogen economy offers the potential of substantial benefits from reduced criteria and greenhouse emissions as well as reduced foreign fuel dependence. Research conducted for the Ford Motor Co. under a US Department of Energy contract suggests that hydrogen fuel, when used in a fuel cell vehicle (FCV), can achieve a cost per vehicle mile less than or equal to the gasoline cost per mile when used in an ICE vehicle. However, fuel cost parity is not sufficient to ensure overall economic success: the PEM fuel cell power system itself must be of comparable cost to the ICE. To ascertain if low cost production of PEM fuel cells is feasible, a powerful set of mechanical engineering tools collectively referred to as Design for Manufacture and Assembly (DFMA) has been applied to several representative PEM fuel cell designs. The preliminary results of this work are encouraging, as presented.

  2. The development of a package for the transport of new mixed oxide fuel assemblies within Europe

    International Nuclear Information System (INIS)

    The use of mixed oxide (MOX) fuels in commercial reactors has increased significantly over the past 10 years as an effective way of using stocks of plutonium produced from reprocessing uranium fuels. Now, with advances in fuel design MOX can give performance approaching that of enriched uranium fuel. To meet demand from European and Japanese utilities, British Nuclear Fuels are currently building a large scale plant at Sellafield to assemble MOX fuels. This required a new transport package to be developed capable of carrying high specification fuels to customers in Europe whilst complying with the latest 1996 IAEA ST-1 Transport Regulations. This package is known as Euromox and currently under development to enter service in 2003. Relatively few packages exist for the transport of MOX fuels and Euromox is the first designed by BNFL for shipments to Europe. Euromox has provided several technical challenges in its development arguably exceeding those typically encountered during the development of new packages for irradiated fuel transports. (author)

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

    International Nuclear Information System (INIS)

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

  4. Verification of plutonium content in spent fuel assemblies using neutron self-interrogation

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, Howard O [Los Alamos National Laboratory; Menlove, Apencer H [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

    2009-01-01

    The large amounts of plutonium in reactor spent fuel assemblies has led to increased research directed toward the measurement of the plutonium for safeguards verification. The high levels of fission product gamma-ray activity and curium neutron backgrounds have made the plutonium measurement difficult. We have developed a new technique that can directly measure both the {sup 235}U concentration and the plutonium fissile concentration using the intrinsic neutron emission fronl the curium in the fuel assembly. The passive neutron albedo reactivity (PNAR) method has been described previously where the curium neutrons are moderated in the surrounding water and reflect back into the fuel assembly to induce fissions in the fissile material in the assembly. The cadmium (Cd) ratio is used to separate the spontaneous fission source neutrons from the reflected thermal neutron fission reactions. This method can measure the sum of the {sup 235}U and the plutonium fissile mass, but not the separate components. Our new differential die-away self-interrogation method (DDSI) can be used to separate the {sup 235}U from the {sup 239}Pu. The method has been applied to both fuel rods and full assemblies. For fuel rods the epi-thermal neutron reflection method filters the reflected neutrons through thin Cd filters so that the reflected neutrons are from the epi-cadmium energy region. The neutron fission energy response in the epi-cadmium region is distinctly different for {sup 235}U and {sup 239}Pu. We are able to measure the difference between {sup 235}U and {sup 239}Pu by sampling the neutron induced fission rate as a function of time and multiplicity after the initial fission neutron is detected. We measure the neutron fission rate using list-mode data collection that stores the time correlations between all of the counts. The computer software can select from the data base the time correlations that include singles, doubles, and triples. The die-away time for the doubles

  5. Thermal simulation experiments with a UO2 fuel rod assembly

    International Nuclear Information System (INIS)

    Evidence is presented which shows that columnar grains can be induced to grow in high-density sintered uranium dioxide specimens by applying a steep temperature gradient at temperatures above 1700oC but below the melting point of 2800oC. Columnar growth apparently is a result of the migration of large transverse voids, whose individual widths define the cross sections of the grains, up a temperature gradient by a sublimation process. The grains grown by this process have a (iii) preferred orientation along their columnar axis. A consequence of such void migration in operating fuel elements containing solid UO2 pellets is the formation of a central void bounded by a region of oxide exhibiting columnar growth. (author)

  6. Water confinement effects in response of fuel assembly to faulted condition loads

    Energy Technology Data Exchange (ETDEWEB)

    Shah, S.J.; Brenneman, B.; Williams, G.T.; Strumpel, J.H. [Framatome-ANP, Lynchburg (United States)

    2004-07-01

    It has been established by other authors that the accelerations of the water confined by the reactor core baffle plates has a significant effect on the responses of all the fuel assemblies during LOCA (loss of coolant accident) or seismic transients. This particular effect is a consequence of the water being essentially incompressible, and thus experiencing the same horizontal accelerations as the imposed baffle plate motions. These horizontal accelerations of the fluid induce lateral pressure gradients that cause horizontal buoyancy forces on any submerged structures. These forces are in the same direction as the baffle accelerations and, for certain frequencies at least, tend to reduce the relative displacements between the fuel and baffle plates. But there is another confinement effect: the imposed baffle plate velocities must also be transmitted to the water. If the fuel assembly grid strips are treated as simple hydro-foils, these horizontal velocity components change the fluid angle of attack on each strip, and thus may induce large horizontal lift forces on each grid in the same direction as the baffle plate velocity. There is a similar horizontal lift due to inclined flow over the rods when axial flow is present. These combined forces appear to reduce the relative displacements between the fuel and baffle plates for any significant axial flow velocity. Modeling this effect is very simple. It was shown in previous papers that the mechanism for the large fuel assembly damping due to axial flow may be the hydrodynamic forces on the grid strips, and that this is very well represented by discrete viscous dampers at each grid elevation. To include the imposed horizontal water velocity effects, on both the grids and rods, these dampers are simply attached to the baffle plate rather than 'ground'. The large flow-induced damping really acts in a relative reference frame rather than an inertial reference frame, and thus it becomes a flow-induced coupling

  7. Acceptance of non-fuel assembly hardware by the Federal Waste Management System

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high-level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. 14 refs., 12 figs., 43 tabs

  8. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    International Nuclear Information System (INIS)

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e’n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program

  9. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio [Universidad Simón Bolívar, Nuclear Physics Laboratory, Apdo 89000, Caracas 1080A (Venezuela, Bolivarian Republic of); Davila, Jesus [Física Médica C. A. and Universidad Central de Venezuela, Caracas (Venezuela, Bolivarian Republic of)

    2015-07-23

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e’n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides {sup 232}Th, {sup 238}U and {sup 237}Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  10. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    Science.gov (United States)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

    2015-07-01

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  11. Predicting fissile content of spent nuclear fuel assemblies with the Passive Neutron Albedo Reactivity technique and Monte Carlo code emulation

    International Nuclear Information System (INIS)

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed. (author)

  12. Predicting fissile content of spent nuclear fuel assemblies with the passive neutron Albedo reactivity technique and Monte Carlo code emulation

    International Nuclear Information System (INIS)

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed.

  13. Influence of Metal Shells around Fuel Assemblies on Criticality Calculations for a Fuel Storage Pool

    OpenAIRE

    Babichev, L.; Khmialeuski, A.

    2012-01-01

    Influence of metal shells and size of cells in a fuel storage pool on the value of the effective neutron multiplication factor was studied. Monte Carlo code MCU-FREE was used in the criticality calculations. A criticality analysis of spent fuel storage pools for different degrees of packing of cells in the rack was performed.

  14. A coating to protect spent aluminium-clad research reactor fuel assemblies during extended wet storage

    International Nuclear Information System (INIS)

    Pitting corrosion of aluminium (Al) alloy clad research reactor (RR) fuel in wet storage facilities can be reduced to a large extent by maintaining water parameters within specified limits. However, factors like bimetallic contact, settled solids and synergistic effects of many storage basin water parameters provoke cladding corrosion. Increase in corrosion resistance of spent Al-clad RR fuels can be achieved through the use of conversion coatings. This paper presents: (a) details about the formation of cerium dioxide as a conversion coating on Al alloys used as RR fuel cladding; (b) the corrosion resistance of cerium dioxide coated Al alloy specimens exposed to NaCl solutions. Marked improvements in corrosion resistance of cerium dioxide coated Al specimens were observed. This paper also presents details of a Latin American Project to develop conversion coatings for long term safe wet storage of spent Al-clad RR spent fuel assemblies. (author)

  15. CFD prediction of flow and phase distribution in fuel assemblies with spacers

    Energy Technology Data Exchange (ETDEWEB)

    Anglart, H.; Nylund, O. [ABB Atom AB, Vasteras (Switzerland); Kurul, N. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling and computation of multi-dimensional two-phase flows in BWR fuel assemblies. The modeling principles are presented based on using a two-fluid model in which lateral interfacial effects are accounted for. This model has been used to evaluate the velocity fields of both vapor and liquid phases, as well as phase distribution, between fuel elements in geometries similar to BWR fuel bundles. Furthermore, this model has been used to predict, in a detailed mechanistic manner, the effects of spacers on flow and phase distribution between, and pressure drop along, fuel elements. The related numerical simulations have been performed using a CFD computer code, CFDS-FLOW3D.

  16. Comparison of hydrogen generation for TVSM and TVSA fuel assemblies for water water energy reactor (VVER)-1000

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, A.E. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)], E-mail: antoanet@inrne.bas.bg; Groudev, P.P. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)], E-mail: pavlinpg@inrne.bas.bg; Atanasova, B.P. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)], E-mail: b_atanasova@inrne.bas.bg

    2009-01-15

    This paper presents the results received during investigation of hydrogen generation for both types fuel assemblies-the old modernistic type of fuel assemblies (TVSM) and recently installed new one alternative type of fuel assemblies (TVSA) in case of severe accident. There are some differences between both types FAs. They have different geometry as well as different burnable poisons. To investigate behavior of new fuel assemblies during the severe conditions it have been performed comparison of fuel behavior of old type TVSM fuel assembly to new one TVSA. To perform this investigation it has been used MELCOR 'input model' for Kozloduy Nuclear Power Plant (KNPP) VVER 1000. The model was developed by Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) for investigation of severe accident scenarios and Probabilistic Safety Analyses (PSA) level 2. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides release outside fuel cladding. It was accepted criteria for vessel integrity about hydrogen concentration to be 8%. This criterion was based on the decision of RSK (Germany commission for reactor safety). Generally based on the received results it was made conclusion that using both types of fuel assemblies it was not disturbance safety conditions of NPP.

  17. Computed isotopic inventory and dose assessment for SRS fuel and target assemblies

    International Nuclear Information System (INIS)

    Past studies have identified and evaluated important radionuclide contributors to dose from reprocessed spent fuel sent to waste for Mark 16B and 22 fuel assemblies and for Mark 31 A and 31B target assemblies. Fission-product distributions after a 5- and 15-year decay time were calculated for a ''representative'' set of irradiation conditions (i.e., reactor power, irradiation time, and exposure) for each type of assembly. The numerical calculations were performed using the SHIELD/GLASS system of codes. The sludge and supernate source terms for dose were studied separately with the significant radionuclide contributors for each identified and evaluated. Dose analysis considered both inhalation and ingestion pathways: The inhalation pathway was analyzed for both evaporative and volatile releases. Analysis of evaporative releases utilized release fractions for the individual radionuclides as defined in the ICRP-30 by DOE guidance. A release fraction of unity was assumed for each radionuclide under volatile-type releases, which would encompass internally initiated events (e.g., fires, explosions), process-initiated events, and externally initiated events. Radionuclides which contributed at least 1% to the overall dose were designated as significant contributors. The present analysis extends and complements the past analyses through considering a broader spectrum of fuel types and a wider range of irradiation conditions. The results provide for a more thorough understanding of the influences of fuel composition and irradiation parameters on fission product distributions (at 2 years or more). Additionally, the present work allows for a more comprehensive evaluation of radionuclide contributions to dose and an estimation of the variability in the radionuclide composition of the dose source term that results from the spent fuel sent to waste encompassing a broad spectrum of fuel compositions and irradiation conditions

  18. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G. [and others

    1997-10-01

    The Solution High-Energy Burst Assembly (SHEBA) was originally constructed during 1980 and was designed to be a clean free-field geometry, right-circular, cylindrically symmetric critical assembly employing U(5%)O{sub 2}F{sub 2} solution as fuel. A second version of SHEBA, employing the same fuel but equipped with a fuel pump and shielding pit, was commissioned in 1993. This report includes data and operating experience for the 1993 SHEBA only. Solution-fueled benchmark work focused on the development of experimental measurements of the characterization of SHEBA; a summary of the results are given. A description of the system and the experimental results are given in some detail in the report. Experiments were designed to: (1) study the behavior of nuclear excursions in a low-enrichment solution, (2) evaluate accidental criticality alarm detectors for fuel-processing facilities, (3) provide radiation spectra and dose measurements to benchmark radiation transport calculations on a low-enrichment solution system similar to centrifuge enrichment plants, and (4) provide radiation fields to calibrate personnel dosimetry. 15 refs., 37 figs., 10 tabs.

  19. The influence of ribbon holes on fuel assembly thermal hydraulic performance

    International Nuclear Information System (INIS)

    Two kinds of PWR space grids, structure mixing grid (MG) and mid span mixing grid (MSMG) exist in fuel assembly. There are spring, dimple, ribbon and mixing vanes in Structure mixing grid. The ribbon with hole or without hole will influence the fuel assembly thermal-hydraulic performance. Two 5×5 rod bundle with two space girds, one is structure mixing grid, another is mid span mixing grid are analyzed using CFD method. The two calculations are different in structure mixing grid, one with holes on ribbon, the other one without holes on ribbon. The fuel rods are heated. The whole meshes are about 20 million. The result shows that the ribbon with holes increases pressure loss in space grid area. The holes’ influence on temperature field and flow field are mainly observed at structure space grid downstream, near the mid span mixing grid. Though the ribbon with holes shows limit influence on temperature field, it makes the low pressure area appears around peripheral fuel rods, and the swirl near the ribbon somewhere disappear, thus bubbles are more possible to cluster near the peripheral fuel rods. For further research, two phase CFD could be used to analyze the bubble behaviors. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-01

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

  1. On the evaluation of a fuel assembly design by means of uncertainty and sensitivity measures

    International Nuclear Information System (INIS)

    This paper will provide results of an uncertainty and sensitivity study in order to calculate parameters of safety related importance like the fuel centerline temperature, the cladding temperature and the fuel assembly pressure drop of a lead-alloy cooled fast system. Applying best practice guidelines, a list of uncertain parameters has been identified. The considered parameter variations are based on the experience gained during fabrication and operation of former and existing liquid metal cooled fast systems as well as on experimental results and on engineering judgment. (orig.)

  2. COBRA-SFS predictions of single assembly spent fuel heat transfer data

    International Nuclear Information System (INIS)

    The study reported here is one of several efforts to evaluate and qualify the COBRA-SFS computer code for use in spent fuel storage system thermal analysis. The ability of COBRA-SFS to predict the thermal response of two single assembly spent fuel heat transfer tests was investigated through comparisons of predictions with experimental test data. From these comparisons, conclusions regarding the computational treatment of the physical phenomena occurring within a storage system can be made. This objective was successfully accomplished as reasonable agreement between predictions and data were obtained for the 21 individual test cases of the two experiments

  3. Evaluation of CASMO-3 and HELIOS for Fuel Assembly Analysis from Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hyung Jin; Song, Jae Seung; Lee, Chung Chan

    2007-05-15

    This report presents a study comparing deterministic lattice physics calculations with Monte Carlo calculations for LWR fuel pin and assembly problems. The study has focused on comparing results from the lattice physics code CASMO-3 and HELIOS against those from the continuous-energy Monte Carlo code McCARD. The comparisons include k{sub inf}, isotopic number densities, and pin power distributions. The CASMO-3 and HELIOS calculations for the k{sub inf}'s of the LWR fuel pin problems show good agreement with McCARD within 956pcm and 658pcm, respectively. For the assembly problems with Gadolinia burnable poison rods, the largest difference between the k{sub inf}'s is 1463pcm with CASMO-3 and 1141pcm with HELIOS. RMS errors for the pin power distributions of CASMO-3 and HELIOS are within 1.3% and 1.5%, respectively.

  4. Parametric study on parallel flow induced damping of PWR fuel assembly

    International Nuclear Information System (INIS)

    This paper reports on a mechanism of parallel flow-induced changes in vibrational characteristics of PWR fuel assemblies that has been studied through a series of hydraulic tests using reduced-and full-scale prototype mockups. Measured data and analytical evaluations showed the phenomenon stands on essentially the same basis as the dynamics and stability of flexible cylinders subjected to a parallel flow. In the mathematical model, the effects of rod bundle geometries and boundaries formed by walls or adjacent bundles can be exactly incorporated in the form of added mass coefficients, velocity coupling coefficients and other fluid forces. From a full scale test, it has been shown that coolant temperature has little effect up to reactor operating conditions. The updated FEM model has been verified to be applicable in describing the vibrational characteristics of from an isolated cylinder to a full scale fuel assembly in terms of the consistent properties

  5. Droplet entrainment and deposition rate models for determination of boiling transition in BWR fuel assembly

    International Nuclear Information System (INIS)

    Droplet entrainment and deposition rates are of vital importance for mechanistic determination of critical power and location of boiling transition in a BWR fuel assembly. Data from high-pressure, high-temperature steam-water adiabatic experiments conducted in very tall test sections are used to develop a combination of equilibrium entrainment-deposition rate. Application of this combination to the heated tests conducted in a shorter test section of typical height of a BWR fuel assembly shows that correct split of total liquid in form of the film and droplets at the onset of annular-mist flow regime is also important to obtain good prediction of film flow rates/entrainment fraction. The improved model is then applied to simulate critical power tests in annulus and rod bundles. (author)

  6. Two dimensional, two fluid model for sodium boiling in LMFBR fuel assemblies

    International Nuclear Information System (INIS)

    A two dimensional numerical model for the simulation of sodium boiling transient was developed using the two fluid set of conservation equations. A semiimplicit numerical differencing scheme capable of handling the problems associated with the ill-posedness implied by the complex characteristic roots of the two fluid problems was used, which took advantage of the dumping effect of the exchange terms. Of particular interest in the development of the model was the identification of the numerical problems caused by the strong disparity between the axial and radial dimensions of fuel assemblies. A solution to this problem was found which uses the particular geometry of fuel assemblies to accelerate the convergence of the iterative technique used in the model. Three sodium boiling experiments were simulated with the model, with good agreement between the experimental results and the model predictions

  7. Blockages in LMFBR fuel assemblies: a review of experimental and theoretical studies

    Energy Technology Data Exchange (ETDEWEB)

    Han, J. T.

    1977-08-08

    This is a state-of-the-art report on the thermal-hydraulic effects of flow-channel blockages in liquid-metal fast breeder reactor (LMFBR) pin bundles. Most of the experimental and theoretical studies for simulating blockages in various prototype LMFBR fuel assemblies done in the United States and abroad through 1976 are presented and summarized. A brief summary on blockage detection is included.

  8. 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)

  9. Quality assurance in the procurement, design and manufacture of nuclear fuel assemblies

    International Nuclear Information System (INIS)

    This Safety Guide provides requirements and recommendations for quality assurance programmes that are relevant for the unique features of the procurement, design, manufacture, inspection, testing, packaging, shipping, storage, and receiving inspection of fuel assemblies for nuclear power plants. The generic quality assurance requirements of the Code and related Safety Guides are referred to where applicable, and are duplicated in this document where increased emphasis is desirable

  10. Castor transport and storage casks for VVER and RBMK fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Gartz, R.; Gobler, A.; John, R.; Diersch, R. [GNB Gesellschaft fur Nuklear-Behalter mbH, Essen (Germany); Nemec, P. [Skoda Nuclear Machinery Plzen (Czech Republic)

    1998-12-31

    CASTOR casks have been successfully developed, manufactured and delivered for Russian type reactor fuel assemblies. These casks fulfill both the requirements for type B packages according to IAEA regulations and the requirements covering different accident situations to be assumed at the storage site. In the following, the CASTOR casks CASTOR 440/84, CASTOR RBMK and CASTOR VVER 1000 are described, the nuclear content is characterized and an overview about the status of licensing, manufacturing and delivery is given. (authors) 3 refs.

  11. Space Shuttle booster recovery planning.

    Science.gov (United States)

    Godfrey, R. E.

    1973-01-01

    At the initiation of the Space Shuttle Program, recoverable solid rocket boosters were base-lined, with an estimated savings of 30 per cent over expendable solid rockets. Present studies indicate that the solid rocket boosters in the 142-inch diameter range can be recovered using state-of-the-art recovery systems. Marshall Space Flight Center is conducting extensive studies to establish the most cost effective recovery system for the present Shuttle boosters. Model drop testing, in various facilities, and structural load testing are being conducted with model sizes ranging from 6 inches to 120 inches in diameter.

  12. Development of image processing system for innovative laser welding of nuclear fuel assembly

    International Nuclear Information System (INIS)

    Laser welding technique has been rapidly developed with respect to high power lasers such as CO2 laser, Nd:YAG laser, and etc. Laser welding has advantages to weld materials more precisely due to precise focusing of the laser beam and a little heat affected zone(HAZ). Nowadays, Nd:YAG lasers are used for manufacturing spacer grid assembly of nuclear fuel. The correct positioning is very important factor in laser welding of space grid. We have developed an image processing system and optical assembly for correcting weld positions of space grid assembly in laser welding process. In this paper, the development of optical design and image processing system for a new innovative laser welding system is described

  13. Total evaluation of in bundle void fraction measurement test of PWR fuel assembly

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation is performing the various proof or verification tests on safety and reliability of nuclear power plants under the sponsorship of the Ministry of International Trade and Industry. As one program of these Japanese national projects, an in bundle void fraction measurement test of a pressurized water reactor (PWR) fuel assembly was started in 1987 and finished at the end of 1994. The experiments were performed using the 5 x 5 square array rod bundle test sections. The rod bundle test section simulates the partial section and full length of a 17 x 17 type Japanese PWR fuel assembly. A distribution of subchannel averaged void fraction in a rod bundle test section was measured by the gamma-ray attenuation method using the stationary multi beam systems. The additional single channel test was performed to obtain the required information for the calibration of the rod bundle test data and the assessment of the void prediction method. Three test rod bundles were prepared to analyze an axial power distribution effect, an unheated rod effect, and a grid spacer effect. And, the obtained data were used for the assessment of the void prediction method relevant to the subchannel averaged void fraction of PWR fuel assemblies. This paper describes the outline of the experiments, the evaluation of the experimental data and the assessment of void prediction method

  14. Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

    Energy Technology Data Exchange (ETDEWEB)

    Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

    2007-03-15

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.

  15. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    International Nuclear Information System (INIS)

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution

  16. Iterative ct reconstruction from few projections for the nondestructive post irradiation examination of nuclear fuel assemblies

    Science.gov (United States)

    Abir, Muhammad Imran Khan

    The core components (e.g. fuel assemblies, spacer grids, control rods) of the nuclear reactors encounter harsh environment due to high temperature, physical stress, and a tremendous level of radiation. The integrity of these elements is crucial for safe operation of the nuclear power plants. The Post Irradiation Examination (PIE) can reveal information about the integrity of the elements during normal operations and off?normal events. Computed tomography (CT) is a tool for evaluating the structural integrity of elements non-destructively. CT requires many projections to be acquired from different view angles after which a mathematical algorithm is adopted for reconstruction. Obtaining many projections is laborious and expensive in nuclear industries. Reconstructions from a small number of projections are explored to achieve faster and cost-efficient PIE. Classical reconstruction algorithms (e.g. filtered back projection) cannot offer stable reconstructions from few projections and create severe streaking artifacts. In this thesis, conventional algorithms are reviewed, and new algorithms are developed for reconstructions of the nuclear fuel assemblies using few projections. CT reconstruction from few projections falls into two categories: the sparse-view CT and the limited-angle CT or tomosynthesis. Iterative reconstruction algorithms are developed for both cases in the field of compressed sensing (CS). The performance of the algorithms is assessed using simulated projections and validated through real projections. The thesis also describes the systematic strategy towards establishing the conditions of reconstructions and finds the optimal imaging parameters for reconstructions of the fuel assemblies from few projections.

  17. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  18. Benchmark physics experiment of metallic-fueled LMFBR at FCA. 2; Experiments of FCA assembly XVI-1 and their analyses

    Energy Technology Data Exchange (ETDEWEB)

    Iijima, Susumu; Oigawa, Hiroyuki; Ohno, Akio; Sakurai, Takeshi; Nemoto, Tatsuo; Osugi, Toshitaka; Satoh, Kunio; Hayasaka, Katsuhisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Bando, Masaru

    1993-10-01

    An availability of data and method for a design of metallic-fueled LMFBR is examined by using the experiment results of FCA assembly XVI-1. Experiment included criticality and reactivity coefficients such as Doppler, sodium void, fuel shifting and fuel expansion. Reaction rate ratios, sample worth and control rod worth were also measured. Analysis was made by using three-dimensional diffusion calculations and JENDL-2 cross sections. Predictions of assembly XVI-1 reactor physics parameters agree reasonably well with the measured values, but for some reactivity coefficients such as Doppler, large zone sodium void and fuel shifting further improvement of calculation method was need. (author).

  19. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Waseem,, E-mail: wazim_me@hotmail.com; Murtaza, Ghulam; Elahi, Nadeem

    2014-12-15

    Highlights: • Finite element model of CHASNUPP-1 fuel assembly produced, using Shell181 elements. • Non-linear contact and buckling analysis have been performed. • Structural integrity and stress measurement of fuel assembly is calculated. • Calculated stresses and deformations, are compared with test results. • Results of both studies are comparable, which validate finite element methodology. - Abstract: Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA) of Chashma Nuclear Power Plant-1 (CHASNUPP-1) at room temperature in air. Non-linear contact and buckling analyses have been performed using ANSYS 13.0, in-order to determine the FA's deformation behaviour as well as the location/values of the maximum stress intensity and stresses developed in axial direction under applied compression load of 7350 N or 1.6 g being the FA's handling load (Zhang et al., 1994). The finite element (FE) model comprises spacer grids, fuel rods, flexible contact between the fuel rods and grids’ supports system (springs and dimples) and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behaviour of the geometry is non-linear. The value of the perturbation force is related to the geometry of the model and/or the tolerance defined for the geometry. Therefore, a sensitivity study has been made to determine the appropriate value of an arbitrary perturbation load. It has been observed that FA deformation values obtained through FE analysis and experiment (SNERDI Tech Doc, 1994) under applied compression load are comparable and show linear behaviours. Therefore, it is confirmed that buckling of FA will not occur at the specified load. Moreover, the values of stresses obtained

  20. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    International Nuclear Information System (INIS)

    Highlights: • Finite element model of CHASNUPP-1 fuel assembly produced, using Shell181 elements. • Non-linear contact and buckling analysis have been performed. • Structural integrity and stress measurement of fuel assembly is calculated. • Calculated stresses and deformations, are compared with test results. • Results of both studies are comparable, which validate finite element methodology. - Abstract: Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA) of Chashma Nuclear Power Plant-1 (CHASNUPP-1) at room temperature in air. Non-linear contact and buckling analyses have been performed using ANSYS 13.0, in-order to determine the FA's deformation behaviour as well as the location/values of the maximum stress intensity and stresses developed in axial direction under applied compression load of 7350 N or 1.6 g being the FA's handling load (Zhang et al., 1994). The finite element (FE) model comprises spacer grids, fuel rods, flexible contact between the fuel rods and grids’ supports system (springs and dimples) and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behaviour of the geometry is non-linear. The value of the perturbation force is related to the geometry of the model and/or the tolerance defined for the geometry. Therefore, a sensitivity study has been made to determine the appropriate value of an arbitrary perturbation load. It has been observed that FA deformation values obtained through FE analysis and experiment (SNERDI Tech Doc, 1994) under applied compression load are comparable and show linear behaviours. Therefore, it is confirmed that buckling of FA will not occur at the specified load. Moreover, the values of stresses obtained

  1. Coupled computational fluid dynamics and MOC neutronic simulations of Westinghouse PWR fuel assemblies with grid spacers

    International Nuclear Information System (INIS)

    Neutronic coupling with Computational Fluid Dynamics (CFD) has been under development within the US DOE sponsored “Nuclear Simulation Hub”. The method of characteristics (MOC) neutronics code DeCART ([Joo, 2004], [Kochunas, 2009]) under development at the University of Michigan was coupled with the CFD code STAR-CCM+ to achieve more accurate predictions of fuel assembly performance. At Westinghouse, lower order, neutronic codes such as the nodal code ANC have been coupled to thermal-hydraulics codes such the subchannel code VIPRE to predict the heat flux and fuel nuclear behavior. However, a more detailed neutronics and temperature / fluid field simulation of fuel assembly models which includes explicit representation of spacer grids would considerably improve the design and assessment of new fuel assembly designs. Coupled STAR-CCM+ / DeCART calculations have been performed for various representative three-dimensional models with explicit representation of spacer grids with mixing vanes. The high fidelity results have been compared to lower order simulations. The coupled CFD/MOC solution has provided a more truthful model which includes a more accurate representation of all the important physics such as fission energy, heat convection, heat conduction, and turbulence. Of particular significance is the ability to assess the effects of the mixing grid on the coolant temperature and density distribution using coupled thermal/fluids and neutronic solutions. A more precise cladding temperature can be derived by this approach which will also enable more accurate prediction of departure from nucleate boiling (DNB), as well as a better understanding of DNB margin and crud build up on the fuel rod. (author)

  2. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    International Nuclear Information System (INIS)

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs

  3. Advanced manufacturing of intermediate temperature, direct methane oxidation membrane electrode assemblies for durable solid oxide fuel cell Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ITN proposes to create an innovative anode supported membrane electrode assembly (MEA) for solid oxide fuel cells (SOFCs) that is capable of long-term operation at...

  4. Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

    International Nuclear Information System (INIS)

    The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling

  5. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    International Nuclear Information System (INIS)

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO2 fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  6. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  7. FEM analysis of the top nozzle with welding parts of 16NGF fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Schettino, Carlos Frederico Mattos; Brittes, Luiz Henrique Alves; Silva, Marcio Adriano Coelho da [Industrias Nucleares do Brasil (INB), Resende, RJ (Brazil)]. E-mails: carlosschettino@inb.gov.br; brittes@inb.gov.br; marcio.adriano@inb.gov.br

    2007-07-01

    The present work aims to evaluate structurally the welded Top Nozzle (previously casting) used in the fuel assembly type 16 x 16 used in the Nuclear Power Plant Angra I. The full solid model of this component was generated with SOLIDWORKS program and later imported to ANSYS 10.0 program. For the Finite Element model were used elements SOLID-92 and BEAM3. The validation of the analysis was lead comparing the results got from simulation using ANSYS to physical tests already performed previously in the Westinghouse Electric Company (U.S.A.). The analysis covered specific loads simulating the conditions found during the shipping and handling of the Fuel Element (static loads corresponding to 4g - four times the Fuel Element weight) as well as simulating the conditions found during the operation of the nuclear power plant (Conditions I, II, III and IV). The structural integrity of the Top Nozzle is assured when the design criteria, defined in the ASME Code Section III (Boilers and Pressures Vessels), are satisfied. The results of these analysis were used to prove that the welded Top Nozzle is capable to keep the dimensional stability for which it was designed when submitted to loads that correspond to required stresses for its use in Nuclear Fuel Assemblies. The performed analysis provided INB to get more information of extreme importance for the continuity of the development of the welded Top Nozzle and its later production. (author)

  8. Performance evaluation of booster materials in the plastic bonded explosive PBX 9502 in a hemispherical wave breakout test

    Energy Technology Data Exchange (ETDEWEB)

    Hooks, Daniel E [Los Alamos National Laboratory; Morris, John S [Los Alamos National Laboratory; Hill, Larry G [Los Alamos National Laboratory; Francois, Elizabeth [Los Alamos National Laboratory

    2008-01-01

    An explosive booster is normally required to initiate detonation in an insensitive high explosive (lHE). Booster materials must be ignitable by a conventional detonator and deliver sufficient energy and favorable pulse shape to initiate the IHE charge. The explosive booster should be as insensitive as reasonably possible to maintain the overall safety margin of the explosive assembly. A hemispherical wave breakout test termed the on ionskin test is one of the methods of testing the performance of booster materials in an initiation train assembly. There are several variations of this basic test which are known by other names. In this test, the wave breakout time-position history at the surface of a hemispherical IHE acceptor charge is recorded, and the relative uniformity of breakout allows qualitative comparison between booster candidates and quantitative comparison of several metrics. The results of a series of onionskin experiments evaluating the performance of some new booster formulations in the triaminotrinitrobenzene (TA TB) -based plastic bonded explosive PBX 9502 will be presented. The boosters were tested in an onionskin arrangement in which the booster pellet was cylindrical, and the tests were performed at a temperature of-55{sup o}C to emphasize variations in spreading performance. The modification from the traditional hemispherical geometry facilitated efficient explosive fabrication and charge assembly, but the results indicate that this geometry was not ideal for several reasons. Despite the complications arising from geometry, promising performance was observed from booster formulations including 3,3' -diamino-4,4'azoxyfurazan.

  9. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  10. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Science.gov (United States)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  11. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study

  12. Conceptual design report: superconducting booster

    International Nuclear Information System (INIS)

    The Superconducting Booster project includes the construction of a new high-voltage injector and buncher for the existing tandem, a magnetic transport system, an rf linac with superconducting resonators, and a rebuncher-debuncher. The booster will fit in existing space so that a new building is not required. The layout of the accelerator is given in Fig. I-1. The University of Washington is contributing approximately $1 M to this project

  13. Effect of moderator density distribution of annular flow on fuel assembly neutronic characteristics in boiling water reactor cores

    International Nuclear Information System (INIS)

    The effect of the moderator density distribution of annular flow on the fuel assembly neutronic characteristics in a boiling water nuclear reactor was investigated using the SRAC95 code system. For the investigation, a model of annular flow for fuel assembly calculation was utilized. The results of the assembly calculation with the model (Method 1) and those of the fuel assembly calculation with the uniform void fraction distribution (Method 2) were compared. It was found that Method 2 underestimates the infinite multiplication factor in the fuel assembly including the gadolinia rod (type 1 assembly). This phenomenon is explained by the fact that the capture rate in the thermal energy region in gadolinia fuel is estimated to be smaller when the liquid film of annular flow at the fuel rod surface is considered. A burnup calculation was performed under the condition of a void fraction of 65% and a volumetric fraction of the liquid film in liquid phase of 1. It is found that Method 2 underestimates the infinite multiplication factor in comparison to Method 1 in the early stage of burnup, and that Method 2 becomes to overestimate the factor after a certain degree of burnup. This is because Method 2 overestimates the depletion rate of the gadolinia. (author)

  14. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    Science.gov (United States)

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

  15. Impedance Analysis of the Conditioning of PBI–Based Electrode Membrane Assemblies for High Temperature PEM Fuel Cells

    DEFF Research Database (Denmark)

    Araya, Samuel Simon; Vang, Jakob Rabjerg; Andreasen, Søren Juhl;

    2013-01-01

    This work analyses the conditioning of single fuel cell assemblies based on different membrane electrode assembly (MEA) types, produced by different methods. The analysis was done by means of electrochemical impedance spectroscopy, and the changes in the fitted resistances of the all the tested...

  16. Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

    Energy Technology Data Exchange (ETDEWEB)

    Ammirabile, Luca, E-mail: luca.ammirabile@ec.europa.e [European Commission, JRC, Institute for Energy, Westerduinweg 3, 1755 LE Petten (Netherlands)

    2010-10-15

    In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).

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

    International Nuclear Information System (INIS)

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

  18. Depleting a CANDU-6 fuel assembly using detailed burnup data and reactionwise energy release

    International Nuclear Information System (INIS)

    Temporal behavior of reactor fuel assembly due to neutron exposure is an integral part of lattice analysis. It is important to estimate the production of actinides and fission products as a function of burnup so as to decide the quality of fuel for further energy production. It is also important from the point of view of post irradiation behavior of fuel. The information on heat production during and after irradiation helps in determining the amount of time a fuel assembly needs to be cooled before taking it up for storage or reprocessing. In the present study we have considered the CANDU-6 fuel assembly as reference. Lattice analysis has been performed using development version of code DRAGON. A total of 192 nuclides have been selected as part of the analysis, of which 19 are actinides, 151 are fission products and the rest are structural elements. The fission products have been treated explicitly. There is no pseudo fission product. Using DRAGR module, a multigroup microscopic cross section library in DRAGLIB format has been generated. An important aspect of this library is the explicit treatment of most neutron induced reactions. We have for the first time attempted to perform power normalization due to energy from various neutron induced reactions including (n, γ), (n, f), (n, 2n), (n, 3n), (n, 4n), (n, α), (n, p), (n, 2α), (n, np), (n, d), (n, t). Energy due to decay has also been considered explicitly. Even though the decay energy contributes very little relative to the neutron induced reactions, the information will be very useful for post irradiation behavior of fuel. It was observed that the maximum contributing reactions for the power normalization are (n, f), (n, γ) and (n, 2n). We have assessed the contribution of fission products and actinides towards power normalization as a function of burnup. We have also studied the pinwise contribution towards power normalization in each ring of CANDU-6 fuel. We have attempted to compare the effect of

  19. Mechanical behaviour of membrane electrode assembly (MEA during cold start of PEM fuel cell from subzero environment temperature

    Directory of Open Access Journals (Sweden)

    Maher A.R. Sadiq Al-Baghdadi

    2015-01-01

    Full Text Available Durability is one of the most critical remaining issues impeding successful commercialization of broad PEM fuel cell transportation energy applications. Automotive fuel cells are likely to operate with neat hydrogen under load-following or load-levelled modes and be expected to withstand variations in environmental conditions, particularly in the context of temperature and atmospheric composition. In addition, they are also required to survive over the course of their expected operational lifetimes i.e., around 5,500 hrs, while undergoing as many as 30,000 startup/shutdown cycles. Cold start capability and survivability of proton exchange membrane fuel cells (PEM in a subzero environment temperature remain a challenge for automotive applications. A key component of increasing the durability of PEM fuel cells is studying the behaviour of the membrane electrode assembly (MEA at the heart of the fuel cell. The present work investigates how the mechanical behaviour of MEA are influenced during cold start of the PEM fuel cell from subzero environment temperatures. Full three-dimensional, non-isothermal computational fluid dynamics model of a PEM fuel cell has been developed to simulate the stresses inside the PEM fuel cell, which are occurring during fuel cell assembly (bolt assembling, and the stresses arise during fuel cell running due to the changes of temperature and relative humidity. The model is shown to be able to understand the many interacting, complex electrochemical, transport phenomena, and stresses distribution that have limited experimental data.

  20. Distribution of fission products in graphite sleeves and blocks of the eleventh and twelfth OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    The 11th and 12th fuel assemblies were irradiated in an in-pile gas loop, OGL-1, installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Distribution of fission products in the graphite sleeves and blocks of the assemblies was measured by gamma-ray spectrometry. The 11th fuel assembly was aimed at testing the irradiation performance of mass product fuels in trial manufacturing of the first charge fuel for the High Temperature Engineering Test Reactor (HTTR) in relatively short irradiation, and the 12th assembly in long-term irradiation. The 12th assembly attained a burnup approximately as high as that of the HTTR driver fuel design. In the graphite sleeve of the 11th assembly, high concentration peaks of fission products were found in the axial distribution. Exposure of failed fuel particles was not detected on the surface of fuel compacts, while fissures of graphite matrix at overcoat boundaries were observed on the surface. These results led to a presumption that fission products, which were released from failed particles located inside of the fuel compact, was easily transported through the fissures of the matrix to the inner surface of the sleeve. In the graphite sleeve of the 12th assembly, 110mAg was detected together with other fission products of 137Cs, 134Cs etc. Silver-110m showed characteristic distribution: this nuclides was less concentrated at the region with highly concentrated 60Co which is presumed to have been transported from melted sheath material of thermocouples to the graphite sleeve. It was inferred from the distribution that the transport behavior of 110mAg had been influenced by co-sorption or by pore structure change in the graphite material of the sleeve, which had been induced by metallic elements including cobalt. (author)

  1. Distribution of fission products in graphite sleeves and blocks of the eleventh and twelfth OGL-1 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio; Fukuda, Kousaku; Kikuchi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tsuruta, Harumichi

    1994-06-01

    The 11th and 12th fuel assemblies were irradiated in an in-pile gas loop, OGL-1, installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Distribution of fission products in the graphite sleeves and blocks of the assemblies was measured by gamma-ray spectrometry. The 11th fuel assembly was aimed at testing the irradiation performance of mass product fuels in trial manufacturing of the first charge fuel for the High Temperature Engineering Test Reactor (HTTR) in relatively short irradiation, and the 12th assembly in long-term irradiation. The 12th assembly attained a burnup approximately as high as that of the HTTR driver fuel design. In the graphite sleeve of the 11th assembly, high concentration peaks of fission products were found in the axial distribution. Exposure of failed fuel particles was not detected on the surface of fuel compacts, while fissures of graphite matrix at overcoat boundaries were observed on the surface. These results led to a presumption that fission products, which were released from failed particles located inside of the fuel compact, was easily transported through the fissures of the matrix to the inner surface of the sleeve. In the graphite sleeve of the 12th assembly, {sup 110m}Ag was detected together with other fission products of {sup 137}Cs, {sup 134}Cs etc. Silver-110m showed characteristic distribution: this nuclides was less concentrated at the region with highly concentrated {sup 60}Co which is presumed to have been transported from melted sheath material of thermocouples to the graphite sleeve. It was inferred from the distribution that the transport behavior of {sup 110m}Ag had been influenced by co-sorption or by pore structure change in the graphite material of the sleeve, which had been induced by metallic elements including cobalt. (author).

  2. Thermal analysis model for the temperature distribution of the CANDU spent fuel assembly

    International Nuclear Information System (INIS)

    The purpose of this technical is to introduce the methodology and experimental process for the experimental research work with the mock-up test performed to verify and validate the MAXROT code which is a thermal analysis method for Wolsong (CANDU) spent fuel dry storage canister. The experiment was conducted simulating the heat transfer characteristics of combinations of equilateral triangular and square pitch arrays of heater rods, similar to a CANDU spent fuel bundle. After assembly of the heater rod bundle into the containment vessel, the experimental apparatus was operated under the same operating and boundary conditions as an interim dry storage condition at the nuclear power plant site. The reduced data from this experiment has been utilized to verity a model developed to predict the maximum fuel rod surface temperature in a fuel bundle. These test procedures and the experiment can be utilized to establish the fine thermal analysis method applicable to dry storage system for the spent fuel. 12 figs., 5 tabs., 36 refs. (Author) .new

  3. Swelling and creep observed in AISI 304 fuel pin cladding from three MOX fuel assemblies irradiated in EBR-II

    Science.gov (United States)

    Garner, F. A.; Makenas, B. J.; Chastain, S. A.

    2011-06-01

    Three 37-pin MOX-fueled experimental subassemblies were irradiated in EBR-II with fuel pin cladding constructed from annealed AISI 304 stainless steel. Analysis of the swelling and irradiation creep of the cladding showed that the terminal swelling rate of AISI 304 stainless steel appears to be ˜1%/dpa and that swelling is very reproducible for identical irradiation conditions. The swelling at a given neutron fluence is rather sensitive to both irradiation temperature and especially to the neutron flux, however, with the primary influence residing in the transient regime. As the neutron flux increases the duration of the transient regime is increased in agreement with other recent studies. The duration of the transient regime is also decreased by increasing irradiation temperature. In these assemblies swelling reached high levels rather quickly, reducing the opportunity for fuel pin cladding interaction and thereby reducing the contribution of irradiation creep to the total deformation. It also appears that in this swelling-before-creep scenario that the well-known "creep disappearance" phenomenon was operating strongly.

  4. The coolant quality effect on Zr fuel assembly reliability in NPP with RBMK-1000

    Energy Technology Data Exchange (ETDEWEB)

    Berezina, I.G.; Kritski, V.G.; Styazhkin, P.S. [All-Russian Design and Scientific Research Inst. of Complex Power Technology (VNIPIET), St. Petersburg (Russian Federation)

    2002-07-01

    1. A model was developed to describe the joint effect of physical and chemical parameters on corrosion of cladding zirconium alloys under LWR operation conditions. For the first time the influence of alloying components content, coolant impurities and the vapour content is taken into account. Verification of the model was carried out on the base of experimental and operation data on corrosion of zirconium claddings of fuel elements in research loops and at the NPP. 2. The chemical part of the model is based on the influence of temperature, pH{sub T} values and the concentration of hydrogen peroxide on solubility of corrosion products of zirconium alloys. The model describes corrosion of zirconium alloy during the entire fuel cycle at the NPP (under normal operation of fuel assemblies in the core and under of spent fuel storage in water ponds) and may be used to predict the influence of changing of the water chemistry quality on corrosion. 3. The model of defect formation accounting for the debris-effect and CILC effects was developed. Calculations showed that to achieve reliable operation of fuel assemblies of the RBMK (with cladding failures 2-3 pc/block in a year) at burn-up 32 MW.day/kg U, it is necessary to make standards for water chemistry more stringent. Limitation of the number of large particles and corrosion products causing the debris-effect is realized into proposed standards indirectly, through controlling the quality indices: X{sub 25}, Cl{sup -}, SiO{sub 3}{sup 2-}, Fe, Na, hardness. (author)

  5. Hanford MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site (SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. Hanford has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 1 facility. In all, a total of three LA MOX fuel fabrication options were identified by Hanford that could accommodate the program. In every case, only minor modification would be required to ready any of the facilities to accept the equipment necessary to accomplish the LA program

  6. Hanford MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site (SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. Hanford has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 1 facility. In all, a total of three LA MOX fuel fabrication options were identified by Hanford that could accommodate the program. In every case, only minor modification would be required to ready any of the facilities to accept the equipment necessary to accomplish the LA program.

  7. Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configuration with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)

  8. SRS MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site(SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. SRS has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 2 or 3 facility with storage of bulk PuO2 and assembly, storage, and shipping of fuel bundles in an S and S Category 1 facility. The total Category 1 approach, which is the recommended option, would be done in the 221-H Canyon Building. A facility that was never in service will be removed from one area, and a hardened wall will be constructed in another area to accommodate execution of the LA fuel fabrication. The non-Category 1 approach would require removal of process equipment in the FB-Line metal production and packaging glove boxes, which requires work in a contamination area. The Immobilization Hot Demonstration Program

  9. SRS MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site(SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. SRS has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 2 or 3 facility with storage of bulk PuO{sub 2} and assembly, storage, and shipping of fuel bundles in an S and S Category 1 facility. The total Category 1 approach, which is the recommended option, would be done in the 221-H Canyon Building. A facility that was never in service will be removed from one area, and a hardened wall will be constructed in another area to accommodate execution of the LA fuel fabrication. The non-Category 1 approach would require removal of process equipment in the FB-Line metal production and packaging glove boxes, which requires work in a contamination area. The Immobilization Hot Demonstration Program

  10. Determination of spent nuclear fuel assembly multiplication with the differential die-away self-interrogation instrument

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Alexis C. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States); Flaska, Marek; Pozzi, Sara A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States)

    2014-09-01

    We present a novel method for determining the multiplication of a spent nuclear fuel assembly with a Differential Die-Away Self-Interrogation (DDSI) instrument. The signal, which is primarily created by thermal neutrons, is measured with four {sup 3}He detector banks surrounding a spent fuel assembly. The Rossi-alpha distribution (RAD) at early times reflects coincident events from single fissions as well as fission chains. Because of this fact, the early time domain contains information about both the fissile material and spontaneous fission material in the assembly being measured. A single exponential function fit to the early time domain of the RAD has a die-away time proportional to the spent fuel assembly (SFA) multiplication. This correlation was tested by simulating assay of 44 different SFAs with the DDSI instrument. The SFA multiplication was determined with a variance of 0.7%.

  11. Supplemental description of ROSA-IV/LSTF with No.1 simulated fuel-rod assembly

    International Nuclear Information System (INIS)

    Forty-two integral simulation tests of PWR small break LOCA (loss-of-coolant accident) and transient were conducted at the ROSA-IV Large-Scale Test Facility (LSTF) with the No.1 simulated fuel-rod assembly between March 1985 and August 1988. Described in the report are supplemental information on modifications of the system hardware and measuring systems, results of system characteristics tests including the initial fluid mass inventory and heat loss distribution for the primary system, and thermal properties for the heater rod materials. These are necessary to establish the correct boundary conditions of each LSTF experiment with the No.1 core assembly in addition to the system data given in the system description report (JAERI-M 84-237). (author)

  12. A versatile passive and active non-destructive device for spent fuel assemblies monitoring

    International Nuclear Information System (INIS)

    The monitoring of spent fuel assemblies in reactor pools or in reprocessing plants with NDA methods is interesting (non-destructivity, non-intrusivity) for process control, safety-criticality and/or nuclear material management. In this context, the authors present the results of the development and design of a prototype device (physical methods used, qualification...) called PYTHON. The aim of PYTHON is to check the declared characteristic values of an irradiated assembly before taking it into a transport cask for safety criticality control. The PYTHON device consists of a detector head in two sections and a 252Cf source if active neutron counting is to be used. Each section of the detection head consists of two detectors: one fission chamber and one ionization chamber

  13. Influence of axial coolant flow on fuel assembly damping for the response to horizontal seismic loads

    International Nuclear Information System (INIS)

    The existence of a large damping increase under axial flow is clearly demonstrated by test results. Therefore this phenomenon concerns not only isolated tubes, but also rod bundles with multiple support such as fuel assemblies, and it may represent a large conservatism margin when not allowed for in their modelling. Test results also demonstrate that the mock-up behaviour is representative of that of a full-scale assembly, and that actual confinement conditions are not of much concern since their influence remains small. However, applying test results under axial flow to a core model can be envisaged only with caution, since no clear physical interpretation of this phenomenon has been found (at least for such large damping values). Further studies will comprise a complementary experimental program, in order to specify the physical nature and the application range of this effect, and to provide hints for a theoretical interpretation. (author)

  14. Expert System analysis of non-fuel assembly hardware and spent fuel disassembly hardware: Its generation and recommended disposal

    International Nuclear Information System (INIS)

    Almost all of the effort being expended on radioactive waste disposal in the United States is being focused on the disposal of spent Nuclear Fuel, with little consideration for other areas that will have to be disposed of in the same facilities. one area of radioactive waste that has not been addressed adequately because it is considered a secondary part of the waste issue is the disposal of the various Non-Fuel Bearing Components of the reactor core. These hardware components fall somewhat arbitrarily into two categories: Non-Fuel Assembly (NFA) hardware and Spent Fuel Disassembly (SFD) hardware. This work provides a detailed examination of the generation and disposal of NFA hardware and SFD hardware by the nuclear utilities of the United States as it relates to the Civilian Radioactive Waste Management Program. All available sources of data on NFA and SFD hardware are analyzed with particular emphasis given to the Characteristics Data Base developed by Oak Ridge National Laboratory and the characterization work performed by Pacific Northwest Laboratories and Rochester Gas ampersand Electric. An Expert System developed as a portion of this work is used to assist in the prediction of quantities of NFA hardware and SFD hardware that will be generated by the United States' utilities. Finally, the hardware waste management practices of the United Kingdom, France, Germany, Sweden, and Japan are studied for possible application to the disposal of domestic hardware wastes. As a result of this work, a general classification scheme for NFA and SFD hardware was developed. Only NFA and SFD hardware constructed of zircaloy and experiencing a burnup of less than 70,000 MWD/MTIHM and PWR control rods constructed of stainless steel are considered Low-Level Waste. All other hardware is classified as Greater-ThanClass-C waste

  15. Sparse-view neutron CT reconstruction of irradiated fuel assembly using total variation minimization with Poisson statistics

    International Nuclear Information System (INIS)

    We inspect the nuclear fuel assembly by demonstrating the potential use of sparse-view neutron computed tomography. The projection images of the fuel assembly were collected at the Idaho National Laboratory hot fuel examination facility using indirect foil-film transfer technique. The radiographs were digitized using a commercial film digitizer and registered spatially for reconstruction. Digitized data were reconstructed using simultaneous algebraic reconstruction technique (SART) with total variation minimization using a dual approach for numerical solution assuming the projection data are corrupted by Poisson noise. To validate and evaluate the performance of the algorithm, visual inspections, as well as quantitative evaluation studies using a computer simulation data and the experimental data of the fuel assembly were carried out. The proposed method provides better reconstruction for both simulated and experimental case in terms of artifact reduction, higher SNR, and better spatial resolution compared to the reconstruction yielded by filtered back projection and SART reconstruction. (author)

  16. Method for fixing a spring package to a top nozzle in a fuel assembly of a nuclear power reactor

    International Nuclear Information System (INIS)

    A method of fixing a spring package to a top nozzle of a fuel assembly of a nuclear reactor so as to press the fuel assembly against the bottom of the reactor core of the nuclear reactor, the fuel assembly including fuel rods, guide tubes and spacers arranged in a bundle between a top nozzle and a bottom nozzle, the method is described including the steps of: (a) welding a clamp to the top nozzle, (b) milling out a T-shaped slot in the clamp for receiving one end of the spring package with a close fit, (c) inserting said one end of the spring package into the slot and (d) fixing said one end of the spring package in the slot with a locking pin such that all moment forces from the spring package are taken up by the clamp and no moment forces are applied to the locking pin

  17. Final Report on IFA-10, the first Swedish Instrumented Fuel Assembly Irradiated in HBWR, Norway

    Energy Technology Data Exchange (ETDEWEB)

    Gyllander, J.Aa.

    1967-12-15

    A final report is given on IFA-10, the first Swedish instrumented fuel assembly irradiated in HBWR. The post-irradiation data are presented and correlated with the irradiation statistics. No bowing of the bundle was observed, no equi-axed grain growth was discernible, the fission gas release was very small, and the relative dimensional changes in length and diameter were of the order of magnitude 9 x 10{sup -4} The hydride content of the can increased from 35 ppm to 65 ppm and, in the contact point of the spacer, to 180 ppm.

  18. Fuel cleanup system for the tritium systems test assembly: design and experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kerr, E.C.; Bartlit, J.R.; Sherman, R.H.

    1980-01-01

    A major subsystem of the Tritium Systems Test Assembly is the Fuel Cleanup System (FCU) whose functons are to: (1) remove impurities in the form of argon and tritiated methane, water, and ammonia from the reactor exhaust stream and (2) recover tritium for reuse from the tritiated impurities. To do this, a hybrid cleanup system has been designed which utilizes and will test concurrently two differing technologies - one based on disposable, hot metal (U and Ti) getter beds and a second based on regenerable cryogenic asdorption beds followed by catalytic oxidation of impurities to DTO and stackable gases and freezout of the resultant DTO to recover essentially all tritium for reuse.

  19. Measurement of a fresh MOX-LWR type fuel assembly under water

    International Nuclear Information System (INIS)

    A fresh MOX-PWR type fuel assembly in a 17x17 configuration has been assayed under water with an adapted FORK type detector wherein fission chambers were replaced by 3He tubes. Shift register electronics were used, revealing total neutrons and real coincidence rate. The influence of pin removal was investigated at the periphery and in a central position. The boron concentration in the water was gradually increased to a maximum value of 2200 ppm and the influence on totals and reals was examined

  20. Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Durkee, Jr., Joe W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-06-19

    A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235U a% enrichments, burnups of 5, 10, 15, 20, 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134Cs/137Cs 134Cs/154Eu, and 154Eu/137Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the

  1. Implement of MOX fuel assemblies in the design of the fuel reload for a BWR; Implemento de ensambles de combustible MOX en el diseno de la recarga de combustible para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Enriquez C, P.; Ramirez S, J. R.; Alonso V, G.; Palacios H, J. C., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    At the present time the use of mixed oxides as nuclear fuel is a technology that has been implemented in mixed reloads of fuel for light water reactors. Due to the plutonium production in power reactors, is necessary to realize a study that presents the plutonium use like nuclear fuel. In this work a study is presented that has been carried out on the design of a fuel assembly with MOX to be proposed in the supply of a fuel reload. The fissile relationship of uranium to plutonium is presented for the design of the MOX assembly starting from plutonium recovered in the reprocessing of spent fuel and the comparison of the behavior of the infinite multiplication factor is presented and of the local power peak factor, parameters of great importance in the fuel assemblies design. The study object is a fuel assembly 10 x 10 GNF2 type for a boiling water reactor. The design of the fuel reload pattern giving fuel assemblies with MOX, so the comparison of the behavior of the stop margin for a fuel reload with UO{sub 2} and a mixed reload, implementing 12 and 16 fuel assemblies with MOX are presented. The results show that the implement of fuel assemblies with MOX in a BWR is possible, but this type of fuels creates new problems that are necessary to study with more detail. In the development of this work the calculus tools were the codes: INTREPIN-3, CASMO-4, CMSLINK and SIMULATE-3. (Author)

  2. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    International Nuclear Information System (INIS)

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET

  3. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

  4. Monte Carlo simulations of differential die-away instrument for determination of fissile content in spent fuel assemblies

    Science.gov (United States)

    Lee, Tae-Hoon; Menlove, Howard O.; Swinhoe, Martyn T.; Tobin, Stephen J.

    2011-10-01

    The differential die-away (DDA) technique has been simulated by using the MCNPX code to quantify its capability of measuring the fissile content in spent fuel assemblies. For 64 different spent fuel cases of various initial enrichment, burnup and cooling time, the count rate and signal to background ratios of the DDA system were obtained, where neutron backgrounds are mainly coming from the 244Cm of the spent fuel. To quantify the total fissile mass of spent fuel, a concept of the effective 239Pu mass was introduced by weighing the relative contribution to the signal of 235U and 241Pu compared to 239Pu and the calibration curves of DDA count rate vs. 239Pu eff were obtained by using the MCNPX code. With a deuterium-tritium (DT) neutron generator of 10 9 n/s strength, signal to background ratios of sufficient magnitude are acquired for a DDA system with the spent fuel assembly in water.

  5. Thermomechanical evaluation of the fuel assemblies fabricated in the ININ; Evaluacion termomecanica de los ensambles combustibles fabricados en el ININ

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2005-07-01

    The pilot plant of fuel production of the National Institute of Nuclear Research (ININ) provided to the Laguna Verde Nuclear Power Plant (CNLV) four fuel assemblies type GE9B. The fuel irradiation was carried out in the unit 1 of the CNLV during four operation cycles, highlighting the fact that in their third cycle the four assemblies were placed in the center of the reactor core. In the Nuclear Systems Department (DSN) of the ININ it has been carried out studies to evaluate their neutron performance and to be able to determine the exposure levels of this fuels. Its also outlines the necessity to carry out a study of the thermomechanical behavior of the fuel rods that compose the assemblies, through computational codes that simulate their performance so much thermal as mechanical. For such purpose has been developing in the DSN the FETMA code, together with the codes that compose the system Fuel Management System (FMS), which evaluates the thermomechanical performance of fuel elements. In this work were used the FETMA and FEMAXI codes (developed by JAERI) to study the thermomechanical performance of the fuel elements manufactured in the ININ. (Author)

  6. Rail Shock and Vibration Pre-Test Modeling of a Used Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B.; Klymyshyn, Nicholas A.; Jensen, Philip J.; Best, Ralph E.; Maheras, Steven J.; McConnell, Paul E.; Orchard, John

    2015-04-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste (HLW). The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and HLW generated by existing and future nuclear fuel cycles. The Storage and Transportation staff within the UFDC is responsible for addressing issues regarding the long-term or extended storage (ES) of UNF and its subsequent transportation. Available information is not sufficient to determine the ability of ES UNF, including high-burnup fuel, to withstand shock and vibration forces that could occur when the UNF is shipped by rail from nuclear power plant sites to a storage or disposal facility. There are three major gaps in the available information – 1) the forces that UNF assemblies would be subjected to when transported by rail, 2) the mechanical characteristics of fuel rod cladding, which is an essential structure for controlling the geometry of the UNF, a safety related feature, and 3) modeling methodologies to evaluate multiple possible degradation or damage mechanisms over the UNF lifetime. In order to address the first gap, options for tests to determine the physical response of surrogate UNF assemblies subjected to shock and vibration forces that are expected to be experienced during normal conditions of transportation (NCT) by rail must be identified and evaluated. The objective of the rail shock and vibration tests is to obtain data that will help researchers understand the mechanical loads that ES UNF assemblies would be subjected to under normal conditions of transportation and to fortify the computer modeling that will be necessary to evaluate the impact

  7. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    Science.gov (United States)

    LaFleur, Adrienne M.; Menlove, Howard O.

    2015-05-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies.

  8. Post-Irradiation Examination Test for an Evaluation of In-Core Performance of the Parts of Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. S.; Ryu, W. S.; Choo, Y. S. (and others)

    2007-11-15

    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 report are used to produce the irradiation data of the grid 1x1 cell spring with fuel cladding, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the guide tube. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 270 {approx} 375 .deg. C up to a fast neutron fluence of 5.7x10{sup 20} n/cm{sup 2}(E>1 MeV). From the spring test of mid grid 1x1 cell and bottom grid plate, the irradiation effects were not found. The irradiation effects on the irradiation growth also were not found. The buckling load of mid grid 1x1 cell tends to increase due to a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. Through these tests of the 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.

  9. Development of whole core thermal-hydraulic analysis program ACT. 4. Simplified fuel assembly model and parallelization by MPI

    International Nuclear Information System (INIS)

    A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including the effect of the flow between wrapper-tube walls (inter-wrapper flow) under various reactor operation conditions. As appropriate boundary conditions in addition to a detailed modeling of the core are essential for accurate simulations of in-core thermal hydraulics, ACT consists of not only fuel assembly and inter-wrapper flow analysis modules but also a heat transport system analysis module that gives response of the plant dynamics to the core model. This report describes incorporation of a simplified model to the fuel assembly analysis module and program parallelization by a message passing method toward large-scale simulations. ACT has a fuel assembly analysis module which can simulate a whole fuel pin bundle in each fuel assembly of the core and, however, it may take much CPU time for a large-scale core simulation. Therefore, a simplified fuel assembly model that is thermal-hydraulically equivalent to the detailed one has been incorporated in order to save the simulation time and resources. This simplified model is applied to several parts of fuel assemblies in a core where the detailed simulation results are not required. With regard to the program parallelization, the calculation load and the data flow of ACT were analyzed and the optimum parallelization has been done including the improvement of the numerical simulation algorithm of ACT. Message Passing Interface (MPI) is applied to data communication between processes and synchronization in parallel calculations. Parallelized ACT was verified through a comparison simulation with the original one. In addition to the above works, input manuals of the core analysis module and the heat transport system analysis module have been prepared. (author)

  10. LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

    International Nuclear Information System (INIS)

    The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium

  11. Loss of stability and possible bending shape of WWER-1000 fuel assemblies guide tubes

    International Nuclear Information System (INIS)

    The paper defines the critical value of compressive force and the corresponding possible bending shapes of WWER-1000 fuel assemblies guide tubes resulting from loss of stability during operation. The guide tubes are considered as the long rods that are bent due to the loss of stability under the longitudinal compressive force. The three design schemes of the guide tubes fixations with the fuel assembly top and bottom nozzles corresponding to the motionless and sliding cylindrical joints and rigid fixations are considered in this paper. The minimal values of the guide tubes critical compressive force correspondent to loss of stability are obtained as 1.87 MPa for the both edges cylindrical joints fixing, 3.83 MPa for the cylindrical joint fixing of the one edge and the rigid fixing of the another edge, 7.49 MPa for the both edges rigid fixing. The relatively small values of the critical compressive force show that the shape of guide tube bent due to loss of stability can include several bending points leading to increased time for drop of absorber element cluster to the core bottom and incompliance with reactor normal operation conditions.

  12. An anisotropic numerical model for thermal hydraulic analyses: application to liquid metal flow in fuel assemblies

    Science.gov (United States)

    Vitillo, F.; Vitale Di Maio, D.; Galati, C.; Caruso, G.

    2015-11-01

    A CFD analysis has been carried out to study the thermal-hydraulic behavior of liquid metal coolant in a fuel assembly of triangular lattice. In order to obtain fast and accurate results, the isotropic two-equation RANS approach is often used in nuclear engineering applications. A different approach is provided by Non-Linear Eddy Viscosity Models (NLEVM), which try to take into account anisotropic effects by a nonlinear formulation of the Reynolds stress tensor. This approach is very promising, as it results in a very good numerical behavior and in a potentially better fluid flow description than classical isotropic models. An Anisotropic Shear Stress Transport (ASST) model, implemented into a commercial software, has been applied in previous studies, showing very trustful results for a large variety of flows and applications. In the paper, the ASST model has been used to perform an analysis of the fluid flow inside the fuel assembly of the ALFRED lead cooled fast reactor. Then, a comparison between the results of wall-resolved conjugated heat transfer computations and the results of a decoupled analysis using a suitable thermal wall-function previously implemented into the solver has been performed and presented.

  13. LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

    Energy Technology Data Exchange (ETDEWEB)

    Bronson, M.C.

    1997-10-01

    The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium.

  14. Experimental study of pressure drops through LOCA-generated debris deposited on a fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jeong Kwan, E-mail: jksuh@khnp.co.kr [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Kim, Jae Won; Kwon, Sun Guk; Lee, Jae Yong [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Cho, Hyoung Kyu; Park, Goon Cherl [Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of)

    2015-08-15

    Highlights: • In-vessel downstream effect tests were performed in the presence of LOCA-generated debris. • Available driving heads under each LOCA scenario were verified using experimental data. • Fibrous debris was prepared to satisfy the length distribution obtained from the bypass test. • Limiting test conditions were identified through sensitivity studies. - Abstract: Under post loss-of-coolant accident (LOCA) conditions, it is postulated that debris can be generated and transported to the containment sump strainer. Some of the debris may pass through the strainer and could challenge the long-term core cooling capability of the plant. To address this safety issue, in-vessel downstream effect tests for the advanced power reactor (APR) 1400 were performed. Fibrous debris is the most crucial material in terms of causing pressure drops, and was prepared in this study to satisfy the fiber length distribution obtained through a strainer bypass test. Sensitivity studies on pressure drops through LOCA-generated debris deposited on a fuel assembly were performed to evaluate the effects of water chemistry and fiber length distribution. The pressure drops with debris laden pure water were substantially less than those with debris laden ordinary tap water. The experiment with fiber length distribution suggested by WCAP-16793 showed lower pressure drops than those with the APR1400 specific fiber length distribution. All the experimental results showed that the pressure drops in the mock-up fuel assembly were less than the available driving head at each LOCA scenario.

  15. An estimate of the reactivity of assemblies of NRX fuel elements in light water

    International Nuclear Information System (INIS)

    This report contains calculations on the criticality of assemblies of NRX fuel elements in light water. The elements are dealt with in three sections, 'X rods' of natural uranium, enriched elements of U235/A1 alloy and enriched elements of Pu/Al alloy. Values of k∞ and B2 are provided for two fuel concentrations for each of the two enriched types and for a range of irradiations of the X rods. The calculations for the X rods provide maximum and minimum values of k∞. The maximum values for some lattices are a few per cent above unity. Unfortunately, the present experimental evidence does not prove that it is impossible to achieve values of k∞ greater than unity in lattices of natural uranium in light water. Hence for safety predictions maximum values have been used. The resulting restrictions are not very severe. It is possible to make critical assemblies of the enriched elements, Part (5) contains a set of recommended minimum spacings such that elements of all kinds may safely be mixed in a stack together. There are also predictions of the minimum critical numbers of complete elements or elements cut into slugs. (author)

  16. Evaluation of the in-pile pressure data from instrumented fuel assemblies IFA-431 and IFA-432

    International Nuclear Information System (INIS)

    This report includes results of the examination of the in-pile pressure data from instrumented test assemblies IFA-431 and 432. The pressure data have been used to estimate the fission gas release fraction as a function of fuel burnup. Included are comparisons of the estimated release functions and those predicted by three fission gas release models using the experimental temperature histories of the fuel rods. These comparisons show that fuel temperature is the primary factor in determining fission gas release and that burnup-enhanced fission gas release is not important in UO2 fuels irradiated to 1700 GJ/kgU

  17. Validation of SWAT for burnup credit problems by analysis of post irradiation examination of 17*17 PWR fuel assembly

    International Nuclear Information System (INIS)

    For adopting burnup credit in transport or storage of spent fuel (SF), development of a reliable burnup calculation code is crucial. For this purpose, data of Post Irradiation Examination (PIE) have been extensively analyzed to evaluate accuracy of burnup calculation codes for a 14*14 or 15*15 PWR fuel assembly. This study shows results of analysis of this latest PIE with SWAT and ORIGEN2.1. SWAT is an integrated burnup code system for a 17*17 PWR fuel assembly that has been developed by Tohoku University and JAERI. The results show that SWAT can more precisely predict nuclide composition of latest PWR assembly than ORIGEN2.1. (O.M.)

  18. The PS Booster hits 40

    CERN Multimedia

    Joannah Caborn Wengler

    2012-01-01

    Many accelerators’ "round" birthdays are being celebrated at CERN these days – the PS turned 50 in 2009, the SPS was 35 in 2011, and this year it's the turn of the PS Booster to mark its 40th anniversary. Originally designed to accelerate 1013 protons to 800 MeV, it has far exceeded its initial design performance over the years.   The PS Booster in the 1970s. Imagine the scene: a group of accelerator physicists staring expectantly at a monitor, when suddenly a shout of joy goes up as a signal flickers across the screen. Does that sound familiar? Well, turn the clock back 40 years (longer hair, wider trouser legs) and you have the situation at the PS Booster on 26 May 1972. On that day, beam was injected into the Booster for the first time. “It was a real buzz,” says Heribert Koziol, then Chairman of the Running-in Committee. “We were very happy – and also a little relieved – when the beam finally...

  19. FNAL booster: Experiment and modeling

    Energy Technology Data Exchange (ETDEWEB)

    Panagiotis Spentzouris; James Amundson

    2003-06-02

    We present measurements of transverse and longitudinal beam phase space evolution during the first two hundred turns of the FNAL Booster cycle. We discuss the experimental technique, which allowed us to obtain turn-by-turn measurements of the beam profile. The experimental results are compared with the prediction of the Synergia 3D space charge simulation code.

  20. Spent fuel dry storage technology development: thermal evaluation of three adjacent drywells (each containing a 0.6 kW PWR spent fuel assembly)

    International Nuclear Information System (INIS)

    A spent fuel Adjacent Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site utilizing three nearly identical pressurized water reactor spent fuel assemblies each having a decay heat level of approximately 0.6 kW. Each fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in an instrumented near-surface drywell storage cell for thermal testing. Each fuel assembly was sealed inside a 14-in. diam, 168-in.-long stainless steel canister and attached to a concrete-filled, 20-in.-diam, 34-in.-long, shield plug. The canister assembly was then placed in a carbon steel drywell liner which had been grouted into a hole drilled in the soil adjacent to E-MAD. The three drywells were located 25 feet apart in a linear array. Thermocouples, provided to measure canister, liner and soil temperatures, were inserted into tubes on the outside of the canister and drywell liner and were attached to plastic pipes which were grouted into holes in the soil. Temperatures from the three drywells and the adjacent soil were recorded throughout the Adjacent Drywell Test. Drywell thermal data showed virtually no thermal interaction between adjacent drywells. However, peak temperatures reached by the three drywells did show a fairly significant difference. Peak canister and drywell liner temperatures were reached in August 1981 for all three drywells. The two previously unused drywells responded similarly with peak canister and liner temperatures reaching 1990F and 1580F, respectively. Comparable peak temperatures for the third drywell which had previously contained spent fuel for nearly 21 months prior to the Adjacent Drywell Test reached 2100F for the canister and 1690F for the drywell liner. This difference is attributed to a decrease in soil thermal conductivity caused by the dryout of soil around the drywell used for previous spent fuel testing

  1. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  2. AREVA NP's advanced Thermal Hydraulic Methods for Reactor Core and Fuel Assembly Design

    International Nuclear Information System (INIS)

    The main objective of the Thermal Hydraulic (TH) analysis of reactor core and fuel assembly design is the determination of pressure loss and critical heat flux (CHF). Especially the description of the latter effect requires the modeling of a large variety of physical phenomena starting with single phase quantities like turbulence or fluid-wall friction, two phase quantities like void distributions, heat transfer between fuel rod and fluid and ultimately the CHF mechanism itself. Additional complexity is added by the fact that the relevant geometric scales which have to be resolved, cover a wide range from the length of the fuel assembly (∼ 4000 mm), over the typical dimensions of sub-channel cross sections and the vanes on the spacer grids (∼ 10 mm) down to the microscopic scales set by bubble sizes and boundary layers (mm to sub mm). Due to the above described situation the necessary TH quantities are often determined by measurements. The main advantage of this technique is that measurements are widely accepted and trusted if the geometry and flow conditions are sufficiently close to real reactor conditions. The main disadvantage of experiments is that they are expensive both with respect to time and money; especially in high pressure tests they give only limited access to the test object. Consequently there is a strong interest to develop computer codes with the goal of minimizing the need of experiments, and hence, speeding up and reducing costs of fuel assembly and core design. Today most of the design work is based on sub-channel codes, originally developed in the 70's; they provide an effective description of the TH in fuel assemblies by regarding the fuel assembly as a system of communicating channels (the volume enclosed by four fuel rods = one sub-channel). Further development of these codes is one main focus of AREVA NP's Thermal Hydraulic method and code development strategy. To focus the know-how and resources existing in the different regions of

  3. Evaluation of water condition change for long term integrity of fuel assemblies in Fukushima Daiichi NPPs common pool

    International Nuclear Information System (INIS)

    The Spent Fuel Pools (SFPs) of the Fukushima Daiichi nuclear power plants (1F-NPPs) were subjected to seawater injection and fallen rubble from the damaged reactor buildings by the Great East Japan Earthquake on March 11, 2011. The spent fuel assemblies in SFPs have been transported to a common pool. It is necessary to evaluate the long-term integrity of transported fuel assemblies in the common pool. In this study, the effect of fuel assemblies transported from SFPs on change in the water condition of the common pool has been evaluated. For the evaluation, leaching tests of seawater elements, concrete elements and fission products (FPs) were performed and the leaching behavior was evaluated. Chloride, sodium and calcium ions were detected by leaching tests of rubble. Cesium and iodine ions were detected by FP leaching tests. The equation of cumulative dissolution amount and dissolution rates of chloride ion and iodine ion were calculated using the leaching test results. The leaching equations are proposed in order to evaluate the effect of fuel assemblies transported from SFPs on change in the water condition of the common pool. (author)

  4. Booster for African Economy

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    China’s investment is fueling African growth SINCE 2000,driven by the Forum on China-Africa Cooperation,China’s foreign direct investment(FDI) in Africa has been growing rapidly.In the face of the global financial crisis,which led to global FDI flows falling,China’s investment in Africa has been on a steady, upbeat rise without any interruption.In 2009,China’s direct investment in Africa reached $1.44 billion,of which nonfinancial direct investment soared by 55.4 percent from the previous year.Africa

  5. Waste classification of 17 x 17 KOFA spent fuel assembly hardware

    International Nuclear Information System (INIS)

    Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a low and intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17 x 17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGENS module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and 90Sr, respectively. Finally, it was found that 88.7% of the metal waste from the 17 x 17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing

  6. Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Bikeev, Artem S.; Bolshagin, Sergey N.; Kalugin, Mikhail A.; Kosourov, Evgeniy K.; Pavlovichev, Aleksandr M.; Pryanichnikov, Aleksandr V.; Sukhino-Khomenko, Evgenia A.; Shcherenko, Anna I.; Shcherenko, Anastasia I.; Shkarovskiy, Denis A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    Two types of calculations were made to compare BIPR-7A and MCU results for 3D full-scale models. First EPS (emergency protection system) efficiency and in-core power distributions were analyzed for an equilibrium fuel load of VVER-1000 assuming its operation within an 18-month cycle. Computations were performed without feedbacks and with fuel burnup distributed over the core. After 3D infinite lattices of full-scale VVER-1000 fuel assemblies (A's) with uranium fuel 4.4% enrichment and uranium-erbium fuel 4.4% enrichment and Er{sub 2}O{sub 3} 1 % wt were considered. Computations were performed with feedbacks and fuel burnup at the constant power level. For different time moments effective multiplication factor and power distribution were obtained. EPS efficiency and reactivity effects at chosen time moments were analyzed.

  7. Process development for fabrication of zircaloy- 4 of dissolver assembly for spent nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Spent fuel reprocessing for fast breeder reactor (FBR) requires a dissolver made of a material which has resistance to corrosion as the process involves Nitric Acid as the process medium. Various materials to achieve minimum corrosion rates have been tried for this operation. Particularly the focus was on the use of advanced materials with high performance (corrosion rate and product life) for high concentrations greater than 8 N and temperatures (boiling and vapour) of Nitric Acid employed in the dissolver unit. The different commercially available materials like SS316L , Pure Titanium, Ti - 5% Ta and Ti - 5% Ta - 1.8% Nb were tried and the corrosion behavior of these materials was studied in detail. As this is continuous process of evolution of new materials, it was decided to try out zircaloy - 4 as the material of construction for construction due to its excellent corrosion resistance properties in Nitric Acid environment. The specifications were stringent and the geometrical configurations of the assembly were very intricate in shape. On accepting the challenge of fabrication of dissolver, NFC has made different fixtures for Electron Beam Welding and TIG Welding. Various trials were carried out for optimization of various operating parameter like beam current, Acceleration voltage, welding speed to get adequate weld penetration. Both EB welding and TIG welding process were standardized and qualified by carrying out a number of trials and testing these welds by various weld qualification procedures like radiography, Liquid dye penetrant testing etc. for different intricate weld geometries. All the welds were simulated with samples to optimize the weld parameters. Tests such as include metallographic (for microstructure and HAZ), mechanical (for weld strength) and chemical (material analysis for gases) were conducted and all the weld samples met the acceptable criteria. Finally the dissolver was made meeting stringent specifications. All the welds were checked

  8. The use of the BIC set in the characterization of used nuclear fuel assemblies by nondestructive assay

    International Nuclear Information System (INIS)

    Highlights: • The NDA of used enriched-uranium fuel assemblies is a three-dimensional problem. • At least three independent NDA measurements are necessary for accurate assay. • The BIC set is the burnup, Initial enrichment, and cooling time of a used assembly. • The BIC variables are independent with respect to the physics and isotopics. • The BIC set characterizes used enriched-uranium fuel assemblies to first order. - Abstract: This paper explains why the burnup, initial enrichment, and cooling time of a used fuel assembly – collectively called the BIC set of variables – characterize it to first order for the purposes of nuclear-materials safeguards and burnup credit. From an analysis by basic nuclear engineering, it is shown that the physical properties and the isotopic content of a used fuel assembly are basically three-dimensional vector spaces. By extensive referencing of the NDA literature, the paper then shows that the BIC variables are independent variables with respect to the physical properties and the isotopes. Therefore, the knowledge of all three BIC variables is a necessary condition for the accurate characterization of a used low- or high-enriched uranium (LEU or HEU) fuel assembly. For a plutonium mixed-oxide (MOX) fuel assembly, a fourth variable for the BIC set (the curium-producing ability) is also necessary. The paper also discusses other possible variables besides the BIC set, to demonstrate that the knowledge of the BIC set is also a sufficient condition in many cases. Logically, it is therefore necessary to make at least three independent NDA measurements (or four, for MOX) to achieve a unique solution (characterization) if a reliance on information provided by the reactor operator is to be avoided. By this fact, the common question, “What is the accuracy of a particular NDA technique?” is revealed to be a poorly posed one with regard to used fuel assemblies. The result of the paper is a better paradigm for interpreting and

  9. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  10. CFD analysis of coolant channel geometries for a tightly packed fuel rods assembly of Super FBR

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Rui, E-mail: guorui@asagi.waseda.jp; Oka, Yoshiaki

    2015-07-15

    Highlights: • Supercritical water heat transfer is validated against the experimental data. • Thermal hydraulic performance of three coolant channel cross-sectional geometries are analyzed. • Geometry B is superior to the other two geometries. - Abstract: This paper presents the CFD investigation on three cross-sectional geometries of coolant channels in the newly designed tightly packed fuel rods assembly for high breeding of a supercritical-pressure light water cooled fast breeding reactor (Super FBR). The calculations addressed the turbulence models and mesh conditions validated against the experimental data. The cladding temperatures and pressure drop are compared. It is concluded that geometry B (triangle with round corner) is superior to the other two (round and triangle with sharp corner) due to its excellent thermal hydraulic characteristics.

  11. Fusion fuel purification during the Tritium Systems Test Assembly 3-week loop experiment

    International Nuclear Information System (INIS)

    During the time period from April 19, 1989--May 5, 1989, the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL) conducted its longest continuous integrated loop operation to date. This provided an opportunity to test some hitherto unproven capabilities of the TSTA Fuel Cleanup System (FCU). Previous FCU tests were reported. The purpose of the FCU is to remove impurities from a stream of hydrogen isotopes (Q2) representative of torus exhaust gas. During this run impurities loadings ranging from 60 to 179 sccm of 90% N2 and 10% CH4 were fed to the FCU. Each of the two FCU main flow molecular sieve beds (MSB's) were filled to breakthrough three times. The MSB's were regenerated during loop operations. 2 refs., 6 figs., 2 tabs

  12. Partial flow blockage effects within a (liquid metal cooled fast reactor) LMFBR fuel assembly

    International Nuclear Information System (INIS)

    A lumped thermal-hydraulic model was used to calculate the increase in the sodium and cladding temperatures in the wake behind a non-porous partial flow blockage within a typical LMFBR fuel rod assembly. The model predicts that over 25 percent of the cross sectional flow area may be blocked before the wake fluid temperature reaches boiling; the actual size depends on the blockage axial location and radial location. Agreement with the limited sodium flow rod bundle blockage data is achieved by the model if the wide variation observed in the experimental cladding temperatures within the wake region is attributed to variations in local heat transfer coefficients. (29 references) (U.S.)

  13. Final report: Seven-layer membrane electrode assembly - an innovative approach to PEM fuel cell design

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, A.

    2005-07-01

    Costs of materials and fabrication, rather than appropriateness of technology, are the major barriers to the sales of fuel cells. With the objective of reducing costs, potential alternative component materials for (a) the fluid flow plate (FFP) and (b) the gas diffusion layers were investigated. The concept of a 7-layer membrane electrode assembly (MEA), in which components are bonded into a unitised module, was also studied. The advantages of the bonded cell, and the flow field design, are expounded. Low-cost carbon particle composites were developed for the FFPs. The modular 7-layer MEA has an order of magnitude saving over current materials. Overall, the study has led to a greater volumetric power output, lower costs and greater reliability. The work was carried out by Morgan Group Technology Limited and funded by the DTI.

  14. Development of fuel assembly seismic analysis against vertical and horizontal earthquake

    Energy Technology Data Exchange (ETDEWEB)

    Sato, T.; Akitake, J.; Kobayashi, H. [Nuclear Development Corporation, Ibaraki (Japan); Azumi, S. [Kansai Electric Power co., inc., Osaka (Japan); Koike, H.; Takeda, N.; Suzuki, S. [Kobe Shipyard and Machinery Works, Mitsubishi Heavy Industries, LTD., Kobe (Japan)

    2001-07-01

    Vertical vibration with large acceleration was observed in KOBE earthquake in 1995. Concerning PWR fuel assembly, though the vertical response has been calculated by a static analysis, it had better be calculated by a dynamic analysis in detail. Furthermore, mutual effects between horizontal and vertical motions attract our attention. For these reasons, a dynamic analysis method in the vertical direction was developed and linked with the previously developed method in the horizontal direction. This is the method that takes effect of vertical vibration into the horizontal vibration analysis as the change of horizontal stiffness, which is brought by axial compressive force. In this paper, fundamental test results for developing the method are introduced and summary of the advanced method's procedure and analysis results are also described. (authors)

  15. Numerical Simulation for Flow Distribution in ACE7 Fuel Assemblies affected by a Spacer Grid Deformation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jongpil; Jeong, Ji Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In spite of various efforts to understand hydraulic phenomena in a rod bundle containing deformed rods due to swelling and/or ballooning of clad, the studies for flow blockage due to spacer grid deformation have been limited. In the present work, 3D CFD analysis for flow blockage was performed to evaluate coolant flow within ACE7 fuel assemblies (FAs) containing a FA affected by a spacer grid deformation. The real geometry except for inner grids was used in the simulation and the region including inner grid was replaced by porous media. In the present work, the numerical simulation was performed to predict coolant flow within ACE7 FAs affected by a Mid grid deformation. The 3D CFD result shows that approximately 60 subchannel hydraulic diameter is required to fully recover coolant flow under normal operating condition.

  16. Numerical simulation of water flow through the bottom en piece of a nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Moyses A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mail: navarro@cdtn.br; Santos, Andre A. Campagnole dos [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Mecanica]. E-mail: acampagnole@yahoo.com.br; Gomes, Sydney da Silva; Brittes, Luiz Henrique A. [Industrias Nucleares do Brasil S.A. (INB), Reende, RJ (Brazil)]. E-mails: sydney@inb.gov.br; brittes@inb.gov.br

    2007-07-01

    The water flow through the bottom nozzle of a nuclear fuel assembly was simulated using a commercial CFD code, CFX 10.0. Previously, simulations with a perforated plate similar to the bottom nozzle plate were performed to define the appropriate mesh refinement and turbulence model ({kappa}-{epsilon} or SST). Subsequently, the numerical simulation was performed with the optimized mesh using the turbulence model ({kappa}-{epsilon} in a standard bottom nozzle with some geometric simplifications. The numerical results were compared with experimental results to determine the pressure drop through the bottom nozzle in the Reynolds range from {approx}10500 to {approx}95000. The agreement between the numerical simulations and experimental results may be considered satisfactory. The study indicated that the CFD codes can play an important role in the development of pieces with complex geometries, optimizing the planning of the experiments and aiding in the experimental analysis. (author)

  17. Review of thermohydraulic research of fuel assemblies with partial blocking of flow cross-section

    International Nuclear Information System (INIS)

    A review is presented of the theoretical and experimental investigation of blockage formation and of velocity and temperature fields in fuel rod bundles with partial blockage of the flow section. The temperature and velocity fields in cases of flow blockage are analyzed and the range of the recirculation zone length behind the blockage is shown. Formulas for the evaluation of the coolant flow rate changes and the temperature increments in dependence on the operating parameters and the blocked area are given. Questions of blockage identification and of prevention of emergency situations are discussed. Results of the analysis emphasize the necessity to continue research of blockage formation problems and of velocity and temperature conditions in blocked assemblies. (author)

  18. Nuclear data uncertainty and sensitivity analysis with XSUSA for fuel assembly depletion calculations

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, W.; Aures, A.; Gallner, L.; Hannstein, V.; Krazykacz-Hausmann, B.; Velkov, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany); Martinez, J. S. [Dept. of Nuclear Engineering, Universidad Politecnica de Madrid, Madrid (Spain)

    2014-06-15

    Uncertainty and sensitivity analyses with respect to nuclear data are performed with depletion calculations for BWR and PWR fuel assemblies specified in the framework of the UAM-LWR Benchmark Phase II. For this, the GRS sampling based tool XSUSA is employed together with the TRITON depletion sequences from the SCALE 6.1 code system. Uncertainties for multiplication factors and nuclide inventories are determined, as well as the main contributors to these result uncertainties by calculating importance indicators. The corresponding neutron transport calculations are performed with the deterministic discrete-ordinates code NEWT. In addition, the Monte Carlo code KENO in multi-group mode is used to demonstrate a method with which the number of neutron histories per calculation run can be substantially reduced as compared to that in a calculation for the nominal case without uncertainties, while uncertainties and sensitivities are obtained with almost the same accuracy.

  19. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    International Nuclear Information System (INIS)

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties

  20. CAST3M modelling of a spent fuel assembly bending during a handling accident Rod failure risk evaluation from the experimental results of spent fuel rod bending test

    International Nuclear Information System (INIS)

    The fuel handling operating rules exclude any accidental risk. However in the framework of the PRECCI R and D project, the bending of a spent fuel assembly resulting from its locking during a translation displacement is taken into account. This enabled us to develop an approach based on experiments and calculations that allows us to simulate the behaviour of an assembly under such loading. This study was carried out in CEA laboratories with the funding and the technical support of EDF. A three points bending test on a spent fuel rod segment was performed at the Laboratory for Mechanical Behaviour of Irradiated Materials (LCMI). From the experimental strength-displacement curve, a maximum failure strain, a maximum failure curvature and an equivalent constitutive equation were determined. CAST3M modelling of the fuel rod taking into account the elasto-plastic behaviour of the clad and the cracking of the UO2 fuel pellets was verified by the experimental results. Consequently, the identification of the respective contributions of the clad and of the pellets to the rod global behaviour was made possible. A two dimensional assembly with beam elements was modelled with CAST3M. The properties of the beams modelling the different parts of the assembly (top and bottom nozzle, grids) were chosen and adjusted according to their materials (zirconium alloys, steel) in order to obtain stiffness, tensile and shear behaviour, sliding and holding functions close to the experimental ones. Assembly bending calculations were performed. In order to obtain a rod integrity estimator, their maximum calculated strains and curvatures as a function of the bending angles can be compared to the failure experimental ones. (authors)

  1. Study of fuel assemblies for the nuclear reactor GFR; Estudio de ensambles de combustible para el reactor nuclear GFR

    Energy Technology Data Exchange (ETDEWEB)

    Reyes R, R.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: ricarera@yahoo.com.mx

    2008-07-01

    In the present work the criticality calculations for two models of fuel assembly were realized to study the nuclear reactor cooled by gas (Gas Fast Reactor) of IV Generation. Model 1 is an assembly with hexagonal adjustment of fuel rods with reflector in the axial ends higher and lower, the coolant flows between the rods. Model 2 is an hexagonal assembly type block with spheres dispersion and cylindrical channels for where the coolant with reflector in the axial ends also flows. The materials selected for each component of the assemblies, should be resistant to the radiation of fast neutrons and high operation temperatures, for what in both models the following materials were chosen: a mixture of uranium carbide more plutonium for the fuel; a mixture of silicon carbide in different theoretical density percentages for structures and shieldings; helium gas like coolant and a zirconium carbide mixture like reflector, which fulfill the restrictions of being resistant to the high operation temperatures and means of irradiation. General considerations were taken, which are common parameters to both types of assemblies, like size and materials used in the different parts of each model of assembly. The criticality calculations were obtained with the help of the MCNPx code, based on the Monte Carlo method. It was realized a validation of the atomic density data of each component of the assemblies, to have the certainty of the proportionate values that they were introduced of correct way in the code. The results show that model 1 makes better use of the fissile material in a assembly that has the same dimensions externally. That is to say, that from the only considered viewpoint, the neutron one, model 1 is better than model 2. (Author)

  2. AREVA NP Inc next generation fresh UO2 fuel assembly shipping cask: SCALE - CRISTAL comparisons lead to safety criticality confidence

    International Nuclear Information System (INIS)

    AREVA NP as a worldwide PWR fuel provider has to have a fleet of fresh UO2 shipping casks being agreed within a lot of countries including USA (France), Germany, Belgium, Sweden, China, and South Africa... and to accommodate foreseen EPR Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector decided to develop an up to date shipping cask gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and local Safety Authorities. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: -Preferential flooding - Fuel rod lattice pitch expansion for full length of fuel assemblies - Neutron absorber penalty -... Well known computer codes, American SCALE package and French CRISTAL package, were used to check configurations reactivity and to ensure that both codes lead to coherent results. Basic spectral calculations are based on similar algorithms with specific microscopic cross sections ENDF/BV for SCALE and JEF2.2 for CRISTAL. The main differences between the two packages is on one hand SCALE's three dimensional fuel assembly geometry is described by a pin by pin model while an homogenized fuel assembly description is used by CRISTAL and on the other hand SCALE is working with either 44 or 238 neutron energy groups while CRISTAL is with a 172 neutron energy groups. Those two computer packages rely on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask with two fuel assemblies is designed to maximize fuel isolation inside a cask and with neighboring ones even for large array configuration cases. Proven industrial products are used: - BoralTM as neutron absorber - High density polyethylene (HDPE) or Nylon as neutron moderator - Foam as thermal and mechanical protection The cask is designed to handle the complete

  3. AREVA NP next generation fresh UO2 fuel assembly shipping cask: SCALE - CRISTAL comparisons lead to safety criticality confidence

    International Nuclear Information System (INIS)

    AREVA NP as a worldwide PWR fuel provider has to have a fleet of fresh UO2 shipping casks being agreed within a lot of countries including USA, France, Germany, Belgium, Sweden, China, and South Africa - and to accommodate foreseen EPR Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector decided to develop an up-to-date shipping cask (so called MAP project) gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and local Safety Authorities. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: - Preferential flooding; - Fuel rod lattice pitch expansion for full length of fuel assemblies; - Neutron absorber penalty; -... Well known computer codes, American SCALE package and French CRISTAL package, were used to check configurations reactivity and to ensure that both codes lead to coherent results. Basic spectral calculations are based on similar algorithms with specific microscopic cross sections ENDF/BV for SCALE and JEF2.2 for CRISTAL. The main differences between the two packages is on one hand SCALE's three dimensional fuel assembly geometry is described by a pin by pin model while an homogenized fuel assembly description is used by CRISTAL and on the other hand SCALE is working with either 44 or 238 neutron energy groups while CRISTAL is with a 172 neutron energy groups. Those two computer packages rely on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask with two fuel assemblies is designed to maximize fuel isolation inside a cask and with neighboring ones even for large array configuration cases. Proven industrial products are used: - BoralTM as neutron absorber; - High density polyethylene (HDPE) or Nylon as neutron moderator; - Foam as thermal and mechanical protection. The cask is

  4. Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

    Directory of Open Access Journals (Sweden)

    Skrzypek Maciej

    2015-09-01

    Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

  5. CFD study on inlet flow blockage accidents in rectangular fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Fan, Wenyuan, E-mail: fanwy@mail.ustc.edu.cn; Peng, Changhong, E-mail: pengch@ustc.edu.cn; Guo, Yun, E-mail: guoyun79@ustc.edu.cn

    2015-10-15

    Highlights: • 3D CFD and Relap5 simulations on inlet flow blockage are performed. • Transient effects are investigated by dynamic mesh technique. • Similar flow and power redistributions are predicted in both methods. • Local effects of the blockage are captured by CFD method and analyzed. - Abstract: Three-dimensional transient CFD simulation of 90% inlet flow blockage accidents in rectangular fuel assembly is performed, using the dynamic mesh technique. One-dimensional steady calculation is done for comparison, using Relap5 code. Similar mass flow rate redistributions and asymmetric power redistributions of the plate in the blocked scenario are obtained. No boiling is predicted in both simulations, however, CFD approach provides more in-depth investigations of flow transients and the thermal-hydraulic interaction. The development of flow blockage transients is so fast that the rapid redistribution of mass flow rates occurs in only 0.015 s after the formation of the blockage. As a sequence of the inlet flow blockage, jet-flows and reversed flows occur in the blocked channel. This leads to complex temperature distributions of coolants and fuel plates, in which, the highest coolant temperature no longer occurs around the channel outlet. The present study shows the advantage and significance of the application of three-dimensional transient CFD technique in investigating flow blockage accidents.

  6. New LWR Fuel Assembly Concepts using Particle Burnable Poisons for Low Boron Concentration

    International Nuclear Information System (INIS)

    The most importance role of the soluble boron is the control of the long term reactivity to maintain the criticality of the reactor cores by reducing core excess reactivity. However, the use of soluble boron in the coolant leads to several issues. First, boron is corrosive and the presence of boron in the coolant will increase corrosion on the primary coolant loop and the corrosive nuclides will be mixed with the coolant. Furthermore, CVCS (Chemical and Volume Control System) is required to clean these corrosive elements from the coolant and to purify and control the level of boron diluted in the coolant. The presence of CVCS including the corrosive elements requires complicated maintenance and operation leading to increases of additional pipes which can add the possibilities of occurrences of LOCAs (Loss of Coolant Accident). Furthermore, the removal of soluble boron or reduction of soluble boron concentration makes the moderator temperature coefficient (MTC) more negative. In this paper, we suggest use of burnable poison rods where burnable poison particles are distributed in the SiC matrix as in the FCM (Fully Ceramic Micro-encapsulated) fuel and we performed a feasibility study on the use of the new LWR fuel assembly design concepts using this concept of new burnable poison rods to achieve low boron or boron-free cores

  7. Preparation of a self-humidifying membrane electrode assembly for fuel cell and its performance analysis

    Institute of Scientific and Technical Information of China (English)

    王诚; 毛宗强; 徐景明; 谢晓峰; 杨立寨

    2003-01-01

    A novel nano-porous material SiO2-gel was prepared. After being purified by H2O2, then protonized by H2SO4 and desiccated in vacuum, the SiO2-gel, mixed with Nafion solution, was coated between an electrode and a solid electrolyte, which made a new type of self-humidifying membrane electrode assembly. The SiO2 powder was characterized by FTIR, BET and XRD. The surface of the electrodes was characterized by SEM and EDS. The performances of the self-hu- midifying membrane electrodes were analyzed by polarization discharge and AC impedance under the operation modes of external humidification and self-humidification respectively. Experimental results indicated that the SiO2 powder held super-hydrophilicity, and the layer of SiO2 and Nafion polymer between electrode and solid electrolyte expanded three-dimension electrochemistry reaction area, maintained stability of catalyst layer and enhanced back-diffusion of water from cathode to anode, so the PEM Fuel cell can generate electricity at self-humidification mode. The power density of single PEM fuel cell reached 1.5 W/cm2 under 0.2 Mpa, 70℃ and dry hydrogen and oxygen.

  8. Uncertainty Analysis for OECD-NEA-UAM Benchmark Problem of TMI-1 PWR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyuk; Kim, S. J.; Seo, K.W.; Hwang, D. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A quantification of code uncertainty is one of main questions that is continuously asked by the regulatory body like KINS. Utility and code developers solve the issue case by case because the general answer about this question is still opened. Under the circumference, OECD-NEA has attracted the global consensus on the uncertainty quantification through the UAM benchmark program. OECD-NEA benchmark II-2 problem is a problem on the uncertainty quantification of subchannel code. It is a problem that the uncertainty of fuel temperature and ONB location on the TMI-1 fuel assembly are estimated on the transient and steady condition. In this study, the uncertainty quantification of MATRA code is performed on the problem. Workbench platform is developed to produce the large set of inputs that is needed to estimate the uncertainty quantification on the benchmark problem. Direct Monte Carlo sampling is used to the random sampling from sample PDF. Uncertainty analysis of MATRA code on OECD-NEA benchmark problem is estimated using the developed tool and MATRA code. Uncertainty analysis on OECD-NEA benchmark II-2 problem was performed to quantify the uncertainty of MATRA code. Direct Monte Carlo sampling is used to extract 2000 random parameters. Workbench program is developed to generate input files and post process of calculation results. Uncertainty affected by input parameters was estimated on the DNBR, the cladding and the coolant temperatures.

  9. Calculation model of molten material movement following LMFR fuel assemblies meltdown

    Energy Technology Data Exchange (ETDEWEB)

    Vlasichev, G.N. [State Technical Univ., Nizhny Novgorod (Russian Federation); Kiryushin, A.I.; Kuzavkov, N.G.

    1997-12-31

    Movement of molten heat-generating mass from the core to the diagrid is analyzed in an accident with blockage of Liquid Metal Fast Reactor (LMFR) individual fuel assembly (FA). The objective is to develop a calculational model to estimate the time and distance of molten material downward propagation. The calculational model is based on an assumption that a flat melting boundary exists under the heat-generating mass. Downward heat removal by heat conduction of underlying materials: fertile material, steel and sodium is considered. In the present version of the calculational procedure heat removal in radial direction from the heat-generating mass and underlying materials in the damaged FAs group to adjacent intact FAs taken into account. For mathematical representation a model of non-stationary effective heat conductivity is used. Approximate numerical solution of initial two-dimensional equation of heat conductivity with anisotropy in axial and radial directions is obtained. Numerical estimate of fuel downward propagation from the core in one BN-800 FA and 7 FAs is performed. (author)

  10. Efficiency of membrane electrolyte assembly of hydrogen fuel cells : thermodynamic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Nemec, T.; Marsik, F. [Inst. of Thermomechanics ASCR, Prague (Czech Republic). Dept. of Thermodynamics; Mican, O. [Czech Technical Univ., Prague (Czech Republic). Faculty of Technical and Nuclear Physics, Dept. of Mathematics

    2009-07-01

    The performance of hydrogen fuel cells is limited by water diffusivity and electric conductivity in polymer exchange membrane (PEM) and the redox affinity at the electrodes. This study analyzed gas diffusion electrodes, catalyst layers and the diffusion of protons and water through the PEM from an irreversible thermodynamics point of view. The minimum entropy production principle was used to calculate the unknown transport coefficients in the mass balance equations involving water and proton transport through the membrane. Redox processes in the fuel were described along with the catalytic activity of platinum. The redox affinity of electrochemical reactions taking place at the surfaces were coupled with the catalytic activity of platinum. The entire membrane electrode assembly (MEA) was analyzed and the affect of coupling on the total maximum efficiency was determined. The model was used to derive an explicit relation for the optimal coupling between water diffusion and electro-osmotic flux in a PEM, as well as the relations for the characteristic thickness of a PEM membrane. A 2D numerical simulation of water and electric potential distribution in a membrane for an unsteady loading was used to support the simplified analytic solution.

  11. Development of a membrane electrode assembly process for proton exchange membrane fuel cell (PEMFC)

    International Nuclear Information System (INIS)

    In this work, a Membrane Electrode Assembly (MEA) producing process was developed, involving simple steps, aiming cost reduction and good reproducibility for Proton Exchange Membrane Fuel Cell (PEMFC) commercial applications. The electrodes were produced by spraying ink into both sides of the polymeric membrane, building the catalytic layers, followed by hot pressing of Gas Diffusion Layers (GDL), forming the MEA. This new producing method was called 'Spray and hot pressing hybrid method'. Concerning that all the parameters of spray and hot pressing methods are interdependent, a statistical procedure were used in order to study the mutual variables influences and to optimize the method. This study was earned out in two distinct steps: the first one, where seven variables were considered for the analysis and the second one, where only the variables that interfered in the process performance in the first step were considered for analysis. The results showed that the developed process was adequate, including only simple steps, reaching MEA's performance of 651 m A cm-2 at a potential of 600 mV for catalysts loading of 0,4 mg cm-2 Pt at the anode and 0,6 mg cm-2 Pt at the cathode. This result is compared to available commercial MEA's, with the same fuel cell operations conditions. (author)

  12. Burnup Estimation for Plate Type Fuel Assembly Using SCALE6 Code

    Energy Technology Data Exchange (ETDEWEB)

    Alawneh, Luay M.; Park, Chang Je; Lee, Byungchul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    Accurate burnup estimation is not an easy job due to several reasons such as the effect of fission products and the power change caused by fuel refueling and depletion. The presence of fission products may distort the linear relationship between burnup and input parameters including power density and enrichment. The feasibility test of this approach has been done by comparing the results with a Monte Carlo code results. In this paper, it has been tried to get a crude formula to estimate burnup for an open pool type research reactor. In addition, we want to investigate the perturbation of each factor on burnup, and then combine the effects in one fitted formula for each cycle. This work is focused on calculating burnup for plate type fuel assembly of research reactors through a couple of code systems such as TRITON/NEWT and ORIGEN-ARP. Several sensitivity calculations have been done and the least square fitting is carried out to express a unified formula for burnup. The estimated burnup is compared with that of McCARD calculation. It is founded that the fitted burnup agrees well with the McCARD results.

  13. CFD study on inlet flow blockage accidents in rectangular fuel assembly

    International Nuclear Information System (INIS)

    Highlights: • 3D CFD and Relap5 simulations on inlet flow blockage are performed. • Transient effects are investigated by dynamic mesh technique. • Similar flow and power redistributions are predicted in both methods. • Local effects of the blockage are captured by CFD method and analyzed. - Abstract: Three-dimensional transient CFD simulation of 90% inlet flow blockage accidents in rectangular fuel assembly is performed, using the dynamic mesh technique. One-dimensional steady calculation is done for comparison, using Relap5 code. Similar mass flow rate redistributions and asymmetric power redistributions of the plate in the blocked scenario are obtained. No boiling is predicted in both simulations, however, CFD approach provides more in-depth investigations of flow transients and the thermal-hydraulic interaction. The development of flow blockage transients is so fast that the rapid redistribution of mass flow rates occurs in only 0.015 s after the formation of the blockage. As a sequence of the inlet flow blockage, jet-flows and reversed flows occur in the blocked channel. This leads to complex temperature distributions of coolants and fuel plates, in which, the highest coolant temperature no longer occurs around the channel outlet. The present study shows the advantage and significance of the application of three-dimensional transient CFD technique in investigating flow blockage accidents

  14. AHF Booster Tracking with SIMPSONS.

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, D. E. (David E.); Neri, F. (Filippo)

    2002-01-01

    The booster lattice for the Advanced Hydrotest Facility at Los Alamos was tracked in 3-D with the program SIMPSONS, using the full, symplectic lattice from TEAPOT, using the full set of magnet and misalignment errors, as well as full space-charge effects. The only corrections included were a rough closed-orbit correction and chromaticity correction. The lattice was tracked for an entire booster cycle, from multi-turn injection through acceleration to the top energy of 4 GeV, approximately 99,000 turns. An initial injection intensity of 4x1Ol2, injected in 25 turns, resulted in a final intensity of 3 . 2 {approx} 1 0a' {approx}t 4 GeV. Results of the tracking, including emittance growth, particle loss, and particle tune distributions are presented.

  15. Methodologies to determine the Pu content of spent fuel assemblies for input nuclear material accountancy of pyroporcessing

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Taehoon; Shin, Heesung; Kim, Youngsoo; Kim, Hodong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kwon, Taeje [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

    2011-07-01

    This study shows two different non-destructive approaches to determine the Pu mass of spent fuel assemblies, and the analysis results on the errors in their Pu mass. For both methods, the Cm mass of the assembly is obtained based on the neutron measurement results. The Cm ratio of the assembly is determined from the Cm mass and the Pu mass obtained by using either of the two methods. In a comparison of two methods, the second method is simpler than the first and may not need a homogeneously-mixed sample of the spent fuel assembly. On the other hand, the second approach shows larger error in the estimated Pu mass than the first one for many different spent fuel cases of various burnup, initial enrichment, and cooling times. A member state support program for the development of the IAEA safeguards approach for an engineering-scale pyroprocessing facility, which is designated as the Reference Engineering-scale Pyroprocessing Facility(REPF), has been carried out by Korea Atomic Energy Research Institute since 2008. The nuclear material accountancy of the REPF is based on the 'Cm balance' technique. The Pu content of processing materials of pyroprocessing can be determined by measuring the Cm mass of the materials and multiplying it by the Cm ratio. The spent fuel assembly is de-cladded, and the irradiated UO{sub 2} material of the assembly is homogeneously mixed in the homogenization process in order to obtain a representative sample of the spent fuel assembly for determining the mass of Pu, U and Cm elements, as well as the Cm ratio of the campaign. The shipper-receiver difference between the nuclear power plant and HPC of REPF is determined at this point. We found that the error for the Pu mass and Cm ratio determined from the homogenized uranium oxide powder is the most critical for the determination of the material unaccounted for throughout the whole processes. This paper presents two approaches to determine the Pu mass of spent fuel assemblies using non

  16. A study of the friction and wear processes of the structural components of fuel assemblies for water-cooled and water moderated power reactors

    International Nuclear Information System (INIS)

    The friction forces affect the fuel assembly (FA) strength at all the stages of its lifecycle. The paper covers the methods and the results of the pre-irradiation experimental studies of the static and dynamic processes the friction forces are involved in. These comprise the FA assembling at the manufacturer, fuel rod flow-induced vibration and fretting-wear in the fuel rod-to-cell friction pairs, rod cluster control assembly (RCCA) movement in the FA guide tubes, FA bowing, FA loading-unloading into the core, irradiation-induced growth and thermal-mechanical fuel rod-to-spacer grid interaction. (authors)

  17. Non-destructive γ-ray spectrometry and analysis on spent fuel assemblies of the JPDR-I

    International Nuclear Information System (INIS)

    Non-destructive gamma-ray spectrometry was carried out on the spent fuel assemblies of the whole core of JPDR-I which was a BWR. These data were analyzed by considering power distribution, spatial variation of neutron spectrum, and history of reactor operation. The burnup and the Pu/U atom ratio in each assembly were derived from the non-destructively measured distributions of 137Cs activity and 134Cs/137Cs activity ratio, respectively, by using calibration curves established for fuel specimens of a standard assembly of the core. The results were compared with calculational ones based on the operational data, and good correlations were found between them. The total amount of plutonium build-up in the core estimated from the non-destructive measurements agreed quite well with the amount obtained from reprocessing. (author)

  18. In-field Calibration of a Fast Neutron Collar for the Measurement of Fresh PWR Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); De Baere, Paul [European Commission (Luxembourg). Euratom Safeguards

    2015-04-17

    A new neutron collar has been designed for the measurement of fresh LEU fuel assemblies. This collar uses “fast mode” measurement to reduce the effect of burnable poison rods on the assay and thus reduce the dependence on the operator’s declaration. The new collar design reduces effect of poison rods considerably. Instead of 12 pins of 5.2% Gd causing a 20.4% effect, as in the standard thermal mode collar, they only cause a 3.2% effect in the new collar. However it has higher efficiency so that reasonably precise measurements can be made in 25 minutes, rather than the 1 hour of previous collars. The new collar is fully compatible with the use of the standard data collection and analysis code INCC. This report describes the calibration that was made with a mock-up assembly at Los Alamos National Laboratory and with actual assemblies at the AREVA Fuel fabrication Plant in Lingen, Germany.

  19. The passive nondestructive assay of the plutonium content of spent-fuel assemblies from the BN-350 fast-breeder reactor in the city of Aqtau, Kazakhstan

    CERN Document Server

    Lestone, J P; Rennie, J A; Sprinkle, J K; Staples, P; Grimm, K N; Hill, R N; Cherradi, I; Islam, N; Koulikov, J; Starovich, Z

    2002-01-01

    The International Atomic Energy Agency is presently interested in developing equipment and techniques to measure the plutonium content of breeder reactor spent-fuel assemblies located in storage ponds before they are relocated to more secure facilities. We present the first quantitative nondestructive assay of the plutonium content of fast-breeder reactor spent-fuel assemblies while still underwater in their facility storage pond. We have calibrated and installed an underwater neutron coincidence counter (Spent Fuel Coincidence Counter (SFCC)) in the BN-350 reactor spent-fuel pond in Aqtau, Kazakhstan. A procedure has been developed to convert singles and doubles (coincidence) neutron rates observed by the SFCC into the total plutonium content of a given BN-350 spent-fuel assembly. The plutonium content has been successfully determined for spent-fuel assemblies with a contact radiation level as high as approx 10 sup 5 Rads/h. Using limited facility information and multiple measurements along the length of spe...

  20. Experimental and calculational investigation of longitudinal-and-transverse coolant flow in the fuel assembly of a gas cooled reactor

    International Nuclear Information System (INIS)

    A method for calculating fields of rates temperature and pressure of gas coolant in two-dimensional setup in a gas-cooled reactor fuel assembly including an inlt monifold porous energy-release medium and outlet monifold is considered. From the view of structure the fuel assembly coaxially located porous cylinders with spherical fuel elements between them. An experimental large-scale facility permitting to perform a set of necessary measurements is described. Fields of rtes in the inlet manifold were measured as well as pressure distributions on external (imprenetrable) and internal (perforated) walls of the inlet manifold on its end face wall, on the perforated wall of the outlet manifoled. Experimental and calculational results have been compared

  1. Co-flow anode/cathode supply heat exchanger for a solid-oxide fuel cell assembly

    Science.gov (United States)

    Haltiner, Jr., Karl J.; Kelly, Sean M.

    2005-11-22

    In a solid-oxide fuel cell assembly, a co-flow heat exchanger is provided in the flow paths of the reformate gas and the cathode air ahead of the fuel cell stack, the reformate gas being on one side of the exchanger and the cathode air being on the other. The reformate gas is at a substantially higher temperature than is desired in the stack, and the cathode gas is substantially cooler than desired. In the co-flow heat exchanger, the temperatures of the reformate and cathode streams converge to nearly the same temperature at the outlet of the exchanger. Preferably, the heat exchanger is formed within an integrated component manifold (ICM) for a solid-oxide fuel cell assembly.

  2. Development of a Mobile CZT Detector System for Burnup Measurement of Spent Fuel Assembly and On-Site Application

    International Nuclear Information System (INIS)

    The advantages of mobile CdZnTe (CZT) detector for nuclear safeguard applications of spent fuel burnup inspection in assembly storage pond are compactness, low cost and ease of operations. In this work, a mobile detection system shield with tungsten alloy was designed and then performed on-site. Net count rate of the 662 keV line of 137Cs was produced linearly with burnup as experimental data simulations shows, in which the deviation from linearity is smaller than 9%. As a result, the feasibility of the method using CZT detector to monitor spent nuclear fuel assembly burnup in a fuel pond was validated. The results calculated with Monte Carlo procedure Geant4 can provide a theoretical guide for the further burnup measurement. (author)

  3. Assessment of the impacts of spent fuel disassembly alternatives on the Nuclear Waste Isolation System. [Preparing and packaging spent fuel assemblies for geologic disposal

    Energy Technology Data Exchange (ETDEWEB)

    1984-07-01

    The objective of this report was to evaluate four possible alternative methods of preparing and packaging spent fuel assemblies for geologic disposal against the Reference Process of unmodified spent fuel. The four alternative processes were: (1) End fitting removal, (2) Fission gas venting and resealing, (3) Fuel bundle disassembly and close packing of fuel pins, and (4) Fuel shearing and immobilization. Systems analysis was used to develop a basis of comparison of the alternatives. Conceptual processes and facility layouts were devised for each of the alternatives, based on technology deemed feasible for the purpose. Assessments were made of 15 principal attributes from the technical, operational, safety/risk, and economic considerations related to each of the alternatives, including both the surface packaging and underground repository operations. Specific attributes of the alternative processes were evaluated by assigning a number for each that expressed its merit relative to the corresponding attribute of the Reference Process. Each alternative process was then ranked by summing the numbers for attributes in each of the four assessment areas and collectively. Fuel bundle disassembly and close packing of fuel pins was ranked the preferred method of disposal of spent fuel. 63 references, 46 figures, 46 tables.

  4. Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513. Interim report

    International Nuclear Information System (INIS)

    The effects of the thermally-induced cracking and subsequent relocation of UO2 fuel pellets on the thermal and mechanical behavior of light-water reactor fuel rods during irradiation are quantified in this report. Data from the Nuclear Regulatory Commission/Pacific Northwest Laboratory Halden experiments on instrumented fuel assemblies (IFA) IFA-431, IFA-432, and IFA-513 are analyzed. Beginning-of-life in-reactor measurements of fuel center temperatures, linear heat ratings, and cladding axial elongations are used in a new model to solve for the effective thermal conductivity and elastic moduli of the cracked fuel column. The primary assumptions of the new model are that (1) the cracked fuel is in a hydrostatic state of stress in the (r,theta) plane, and that (2) there is no axial slipping between fuel and cladding. Three basic parameters are used to describe the cracked fuel: (1) the crack pattern, (2) the crack roughness, and (3) the fuel surface (gap) roughness. Recommendations are made on refining the model

  5. Doses of the staff during the spent fuel assemblies transportation and storage in Nuhmos 56V concrete system

    International Nuclear Information System (INIS)

    The NUHMOS 56V concrete system provides long-term interim storage (50 years) for spent fuel assemblies, which have been out of the reactor for a sufficient period of time. It consists from horizontal storage modules. The fuel assemblies are confined in a helium atmosphere by a canister containment pressure vessel. The canister is protected and shielded by a massive reinforced concrete module. Decay heat is removed from the canister and concrete module by a passive natural draft convection ventilation system. The project of storage does not foresee the radiation monitoring inside of building and around it. But we provided and realize the radiation monitoring program around storage, it includes tree phases: - determination the zero background around the building before storage put in exploiting; - monitoring of the radioactive particles in air (additional aspiration plant); dose rate monitoring by portable dosimeters and soil monitoring during the process of the fuel storage; - constantly after the completion the fuel storage process - monitoring of the radioactive particles in air (additional aspiration plant); dose rate monitoring by portable dosimeters, and soil monitoring. Also designed the dose rate monitoring by the dosimeter RME3 with the transfer of data by radio channel to central monitor. The canistered spent fuel assemblies are transferred from the plant's spent fuel pool to the concrete storage modules in a transfer cask. The cask is aligned with the storage module and the canister and inserted into the module by means of a hydraulic ram. The system is a totally passive installation that is designed to provide shielding and safe confinement of spent fuel for a range of postulated accident conditions and natural phenomena. (authors)

  6. The need to address the larger universe of HEU-fueled reactors, including critical assemblies, pulsed reactors and propulsion reactors

    International Nuclear Information System (INIS)

    Full text: The RERTR program has focused thus far primarily on ending shipments of HEU fuel to research reactors. This has resulted in giving highest priority to reactors with steady thermal powers of 1 megawatt or more, because they require regular refuelling. Critical facilities and pulsed reactors can also of serious concern, because some of them contain very large amounts of barely-irradiated HEU and plutonium. They could be costly to convert - and conversion to LEU may be impractical for fast-neutron critical assemblies. An assessment should be carried out first, therefore, as to which are still needed. Critical assemblies are required today primarily to benchmark Monte Carlo neutron-transport codes. Perhaps the world nuclear community could share a few instead of each reactor-design institute having its own. There is also a whole universe of HEU-fuelled pressurized-water reactors used to power submarines and other types of nuclear-powered ships. These reactors collectively require much more HEU fuel each year than research reactors. The risk of HEU diversion from their fuel cycles is not zero but it is difficult for outsiders to discuss conversion because of the fuel designs are classified. This makes the conversion of Russia's civilian icebreaker reactors of particular interest because issues of classified fuel design are less problematic and these reactors load annually fuel containing about 400 kg of U-235. Another reason for interest in developing LEU fuel for these reactors is that the KLT-40 icebreaker reactor is being adapted for a floating nuclear power plant. Finally, the research-reactor community is, in any case, faced with developing fuels that can operate at power-reactor-fuel temperatures because there are a few high-powered research reactors that operate in this temperature range. (author)

  7. VVER-1000/V320 decay heat analysis involving TVS-M and TVSA fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Petkov, Plamen V. [' Kozloduy' NPP, 3321 Kozloduy, Vratsa (Bulgaria)], E-mail: pvpetkov@yahoo.com; Hristov, Danail V. [' Kozloduy' NPP, 3321 Kozloduy, Vratsa (Bulgaria)], E-mail: dvhristov@npp.bg

    2008-12-15

    MELCOR 1.8.4 is an integral computer code, developed for severe accident calculations. It is used primarily for the simulation of PWR and BWR types of reactors as there exists an internal database, suitable for modeling of their core inventory. Despite similarity between VVER-1000/V320 and PWR, accounting of specificities of Russian reactor designs is still required. Part of it is the simulation of core decay heat rate after the shutdown. MELCOR 1.8.4 distinguishes fifteen classes. Each of them contains chemical elements with similar properties. Twelve are involved in radioactive products decay. In current paper the authors present two boundary reactor core loadings, designed with corresponding fuel assemblies: TVS-M and TVSA. They have calculated decay heat after reactor shutdown from 100% and 104% of nominal power by SCALE 4.4a package. The amount of generated nuclides had also been estimated. Irradiation history had been accounted as proposed in Kolobashkin et al. (p. 141) [Kolobashkin, V.M., Rubtsov, P.M., Rujanskiy, P.A., Sidorenko, V.D., 1983. Radionuclide Inventory Estimation Handbook (on Russian). Energoatomizdat, Moscow, pp. 138-188]. Newly developed Core Inventory Estimation Tool (CIET), described in this paper, written and tested previously, has been used for the evaluation of core decay heat fractions, distributed over chemical classes. Twelve curves were generated by following the same numerical procedure implemented in MELCOR for representation of decay in W/kg. Comparison of chemical element decay rates to the defaults for PWR shows deviations from the expectations to maximal values of 37% in Uranium for TVSA fuel assemblies. The total number of radionuclides, separated in chemical classes, given in Gauntt et al. [Gauntt, R.O., Cole, R.K., Rodrigez, S.B., Sanders, R.L., Smith, R.C., Stuard, D.S., Summers, R.M., Young, M.F., 1997. MELCOR Computer Code Manuals. NUREG/CR-6119 Report, Vol. 1 and Vol. 2, SAND97-2398] was compared to the ones involved in

  8. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

  9. Assessment of the integrity of spent fuel assemblies used in dry storage demonstrations at the Nevada Test Site

    International Nuclear Information System (INIS)

    This report summarizes the histories of 17 Zircaloy-clad spent fuel assemblies used in dry storage tests and demonstrations at the Engine Maintenance and Disassembly (EMAD) and Climax facilities at the Nevada Test Site (NTS). The 18th assembly was shipped to the Battelle Columbus Laboratory (BCL) and remained there for extensive characterization and as a source of specimens for whole-rod and rod-segment dry storage tests. The report traces the history of the assemblies after discharge from the Turkey Point Unit 3 pressurized-water reactor (1975 and 1977) through shipment (first arrival at EMAD in December 1978), dry storage tests and demonstrations, and shipment by truck cask from EMAD to the Idaho National Engineering Laboratory (INEL) in May/June 1986. The principal objectives of this report are to assess and document the integrity of the fuel during the extensive dry storage activities at NTS and BCL, and to briefly summarize the dry storage technologies and procedures demonstrated in this program. The dry storage tests and demonstrations involved the following concepts and facilities: (1) surface drywells (EMAD); (2) deep drywells (425 m underground in the Climax granite formation); (3) concrete silo (EMAD); (4) air-cooled vault (EMAD); (5) electrically-heated module for fuel assembly thermal calibration and testing (EMAD/FAITM). 20 refs., 43 figs., 9 tabs

  10. Membrane-less cloth cathode assembly (CCA) for scalable microbial fuel cells.

    Science.gov (United States)

    Zhuang, Li; Zhou, Shungui; Wang, Yueqiang; Liu, Chengshuai; Geng, Shu

    2009-08-15

    One of the main challenges for scaling up microbial fuel cell (MFC) technologies is developing low-cost cathode architectures that can generate high power output. This study developed a simple method to convert non-conductive material (canvas cloth) into an electrically conductive and catalytically active cloth cathode assembly (CCA) in one step. The membrane-less CCA was simply constructed by coating the cloth with conductive paint (nickel-based or graphite-based) and non-precious metal catalyst (MnO(2)). Under the fed-batch mode, the tubular air-chamber MFCs equipped with Ni-CCA and graphite-CCA generated the maximum power densities of 86.03 and 24.67 mW m(-2) (normalized to the projected cathode surface area), or 9.87 and 2.83 W m(-3) (normalized to the reactor liquid volume), respectively. The higher power output of Ni-CCA-MFC was associated with the lower volume resistivity of Ni-CCA (1.35 x 10(-2)Omega cm) than that of graphite-CCA (225 x 10(-2)Omega cm). At an external resistance of 100 Omega, Ni-CCA-MFC and graphite-CCA-MFC removed approximately 95% COD in brewery wastewater within 13 and 18d, and achieved coulombic efficiencies of 30.2% and 19.5%, respectively. The accumulated net water loss through the cloth by electro-osmotic drag exhibited a linear correlation (R(2)=0.999) with produced coulombs. With a comparable power production, such CCAs only cost less than 5% of the previously reported membrane cathode assembly. The new cathode configuration here is a mechanically durable, economical system for MFC scalability. PMID:19556120

  11. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2013-03-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.”(Reference 1) The experiment studied in this evaluation was the first of the series and had the fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, , and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configurations are described in Sections 1.3, 1.4 and 1.7, respectively. Information for this

  12. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2012-09-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.”(Reference 1) The experiment studied in this evaluation was the first of the series and had the fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, , and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configurations are described in Sections 1.3, 1.4 and 1.7, respectively. Information for this

  13. Control rod effects on reaction rate distributions in tight pitched PuO2-UO2 fuel assembly

    International Nuclear Information System (INIS)

    Investigations were made for the heterogeneity effects caused by insertion or withdrawal of a B4C control rod on fine structure of reaction rates distributions in a tight pitched PuO2-UO2 fuel assembly. Analysis was carried out by using the VIM and SRAC codes with the libraries based on JENDL-2 for the hexagonal fuel assembly basically corresponding to the PROTEUS-LWHCR experimental core. The reaction rates are affected more remarkably by the withdrawal of the control rod rather than its insertion. The changes of the reaction rates were decomposed into three terms of spectrum shifts, the changes of effective cross sections with fine groups, and their higher order components. From the analysis, it is concluded that most changes of reaction rates are caused by spectral shifts. The SRAC code with fine group constants can predict the distribution of reaction rates and their ratios with the accuracy of about 5 % except for the values related to Pu-242 capture rate, as compared with the VIM results. To increase the accuracy, it is necessary to generate the effective cross sections of the fuel near control rods with consideration of the heterogeneities in the fuel assembly. (author)

  14. Study of boron dilution phenomenon in the core and fuel assemblies of Bushehr VVERr-1000 reactor in normal operating conditions

    International Nuclear Information System (INIS)

    The spatial temperature distributions in fuel and coolant, results in appearing local changes in those elements densities in the reactor core, and also due to the complete solubility of boric acid in the coolant, there will be a direct correlation between the changes in the boron concentration and the coolant density. Because of the gradual reduction of boron concentration, first a local positive reactivity will be inserted into the core which will cause slight thermo-neutronic fluctuations in the reactor core. Of course, the trend of this process in the case of excessive reduction of the density of the coolant and evaporation of water (accident scenarios) will be reversed and subsequently the negative reactivity will be given to the system. With regard to the importance of this phenomenon, the spatial changes of boron concentration in the core and fuel assemblies of Bushehr VVER-1000 reactor have been examined. In line with this, by designing a complete thermo-neutronic cycle and by using CITATION, WIMS D-5 and COBRAN-EN codes, coolant temperature distribution and boron concentration will be calculated through this procedure, which first by using the output results of WIMS and CITATION codes, the thermal power of each fuel assembly will be calculated and finally, by linking these data to COBRA-EN code and using core and sub-channel analysis methods, the three-dimensional (3D) calculations of boron dilution will be obtained in the core as well as the fuel assemblies of the reactor. (authors)

  15. Determination of optimal imaging parameters for the reconstruction of a nuclear fuel assembly using limited angle neutron tomography

    International Nuclear Information System (INIS)

    The core components of nuclear reactors (e.g., fuel assemblies, spacer grids, control rods) encounter harsh environments due to high temperature, physical stress, and a tremendous level of radiation. The integrity of these elements is crucial for safe operation of nuclear power plants; post-irradiation examination (PIE) can reveal information about the integrity of these components. Neutron computed tomography (CT) is one important PIE measurement tool for nondestructively evaluating the structural integrity of these items. CT typically requires many projections to be acquired from different view angles, after which a mathematical algorithm is used for image reconstruction. However, when working with heavily irradiated materials and irradiated nuclear fuel, obtaining many projections is laborious and expensive. Image reconstruction from a smaller number of projections has been explored to achieve faster and more cost-efficient PIE. Classical reconstruction methods (e.g., filtered backprojection), unfortunately, do not typically offer stable reconstructions from a highly asymmetric, few-projection data set and often create severe streaking artifacts. We propose an iterative reconstruction technique to reconstruct curved, plate-type nuclear fuel assemblies using limited-angle CT. The performance of the proposed method is assessed using simulated data and validated through real projections. We also discuss the systematic strategy for establishing the conditions of reconstructions and finding the optimal imaging parameters for reconstructions of the fuel assemblies from few projections using limited-angle CT. Results show that a fuel assembly can be reconstructed using limited-angle CT if 36 or more projections are taken from a particular direction with 1° angular increment

  16. Determination of optimal imaging parameters for the reconstruction of a nuclear fuel assembly using limited angle neutron tomography

    Science.gov (United States)

    Abir, M. I.; Islam, F. F.; Craft, A.; Williams, W. J.; Wachs, D. M.; Chichester, D. L.; Meyer, M. K.; Lee, H. K.

    2016-01-01

    The core components of nuclear reactors (e.g., fuel assemblies, spacer grids, control rods) encounter harsh environments due to high temperature, physical stress, and a tremendous level of radiation. The integrity of these elements is crucial for safe operation of nuclear power plants; post-irradiation examination (PIE) can reveal information about the integrity of these components. Neutron computed tomography (CT) is one important PIE measurement tool for nondestructively evaluating the structural integrity of these items. CT typically requires many projections to be acquired from different view angles, after which a mathematical algorithm is used for image reconstruction. However, when working with heavily irradiated materials and irradiated nuclear fuel, obtaining many projections is laborious and expensive. Image reconstruction from a smaller number of projections has been explored to achieve faster and more cost-efficient PIE. Classical reconstruction methods (e.g., filtered backprojection), unfortunately, do not typically offer stable reconstructions from a highly asymmetric, few-projection data set and often create severe streaking artifacts. We propose an iterative reconstruction technique to reconstruct curved, plate-type nuclear fuel assemblies using limited-angle CT. The performance of the proposed method is assessed using simulated data and validated through real projections. We also discuss the systematic strategy for establishing the conditions of reconstructions and finding the optimal imaging parameters for reconstructions of the fuel assemblies from few projections using limited-angle CT. Results show that a fuel assembly can be reconstructed using limited-angle CT if 36 or more projections are taken from a particular direction with 1° angular increment.

  17. Booster double harmonic setup notes

    Energy Technology Data Exchange (ETDEWEB)

    Gardner, C. J. [Brookhaven National Lab. (BNL), Upton, NY (United States). Collider-Accelerator Dept.

    2015-02-17

    The motivation behind implementing a booster double harmonic include the reduced transverse space charge force from a reduced peak beam current and reduced momentum spread of the beam, both of which can be achieved from flattening the RF bucket. RF capture and acceleration of polarized protons (PP) is first set up in the single harmonic mode with RF harmonic h=1. Once capture and acceleration have been set up in the single harmonic mode, the second harmonic system is brought on and programmed to operate in concert with the single harmonic system.

  18. Increment of capacity of casks for LWR spent fuel transport (3). Demonstration of quantification of neutron leakage from a fuel assembly

    International Nuclear Information System (INIS)

    H(n,γ) method is proposed to quantify the number of leakage neutrons from a fuel assembly immersed in light water pond. The quantification is achieved by counting 2.223 MeV γ rays radiated by neutron capture reactions in hydrogen (H(n,γ)) outside the assembly. In the cases that sufficient thickness of light water surrounds the assembly, the number of neutron leakage almost equals to that of the γ ray emission. To relate the counts rate to the number of the γ ray emission, an evaluation method of detection efficiency is also developed assuming spatial distributions of 6Li(n,t) and 115In(n,γ) reactions are similar to that of the H(n,γ) one. The quantification is demonstrated for light water moderated sub-critical cores mocked up in Kyoto University Critical Assembly facility (KUCA). The 2.223 MeV γ rays are measured with a NaI scintillator. The distributions of the 6Li(n,t) reaction rate are obtained with optical-fiber detectors and those of the 115In(n,γ) reactions are done by the activation method. The experimentally deduced number of neutron absorptions outside the cores agrees with those estimated with a neutronics calculation code within accuracy of 4.5%. With the number of leakage neutrons from a spent fuel assembly quantified by the method, we can estimate the total number of fission neutron emission so that we can certify the burn-up of the assembly. (author)

  19. Numerical Simulation of Water Flow through the Bottom End Piece of a Nuclear Fuel Assembly

    Science.gov (United States)

    Navarro, Moysés A.; Santos, André A. C. Dos

    An experimental and numerical study was conducted on the pressure loss of flows through the bottom end piece of a nuclear fuel assembly. To determine an optimized numerical methodology using the commercial CFD code, CFX 10.0, a series of preliminary simulations of water flows through perforated plates in a square ducts were performed. A perforated plate is a predominant geometry of the bottom end piece, responsible for the majority of the flow's pressure drop. The numerical pressure loss applying an optimized mesh and the k-ɛ turbulence model showed good agreement when compared with a conventional methodology (Idelchik). Numerical results for the standard bottom end piece were obtained applying the previously determined mesh criteria and the k-ɛ turbulence model with some geometric simplifications. The agreement between the numerical simulations and experimental results can be considered satisfactory but suggests further numerical investigations with the bottom piece under real conditions of the experiment, without the geometric simplifications and with a gap between the piece and the wall of the flow channel. Additionally, other turbulence models should be appraised for this complex geometry.

  20. Performance comparison of microbial fuel cells equipped with different membrane electrode assemblies

    International Nuclear Information System (INIS)

    It is important for practical use of microbial fuel cells (MFCs) to not only develop new materials including electrodes and proton exchange membranes but also to understand the bacterial community structure related to electricity generation. Here, four kinds of novel membrane electrode assemblies (MEAs) were made. Four lactate fed MFCs equipped with the membranes were characterized by electrochemical, molecular-dependent and molecular-independent methods. MFC1 equipped with Nafion 117-type MEA (18 μm thickness) exhibited the highest performance. Although the other MEAs with different configurations of three kinds of polymers; poly (diallyldimethylammonium chloride), polyallylamine hydrochloride and poly (2-acrylamino-2-methyl -1-propanesulfonic acid) had thicknesses of about 0.3 μm (MEA 2 and 3) and 1.0 μm (MEA4), their power densities were lower. Denaturing gradient gel electrophoresis (DGGE) and phylogenetic analyses showed that anaerobic bacteria dominated in anode biofilms of MFC1. A bacterium completely corresponding to nucleotide sequence of one of the DGGE bands was isolated from the anode biofilm in MFC1. Interestingly, BLAST search indicated that the bacterium (named strain RO1) belonged to the genus of gram positive bacterium, Propioniferax. It was confirmed that strain RO1 was capable of producing electricity and constructing biofilm on the anode surface in pure culture MFC. These results suggested that the property of MEA affects significantly the bacterial community structure, thereby influencing the MFC-performance.

  1. Treating refinery wastewaters in microbial fuel cells using separator electrode assembly or spaced electrode configurations

    KAUST Repository

    Zhang, Fang

    2014-01-01

    The effectiveness of refinery wastewater (RW) treatment using air-cathode, microbial fuel cells (MFCs) was examined relative to previous tests based on completely anaerobic microbial electrolysis cells (MECs). MFCs were configured with separator electrode assembly (SEA) or spaced electrode (SPA) configurations to measure power production and relative impacts of oxygen crossover on organics removal. The SEA configuration produced a higher maximum power density (280±6mW/m2; 16.3±0.4W/m3) than the SPA arrangement (255±2mW/m2) due to lower internal resistance. Power production in both configurations was lower than that obtained with the domestic wastewater (positive control) due to less favorable (more positive) anode potentials, indicating poorer biodegradability of the RW. MFCs with RW achieved up to 84% total COD removal, 73% soluble COD removal and 92% HBOD removal. These removals were higher than those previously obtained in mini-MEC tests, as oxygen crossover from the cathode enhanced degradation in MFCs compared to MECs. © 2013 Elsevier Ltd.

  2. Improvements by employee motivation in the manufacture of nuclear fuel assemblies for LWRs

    International Nuclear Information System (INIS)

    Nuclear fuel assemblies are manufactured on a very high technical level and automation. However there is still a need for more improvement. One of the most important ways is employees motivation, because improvements lives of the ideas, impulses, initiatives and commitments of its employees. It can be realized by the employee himself or a group. Three ways of improvement by employees are mainly implemented at ANF: (i) ANF's 3i - program, based on the standard implementation within Siemens, is the first and an important strategy to improve processes, products and costs. It is to involve all employees and make use of the full potential for improvement The individual employee or a group make a suggestion and receive a commendation depending on the benefits. (ii) Work groups with a high level of responsibility are the second part. The groups mainly organize their work, working time and improvements by themselves. They help each other in job training, are very flexible and able to do also most of the maintenance work. (iii) CIP - groups (Continuous Improvement Process), based on the philosophy of KAMEN is the third strategy. These groups come together to improve all processes in the manufacturing area, also the administration or logistical processes at ANF. CIP - groups are implemented as so called long-term groups, the members are from different levels and departments. By comparing the different ways in order to achieve manufacturing improvements, employees motivation is one of the most important and cheapest part and will increase in significance in future. (author)

  3. Improving startup performance with carbon mesh anodes in separator electrode assembly microbial fuel cells

    KAUST Repository

    Zhang, Fang

    2013-04-01

    In a separator electrode assembly microbial fuel cell, oxygen crossover from the cathode inhibits current generation by exoelectrogenic bacteria, resulting in poor reactor startup and performance. To determine the best approach for improving startup performance, the effect of acclimation to a low set potential (-0.2V, versus standard hydrogen electrode) was compared to startup at a higher potential (+0.2V) or no set potential, and inoculation with wastewater or pre-acclimated cultures. Anodes acclimated to -0.2V produced the highest power of 1330±60mWm-2 for these different anode conditions, but unacclimated wastewater inocula produced inconsistent results despite the use of this set potential. By inoculating reactors with transferred cell suspensions, however, startup time was reduced and high power was consistently produced. These results show that pre-acclimation at -0.2V consistently improves power production compared to use of a more positive potential or the lack of a set potential. © 2013 Elsevier Ltd.

  4. Basic model for membrane electrode assembly design for direct methanol fuel cells

    Science.gov (United States)

    Krewer, Ulrike; Yoon, Hae-Kwon; Kim, Hee-Tak

    This research proposes a model that predicts the effect of the anode diffusion layer and membrane properties on the electrochemical performance and methanol crossover of a direct methanol fuel cell (DMFC) membrane electrode assembly (MEA). It is an easily extensible, lumped DMFC model. Parameters used in this design model are experimentally obtainable, and some of the parameters are indicative of material characteristics. The quantification of these material parameters builds up a material database. Model parameters for various membranes and diffusion layers are determined by using various techniques such as polarization, mass balance, electrochemical impedance spectroscopy (EIS), and interpretation of the response of the cell to step changes in current. Since the investigation techniques cover different response times of the DMFC, processes in the cell such as transport, reaction and charge processes can be investigated separately. Properties of single layers of the MEA are systematically varied, and subsequent analysis enables identification of the influence of the layer's properties on the electrochemical performance and methanol crossover. Finally, a case study indicates that the use of a membrane with lower methanol diffusivity and a thicker anode micro-porous layer (MPL) yields MEAs with lower methanol crossover but similar power density.

  5. Radionuclide characterization of reactor decommissioning waste and spent fuel assembly hardware

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission (NRC) has recently enacted rules setting forth technical, safety, and financial criteria for decommissioning of licensed nuclear facilities, including commercial nuclear power stations. These rules have addressed six major issues, including decommissioning alternatives, timing, planning, financial assurance, residual radioactivity, and environmental review. Also, the rules governing disposal of low-level radioactive wastes in commercial shallow land burial facilities will be applicable to most of the wastes generated during reactor decommissioning. This study has been implemented to provide the NRC and licensees with a more comprehensive and defensible data base and regulatory assessment of the radiological factors associated with reactor decommissioning and disposal of wastes generated during these activities. The objectives of this study are being accomplished during a two-phase sampling, measurement, and appraisal program utilizing: (1) the decommissioning of Shippingport Atomic Power Station, and (2) neutron activated materials from commercial reactors. Radioactive materials obtained from Shippingport Station and from a number of commercial stations for comprehensive radionuclide and stable element analyses are being utilized to assess the following important aspects of reactor decommissioning and radioactive waste characterization: (1) radiological safety and technology assessment from an actual reactor decommissioning (Shippingport); (2) radiological characterization of intensely radioactive materials (greater than Class-C) associated with the reactor pressure vessel and spent fuel assembly hardware from commercial nuclear power plants; (3) evaluation of the accuracy of computer codes for predicting radionuclide inventories in retired reactors and neutron activated components; and (4) assessment of waste disposal options associated with reactor decommissioning

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

    International Nuclear Information System (INIS)

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

  7. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    International Nuclear Information System (INIS)

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000 degree F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion (''bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled

  8. Monte Carlo simulations of differential die-away instrument for determination of fissile content in spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Tae-Hoon, E-mail: typhoon@kaeri.re.kr [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Korea Atomic Energy Research Institute, 150-1 Dukjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Menlove, Howard O.; Swinhoe, Martyn T.; Tobin, Stephen J. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2011-10-01

    The differential die-away (DDA) technique has been simulated by using the MCNPX code to quantify its capability of measuring the fissile content in spent fuel assemblies. For 64 different spent fuel cases of various initial enrichment, burnup and cooling time, the count rate and signal to background ratios of the DDA system were obtained, where neutron backgrounds are mainly coming from the {sup 244}Cm of the spent fuel. To quantify the total fissile mass of spent fuel, a concept of the effective {sup 239}Pu mass was introduced by weighing the relative contribution to the signal of {sup 235}U and {sup 241}Pu compared to {sup 239}Pu and the calibration curves of DDA count rate vs. {sup 239}Pu{sub eff} were obtained by using the MCNPX code. With a deuterium-tritium (DT) neutron generator of 10{sup 9} n/s strength, signal to background ratios of sufficient magnitude are acquired for a DDA system with the spent fuel assembly in water.

  9. 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

  10. Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control

    Science.gov (United States)

    Zafred, Paolo R.; Gillett, James E.

    2012-04-24

    A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

  11. 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.

  12. Optimization of wire-wrapped fuel assembly of LMR based on three-dimensional analysis of heat transfer

    International Nuclear Information System (INIS)

    In this work, performance evaluation of surrogate models in a wire-wrapped fuel assembly shape optimization is carried out. In addition to three basic models, i.e., Response Surface Approximation (RSA), Kriging (KRG) and Radial Basis Neural Network (RBNN), the multiple surrogate model Press Based Averaging (PBA) is also tested. Two design variables are selected to enhance the performance of wire wrapped fuel assembly and design points are selected using Latin Hypercube Sampling (LHS). Optimization problem has been stated as maximization of the objective function, defined as a linear combination of heat transfer and friction loss related terms with a weighing factor. Among the three basic models Kriging performs better while among the all models multiple surrogate model, PBA performs the best. Use of multiple surrogate PBA gives more robust approximation than individual surrogate. (author)

  13. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Čerba, Štefan, E-mail: stefan.cerba@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Vrban, Branislav; Lüley, Jakub [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Dařílek, Petr [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Zajac, Radoslav, E-mail: radoslav.zajac@vuje.sk [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Nečas, Vladimír; Haščik, Ján [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia)

    2014-02-15

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice.

  14. Optimization of enrichment distributions in nuclear fuel assemblies loaded with uranium and plutonium via a modified linear programming technique

    Science.gov (United States)

    Cuevas Vivas, Gabriel Francisco

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed OXide (MOX) fuels. MOX isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide-range of applicability of the optimization technique. The features of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or

  15. Assessment of RANS Based CFD Methodology using JAEA Experiment with a Wire-wrapped 127-pin Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, J. H.; Yoo, J.; Lee, K. L.; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, we assess the RANS based CFD methodology with JAEA experimental data. The JAEA experiment study with the 127-pin wire-wrapped fuel assembly was implemented using water for validating pressure drop formulas in ASFRE code. Complicated and vortical flow phenomena in the wire-wrapped fuel bundles were captured by vortex structure identification technique based on the critical point theory. The SFR system is one of the nuclear reactors in which a recycling of transuranics (TRUs) by reusing spent nuclear fuel sustains the fission chain reaction. This situation strongly motivated the Korea Atomic Energy Research Institute (KAERI) to start a prototype Gen-4 Sodium-cooled Fast Reactor (PGSFR) design project under the national nuclear R and D program. Generally, the SFR system has a tight package of the fuel bundle and a high power density. The sodium material has a high thermal conductivity and boiling temperature than the water. That can make core design to be more compact than Light Water Reactor (LWR) through narrower sub-channels. The fuel assembly of the SFR system consists of long and thin wire-wrapped fuel bundles and a hexagonal duct, in which wire-wrapped fuel bundles in the hexagonal tube has triangular loose array. The main purpose of a wire spacer is to avoid collisions between adjacent rods. Furthermore, a wire spacer can mitigate a vortex induced vibration, and enhance convective heat transfer due to the secondary flow by helical type wire spacers. Most of numerical studies in the nuclear fields was widely conducted based on the simplified sub-channel analysis codes such as COBRA (Rowe), SABRE (Macdougall and Lillington), ASFRE (Ninokata), and MATRA-LMR (Kim et al.). The relationship between complex flow phenomena and helically wrapped-wire spacers will be discussed. The RANS based CFD methodology is evaluated with JAEA experimental data of the 127-pin wirewrapped fuel assembly. Complicated and vortical flow phenomena in the wire-wrapped fuel

  16. Transportation of 33 irradiated MTR fuel assemblies from FRM/Garching to Savannah River Site, USA, using a GNS transport cask and using a new loading device

    Energy Technology Data Exchange (ETDEWEB)

    Dreesen, K.; Goetze, H.G.; Holst, L. [GNS, D-45127 Essen (Germany); Gerstenberg, H.; Schreckenbach, K. [Technical University of Munich, D-85748 Garching (Germany)

    2000-07-01

    According to the Department of Energy program of the return spent fuel from the foreign research reactors operators, 33 irradiated MTR box shaped fuel assemblies from the Technical University Munich were shipped to SRS/USA. The fuel assemblies were irradiated for typically 800 full days and, after a sufficient cooling time, loaded into a GNS 16 cask. The GNS 16 cask is a new transport cask for box shaped MTR fuel assemblies and TRIGA fuel assemblies and was used for the first time at the FRM Garching. The capacity of the cask is 33 box shaped MTR fuel assemblies. During the loading of the fuel assemblies, a newly developed loading device was used. The main components of the loading device are the transfer flask, the shielded loading lock, adapter plate and a mobile water tank. The loading device works mechanically with manpower. For the handling of the transfer flask, a crane with a capacity of 5 metric tons is necessary. During installation of the lid the mobile water pool is filled with demineralized water and the shielded loading passage is taken away. After that the lid is put on the cask. After drainage, the mobile water pool is disassembled, and the cask is dewatered. Finally leak tests of all seals are made. The achieved leakage rate was <1.0 x 10{sup -5} Pa x I/s. The work in FRM was done between 03.02.99 and 12.02.99 including a dry run and leak test. (author)

  17. 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)

  18. ROSA-V large scale test facility (LSTF) system description for the third and fourth simulated fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Mitsuhiro; Nakamura, Hideo; Ohtsu, Iwao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2003-03-01

    The Large Scale Test Facility (LSTF) is a full-height and 1/48 volumetrically scaled test facility of the Japan Atomic Energy Research Institute (JAERI) for system integral experiments simulating the thermal-hydraulic responses at full-pressure conditions of a 1100 MWe-class pressurized water reactor (PWR) during small break loss-of-coolant accidents (SBLOCAs) and other transients. The LSTF can also simulate well a next-generation type PWR such as the AP600 reactor. In the fifth phase of the Rig-of-Safety Assessment (ROSA-V) Program, eighty nine experiments have been conducted at the LSTF with the third simulated fuel assembly until June 2001, and five experiments have been conducted with the newly-installed fourth simulated fuel assembly until December 2002. In the ROSA-V program, various system integral experiments have been conducted to certify effectiveness of both accident management (AM) measures in beyond design basis accidents (BDBAs) and improved safety systems in the next-generation reactors. In addition, various separate-effect tests have been conducted to verify and develop computer codes and analytical models to predict non-homogeneous and multi-dimensional phenomena such as heat transfer across the steam generator U-tubes under the presence of non-condensable gases in both current and next-generation reactors. This report presents detailed information of the LSTF system with the third and fourth simulated fuel assemblies for the aid of experiment planning and analyses of experiment results. (author)

  19. Structural assembly effects of Pt nanoparticle-carbon nanotube-polyaniline nanocomposites on the enhancement of biohydrogen fuel cell performance

    Energy Technology Data Exchange (ETDEWEB)

    Hoa, Le Quynh, E-mail: hoa@p.eng.osaka-u.ac.jp [Department of Applied Physics, Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Sugano, Yasuhito; Yoshikawa, Hiroyuki; Saito, Masato [Department of Applied Physics, Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Tamiya, Eiichi, E-mail: tamiya@ap.eng.osaka-u.ac.jp [Department of Applied Physics, Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan)

    2011-11-30

    Graphical abstract: - Abstract: In this work, we designed various polyaniline (PANI) nanocomposites with platinum (Pt) nanoparticle-decorated multi-walled carbon nanotubes (MWCNTs), employed them as anodic catalysts, and studied their structural assembly effects with regard to enhancing biohydrogen fuel cell performance. Of two proposed structures, the PANI/Pt/MWCNTs multilayer nanocomposites showed superior electrocatalytic activities in the hydrogen oxidation reaction and in fuel cell power density relative to the Pt/MWCNTs-PANI core-shell design. These enhancements were attributed to the active interface formed between the Pt nanoparticles and polyaniline nanofibers, where the higher electronic and ionic conductivities of the thin PANI nanofiber layers in contact with Pt active sites were better than with the PANI bound Pt/MWCNTs. We also investigated the change in the electronic state of the composites and the charge-transfer rate caused by varying the structural assembly. Finally, the role of each catalyst component was examined to understand its individual effect on fuel cell performance and to understand its structural assembly effect on enhanced power density.

  20. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly skeleton

    Energy Technology Data Exchange (ETDEWEB)

    Waseem,, E-mail: wazim_me@hotmail.com; Elahi, N.; Murtaza, G.; Siddiqui, A.A.

    2014-01-15

    Highlights: • FE model of CHASNUPP-1 FA Skeleton produced, using Shell181 Element. • Non-linear buckling analysis has been performed. • Structural integrity and stress measurement of FA skeleton is calculated. • Test results obtained at each strain gauge is compared with FE results at same locations. • Results of both studies are comparable, which validate the FE methodology. -- Abstract: A fuel assembly (FA) structure without fuel rods is called FA skeleton which is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the Chashma Nuclear Power Plant-1 (CHASNUPP-1) FA Skeleton at room temperature in air. Non-linear buckling analysis has been performed using ANSYS 13.0, in-order to determine the buckling behavior of the FA skeleton as well as the location/values of the maximum stress intensity and stresses developed in axial direction under applied compression load of 4900 N. The finite element (FE) model of spacer grids, guide thimbles with dash-pots and flow holes, in addition to spot welds between spacer grids and guide thimbles, has been developed using Shell 181 element. The FA skeleton is a non-straight structure. Its actual behavior of the geometry is non-linear. The value of the perturbation force is related to the geometry of the model and/or the tolerance defined for the geometry. Therefore, a sensitivity study has been made to determine the appropriate value of an arbitrary perturbation load. It has been observed that FA skeleton deformation values obtained through FE Analysis and Experiment (Technical Report, 1994a,b) under applied compression load are comparable and show linear behaviors. Therefore, it is confirmed that buckling of FA skeleton will not occur at the specified load. Moreover, the values of stresses obtained at different locations of the guide thimbles are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Since the

  1. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly skeleton

    International Nuclear Information System (INIS)

    Highlights: • FE model of CHASNUPP-1 FA Skeleton produced, using Shell181 Element. • Non-linear buckling analysis has been performed. • Structural integrity and stress measurement of FA skeleton is calculated. • Test results obtained at each strain gauge is compared with FE results at same locations. • Results of both studies are comparable, which validate the FE methodology. -- Abstract: A fuel assembly (FA) structure without fuel rods is called FA skeleton which is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the Chashma Nuclear Power Plant-1 (CHASNUPP-1) FA Skeleton at room temperature in air. Non-linear buckling analysis has been performed using ANSYS 13.0, in-order to determine the buckling behavior of the FA skeleton as well as the location/values of the maximum stress intensity and stresses developed in axial direction under applied compression load of 4900 N. The finite element (FE) model of spacer grids, guide thimbles with dash-pots and flow holes, in addition to spot welds between spacer grids and guide thimbles, has been developed using Shell 181 element. The FA skeleton is a non-straight structure. Its actual behavior of the geometry is non-linear. The value of the perturbation force is related to the geometry of the model and/or the tolerance defined for the geometry. Therefore, a sensitivity study has been made to determine the appropriate value of an arbitrary perturbation load. It has been observed that FA skeleton deformation values obtained through FE Analysis and Experiment (Technical Report, 1994a,b) under applied compression load are comparable and show linear behaviors. Therefore, it is confirmed that buckling of FA skeleton will not occur at the specified load. Moreover, the values of stresses obtained at different locations of the guide thimbles are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Since the

  2. A report on the transport of MTR-type spent fuel assemblies of the Philippine Research Reactor (PRR-1)

    International Nuclear Information System (INIS)

    Fifty one (51) fuel assemblies of mixed enrichment from the Philippine Research Reactor (PRR-1), consisting of 50 spent and 1 fresh, were shipped to the United States last 14 March 1999 under the U.S. Return of Foreign Research Reactor (FRR) fuel policy. The shipment was in line with the U.S. initiative to implement its Record of Decision (ROD) which took effect on 13 May 1996 to accept and manage all FRR uranium fuel of U.S. origin and enriched in the United States. The shipment program would last10 years, ending midnight of 13 May 2006. The ROD provided a 3 year extension period within which to accept FRR spent nuclear fuel (SNF) withdrawn from reactors after 2006. The U.S. policy gave priority to the NPT significance of high enriched U, as the prime target of the return of FRR policy. Classified as a developing country, the Philippines, through the PNRI, signed a contract with the U.S. Department of Energy for the cost-free shipment of PRR-1 spent fuel to the United States. Spent fuel loading and transport operations to the port area lasted seven (7) days, from 8 to 14 March 1999. (Author)

  3. Determination of total plutonium content in spent nuclear fuel assemblies with the differential die-away self-interrogation instrument

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Alexis C. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87544 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 500 S State St., Ann Arbor, MI 48109 (United States); Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87544 (United States); Flaska, Marek; Pozzi, Sara A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 500 S State St., Ann Arbor, MI 48109 (United States)

    2014-11-11

    As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

  4. Graphene-Supported Platinum Catalyst-Based Membrane Electrode Assembly for PEM Fuel Cell

    Science.gov (United States)

    Devrim, Yilser; Albostan, Ayhan

    2016-08-01

    The aim of this study is the preparation and characterization of a graphene-supported platinum (Pt) catalyst for proton exchange membrane fuel cell (PEMFC) applications. The graphene-supported Pt catalysts were prepared by chemical reduction of graphene and chloroplatinic acid (H2PtCl6) in ethylene glycol. X-ray powder diffraction, thermogravimetric analysis (TGA) and scanning electron microscopy have been used to analyze structure and surface morphology of the graphene-supported catalyst. The TGA results showed that the Pt loading of the graphene-supported catalyst was 31%. The proof of the Pt particles on the support surfaces was also verified by energy-dispersive x-ray spectroscopy analysis. The commercial carbon-supported catalyst and prepared Pt/graphene catalysts were used as both anode and cathode electrodes for PEMFC at ambient pressure and 70°C. The maximum power density was obtained for the Pt/graphene-based membrane electrode assembly (MEA) with H2/O2 reactant gases as 0.925 W cm2. The maximum current density of the Pt/graphene-based MEA can reach 1.267 and 0.43 A/cm2 at 0.6 V with H2/O2 and H2/air, respectively. The MEA prepared by the Pt/graphene catalyst shows good stability in long-term PEMFC durability tests. The PEMFC cell voltage was maintained at 0.6 V without apparent voltage drop when operated at 0.43 A/cm2 constant current density and 70°C for 400 h. As a result, PEMFC performance was found to be superlative for the graphene-supported Pt catalyst compared with the Pt/C commercial catalyst. The results indicate the graphene-supported Pt catalyst could be utilized as the electrocatalyst for PEMFC applications.

  5. The AGS Booster vacuum systems

    International Nuclear Information System (INIS)

    The AGS Booster is a synchrotron for the acceleration of both protons and heavy ions. The design pressure of low 10-11 mbar is required to minimize beam loss of the partially stripped heavy ions. To remove contaminants and to reduce outgassing, the vacuum chambers and the components located in them will be chemically cleaned, vacuum fired, baked then treated with nitric oxide. The vacuum sector will be insitu baked to a minimum of 200 degree C and pumped by the combination of sputter ion pumps and titanium sublimation pumps. This paper describes the design and the processing of this ultra high vacuum system, and the performance of some half-cell vacuum chambers. 9 refs., 7 figs

  6. Tetanus Booster -A missed opportunity

    Directory of Open Access Journals (Sweden)

    Rajesh Gupta

    2014-06-01

    Full Text Available Tetanus a known childhood killer is an entirely preventable disease with tetanus toxoid (TT. Most of the research today is on neonatal tetanus owing to its high case fatality rate. Though the reported mortality with tetanus is lower in older age groups, the management of tetanus is still a challenge in resource constraint settings of developing countries like India. On this background the present study was designed to know the status of tetanus immunization among school going age group children. This was an OPD based survey targeting school going children (age of 5 to 18 years attending Pediatric out-patient department (OPD of a tertiary care teaching hospital. It was observed that out of 636, 299 (47% children were vaccinated for diphtheria, pertusis and tetanus (DPT booster at 5 years of age. Out of 374 children eligible for TT (10 to 16 years only 37 (9.8 % were immunized with TT at age 10 years. Out of 44 children at age of 16 years only 6 (13.6% were immunized. Though there are strategies to immunize school going children under routine immunization programme, official records documented that the immunization coverage for TT was 68% in school going age group. Majority of (80% the cases of tetanus were in non-neonatal age group (mainly school going group in Madhya Pradesh, India. Based on these observations it can be concluded that the tetanus immunization coverage among children of school going age was poor in the given setting.  Tetanus is an acute, spastic paralytic illness historically called lockjaw that is caused by the neurotoxin produced by Clostridium tetani. Tetanus occurs worldwide and is endemic in approximately 90 developing countries. Tetanus is an entirely preventable disease with active immunization with tetanus toxoid. A serum antibody titer of ≥0.01 U/mL is considered protective. [1] In India, according to National immunization schedule, active immunization against tetanus is done with administration of tetanus toxoid as

  7. Experimentation, modelling and simulation of water droplets impact on ballooned sheath of PWR core fuel assemblies in a LOCA situation

    International Nuclear Information System (INIS)

    In a pressurized water reactor (PWR), during a Loss Of Coolant Accident (LOCA), liquid water evaporates and the fuel assemblies are not cooled anymore; as a consequence, the temperature rises to such an extent that some parts of the fuel assemblies can be deformed resulting in 'ballooned regions'. When reflooding occurs, the cooling of these partially blocked parts of the fuel assemblies will depend on the coolant flow that is a mixture of overheated vapour and under-saturated droplets. The aim of this thesis is to study the heat transfer between droplets and hot walls of the fuel rods. In this purpose, an experimental device has been designed in accordance with droplets and wall features (droplet velocity and diameter, wall temperature) representative of LOCA conditions. The cooling of a hot Nickel disk, previously heated by induction, is cooled down by a stream of monodispersed droplet. The rear face temperature profiles are measured by infrared thermography. Then, the estimation of wall heat flux is performed by an inverse conduction technique from these infrared images. The effect of droplet dynamical properties (diameter, velocity) on the heat flux is studied. These experimental data allow us to validate an analytical model of heat exchange between droplet and hot slab. This model is based on combined dynamical and thermal considerations. On the one hand, the droplet dynamics is considered through a spring analogy in order to evaluate the evolution of droplet features such as the spreading diameter when the droplet is squeezed over the hot surface. On the other hand, thermal parameters, such as the thickness of the vapour cushion beneath the droplet, are determined from an energy balance. In the short term, this model will be integrated in a CFD code (named NEPTUNE-CFD) to simulate the cooling of a reactor core during a LOCA, taking into account the droplet/wall heat exchange. (author)

  8. Infinite fuel element simulation of pin power distributions and control blade history in a BWR fuel assembly

    International Nuclear Information System (INIS)

    Pellet-Cladding Interaction (PCI) is a well known effect in fuel pins. One possible reason for PCI-effects could be local power excursions in the fuel pins, which can led to a rupture of the fuel cladding tube. From a reactor safety point of view this has to be considered as a violence of the barrier principal in order to retain fission products in the fuel pins. This paper focuses on the pin power distributions in a 2D infinite lattice of a BWR fuel element. Lots of studies related PCI effect can be found in the literature. In this compact, coupled neutronic depletion calculations taking the control history effect into account are described. Depletion calculations of an infinite fuel element of a BWR were carried out with controlled, uncontrolled and temporarily controlled scenarios. Later ones are needed to describe the control blade history (CBH) effect. A Monte-Carlo approach is mandatory to simulate the neutron physics. The VESTA code was applied to couple the Monte-Carlo-Code MCNP(X) with the burnup code ORIGEN. Additionally, CASMO-4 is also employed to verify the method of simulation results from VESTA. The cross sections for Monte Carlo and burn-up calculations are derived from ENDF/B-VII.0. (orig.)

  9. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    International Nuclear Information System (INIS)

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s−1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  10. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Science.gov (United States)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-11-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s-1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  11. Fission gas release modelling: Developments arising from instrumented fuel assemblies, out-of-pile experiments and microstructural observations

    International Nuclear Information System (INIS)

    This paper reviews the development of fission gas release modelling in thermal reactor fuel (both steady-state and transient) and in particular, illustrates the way in which experimental data have been (and continue to be) the main driving force behind model development. To illustrate this point various aspects of fuel performance are considered: temperature calculation, steady state and transient fission gas release, grain boundary gas atom capacity and microstructural phenomena. The sources of experimental data discussed include end-of-life fission gas release measurements, instrumented fuel assemblies (e.g. rods with internal pressure transducers, fuel centre thermocouples), swept capsule experiments, out-of-pile annealing experiments and microstructural techniques applied during PIE. In the case of the latter, the benefit of applying many observation/analysis techniques on the same fuel samples (the approach adopted at NRL-Windscale) is stressed. This illustrates a shift of emphasis in the modelling field from the development of large, complex thermo-mechanical computer codes to the assessment of key experimental data in order to develop and evaluate sub-models which correctly predict the observed behaviour. (author). 26 refs, 7 figs, 1 tab

  12. Fission gas release modelling: developments arising from instrumented fuel assemblies, out-of-pile experiments and microstructural observations

    International Nuclear Information System (INIS)

    This paper reviews the development of fission gas release modelling in thermal reactor fuel (both steady-state and transient) and in particular, illustrates the way in which experimental data have been, and continue to be, the main driving force behind model development. To illustrate this point various aspects of fuel performance are considered: temperature calculation, steady-state and transient fission gas release, grain boundary gas atom capacity and microstructural phenomena. The sources of experimental data discussed include end-of-life fission gas release measurements, instrumented fuel assemblies (e.g. rods with internal pressure transducers, fuel centre thermocouples), swept capsule experiments, out-of-pile annealing experiments and microstructural techniques applied during post-irradiation evaluation. In the case of the latter, the benefit of applying many observation and analysis techniques on the same fuel samples (the approach adopted at NRL Windscale) is emphasized. This illustrates a shift of emphasis in the modelling field from the development of large, complex thermo-mechanical computer codes to the assessment of key experimental data in order to develop and evaluate sub-models which correctly predict the observed behaviour. (author)

  13. Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Adorni, Martina [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy)]. E-mail: martina_adorni@tin.it; Bousbia-Salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); Hamidouche, Tewfik [Commissariat a l' Energie Atomique, Centre de Recherche Nucleaire d' Alger-Algeria, 02 Boulevard Frantz fanon, BP 399 Alger-gare (Algeria); Maro, Beniamino Di [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); Pierro, Franco [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); D' Auria, Francesco [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy)

    2005-10-15

    The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated in the open literature notwithstanding the fat that it constitutes a severe accident that may lead to local dryout and eventually to loss of the FA integrity. The transients herein considered are related to partial and total obstruction of the cooling channel of a single Fuel Assembly of the reactor core. This study constitutes the first step of a larger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic approach of the used thermal-hydraulic system code.

  14. Booster 6-GeV study

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Xi; Ankenbrandt, Charles M.; Pellico, William A.; Lackey, James; Padilla, Rene; /Fermilab; Norem, J.; /Argonne

    2004-12-01

    Since a wider aperture has been obtained along the Booster beam line, this opens the opportunity for Booster running a higher intensity beam than ever before. Sooner or later, the available RF accelerating voltage will become a new limit for the beam intensity. Either by increasing the RFSUM or by reducing the accelerating rate can achieve the similar goal. The motivation for the 6-GeV study is to gain the relative accelerating voltage via a slower acceleration.

  15. Thermal-Hydraulic Calculation for Simplified Fuel Assembly of Super Fast Reactor Using Two-Fluid Model Analysis Code ACE-3D

    International Nuclear Information System (INIS)

    To evaluate thermal hydraulic characteristics of a fuel assembly of supercritical water-cooled fast reactor (Super Fast Reactor), a simplified fuel assembly was analyzed with a three-dimensional two-fluid model analysis code ACE-3D which has been enhanced by Japan Atomic Energy Agency. In the ACE-3D code, the two-phase flow turbulent model based on the k-ε model were adopted. The analytical geometry simulates a 19-rod fuel assembly, which is a simplified geometry of the 271-rod fuel assembly and includes all three kinds of different subchannel types; (1): adjoining to the channel box, (2): next to type (1), and (3): located inside types (1) and (2). In this calculation, one-twelfth model is adopted as the computational domain taking advantage of symmetry. As the boundary conditions, mass velocity, inlet enthalpy and power per rod are to be the same as the steady state condition of the Super Fast Reactor. Cross-sectional local power distribution in the fuel assembly is set to be flat. Rod surface temperatures take peak values near the top of the rods. Maximum clad surface temperature (MCST) is observed at the position facing to the narrowest gap on the center rod near the outlet and the value is 902 K (629 deg. C). It was confirmed that the predicted MCST satisfies a thermal design criteria to ensure fuel and cladding integrity: the MCST should be less than 650 deg. C. (author)

  16. Characterization of long-lived activation products in spent fuel assembly hardware and reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Extensive measurements have provided the basis for evaluating the radionuclide concentrations, distributions, inventories, waste classification, and disposal options for activated metal wastes generated during reactor decommissioning. A variety of neutron-activated metal specimens associated with spent fuel assembly hardware from commercial nuclear power stations and pressure vessel steel from the decommissioned Gundremmingen KRB-A reactor were subjected to detailed radionuclide and stable element analyses. Emphasis was placed on the long-lived radionuclides specified in 10CFR61, including Mn-54, Fe-55, Co-60, Ni-59, Ni-63 and Nb-94. In addition, it was discovered that much higher concentrations of Nb-93m were present in activated Inconel and stainless steel than earlier calculations had predicted. The concentrations of Ni-63, Ni-59, and Nb-94 in Inconel components, and Ni-63 and Ni-59 in stainless steel components were often much greater than the Class C limit, indicating that these materials would have to be disposed of as high level waste. The accuracy of calculational methods for predicting radionuclide concentrations in activated metal wastes was evaluated by conducting blind comparisons of empirical versus predicted values. This comparison showed that good agreements were achieved for the fueled regions of the fuel assemblies, but at the tops and bottoms of the assemblies the calculated values were, in some cases, significantly in error. The agreement between measured versus predicted radionuclide concentrations for the Gundremmingen pressure vessel steel was good. These evaluations have provided confidence in the calculational methods and have identified problem areas where improvements are warranted. (orig.)

  17. Studying the vibration and random hydrodynamic loads on the fuel rods bundles in the fuel assemblies of the reactor installations used at nuclear power stations equipped with VVER reactors

    Science.gov (United States)

    Solonin, V. I.; Perevezentsev, V. V.

    2012-05-01

    Random hydrodynamic loads causing vibration of fuel rod bundles in a turbulent flow of coolant are obtained from the results of pressure pulsation measurements carried out over the perimeter of the external row of fuel rods in the bundle of a full-scale mockup of a fuel assembly used in a second-generation VVER-440 reactor. It is shown that the turbulent flow structure is a factor determining the parameters of random hydrodynamic loads and the vibration of fuel rod bundles excited by these loads. The results from a calculation of random hydrodynamic loads are used for estimating the vibration levels of fuel rod bundles used in prospective designs of fuel assemblies for VVER reactors.

  18. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  19. High Performance Fuel Cell and Electrolyzer Membrane Electrode Assemblies (MEAs) for Space Energy Storage Systems

    Science.gov (United States)

    Valdez, Thomas I.; Billings, Keith J.; Kisor, Adam; Bennett, William R.; Jakupca, Ian J.; Burke, Kenneth; Hoberecht, Mark A.

    2012-01-01

    Regenerative fuel cells provide a pathway to energy storage system development that are game changers for NASA missions. The fuel cell/ electrolysis MEA performance requirements 0.92 V/ 1.44 V at 200 mA/cm2 can be met. Fuel Cell MEAs have been incorporated into advanced NFT stacks. Electrolyzer stack development in progress. Fuel Cell MEA performance is a strong function of membrane selection, membrane selection will be driven by durability requirements. Electrolyzer MEA performance is catalysts driven, catalyst selection will be driven by durability requirements. Round Trip Efficiency, based on a cell performance, is approximately 65%.

  20. Comparisons of prediction methods for peak cladding temperature and effective thermal conductivity in spent fuel assemblies of transportation/storage casks

    International Nuclear Information System (INIS)

    Highlights: • Peak cladding temperature (PCT) of spent fuel were evaluated by various methods. • The methods are Wooton–Epstein correlation, two-region model, and CFD. • Temperature difference between two-region and CFD ranges from −0.2 to 9 K. • CFD could be used to calculate PCT because of over-predicting PCT of two-region. • Application using CFD was conducted for spent fuel assembly used in Republic of Korea. - Abstract: When spent fuel assemblies from the reactor of nuclear power plants (NPPs) are transported or stored, the assemblies are exposed to a variety of environments that can affect the peak cladding temperature. There are three models to calculate the peak cladding temperature of spent fuel assemblies in a cask: Manteufel and Todreas’s two-region model, Bahney Lotz’s effective thermal conductivity model, and Wooton–Epstein correlation. The peak cladding temperatures of Babcock and Wilcox (B and W) 15 × 15 PWR spent fuel assembly under helium backfill gas were evaluated by using two-dimensional CFD simulation and compared with two models (Wooton–Epstein correlation, two-region model). The peak cladding temperature difference between the two-region model and CFD simulation ranges from −0.2 K to 9 K. Two-region model over-predicts the measured peak cladding temperature that performs in a spent fuel dry storage cask. Therefore the simulation could be used to calculate peak cladding temperature of spent fuel assemblies. Application using CFD simulation was conducted to investigate the peak cladding temperature and effective thermal conductivity of spent fuel assembly used in Korea NPPs: 16 × 16 (CE type) and 17 × 17 (WH type) PWR spent fuel assembly. CFD simulation results are similar to each other, and the difference of temperature drop between the three arrays occurs slightly in all basket wall temperatures. The effective thermal conductivity calculated from the 16 × 16 PWR spent fuel assembly results was more conservative

  1. On the evaluation of the pressure losses in a lead-bismuth-eutectics cooled fuel assembly with TRACE and SUSA

    International Nuclear Information System (INIS)

    The prediction of the pressure drop in a pool-type reactor operated with lead-bismuth-eutectics is of crucial importance. A pressure drop of e.g. 1 bar is equivalent to a lead-bismuth-eutectics column of about 1 m, which has a big influence on the financial aspects of the design proposal. The paper presents results on the hydraulic evaluation of a fuel assembly with the emphasis on uncertainties and variations of relevant parameters like the mass flow rate, form, and friction loss coefficients. With the subsequent uncertainty and sensitivity study, in connection with thermal hydraulic investigations, the influence of these uncertain parameters was evaluated. (author)

  2. Seismic Shaking Table Requirements and Consideration of Fluid-Structure Interaction Effect in Seismic Response Analysis Model for In-Reactor Fuel Assembly Under Severe Earthquake Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kanghee; Yoon, Kyungho; Kang, Heungsoek; Lee, Youngho; Kim, Hyungkyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Dynamic response of fuel assembly can be significantly affected by added hydrodynamic mass and additional damping from the fluid and flow inside operating reactor core. Added mass or hydrodynamic virtual mass from surrounding fluid medium can be theoretically estimated by the potential flow theory. Solving Laplace equation in terms of velocity potential can leads to calculate mass components in the mass matrix of simplified fuel FE model. Additional damping from the fluid and the flow inside reactor core are originated from fluid drag and flow lift force, respectively. Lift force from axial flow can increase fuel assembly damping by twice compared to still fluid damping from the loop testing. In practice, fuel assembly damping should be measured by mockup loop testing and referred to published data in the literature. The justification is performed via time history analysis with simplified dynamic model using a group of fuel assembly in the core. Key check points in this analysis might be the integrity of intermediate spacer grids when impacting fuels into core shroud plate or into neighboring fuel assembly. Thus, dynamic displacement and impact force at grid elevations are the important structural parameters to be traced out during the analysis and the simulation testing. KAERI have a plan to develop dynamic analysis model and to setup test infrastructure for full scale and several fuel assembly rows seismic simulation testing. This paper briefly discuss on the reference earthquake accident scenario, shaking table requirements for full-scale seismic simulation testing, virtual testing issues before the hardware setup, and modelling issue related to fluid-structure interaction effect in accident core analysis.

  3. Assessing the Feasibility of Using Neutron Resonance Transmission Analysis (NRTA) for Assaying Plutonium in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Chichester; J. W. Sterbentz

    2012-07-01

    Neutron resonance transmission analysis (NRTA) is an active-interrogation nondestructive assay (NDA) technique capable of assaying spent nuclear fuel to determine plutonium content. Prior experimental work has definitively shown the technique capable of assaying plutonium isotope composition in spent-fuel pins to a precision of approximately 3%, with a spatial resolution of a few millimeters. As a Grand Challenge to investigate NDA options for assaying spent fuel assemblies (SFAs) in the commercial fuel cycle, Idaho National Laboratory has explored the feasibility of using NRTA to assay plutonium in a whole SFA. The goal is to achieve a Pu assay precision of 1%. The NRTA technique uses low-energy neutrons from 0.1-40 eV, at the bottom end of the actinide-resonance range, in a time-of-flight arrangement. Isotopic composition is determined by relating absorption of the incident neutrons to the macroscopic cross-section of the actinides of interest in the material, and then using this information to determine the areal density of the isotopes in the SFA. The neutrons used for NRTA are produced using a pulsed, accelerator-based neutron source. Distinguishable resonances exist for both the plutonium (239,240,241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Based on extensive modeling of the problem using Monte Carlo-based simulation codes, our preliminary results suggest that by rotating an SFA to acquire four symmetric views, sufficient neutron transmission can be achieved to assay a SFA. In this approach multiple scan information for the same pins may also be unfolded to potentially allow the determination of plutonium for sub-regions of the assembly. For a 17 ? 17 pressurized water reactor SFA, a simplistic preliminary

  4. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study

    International Nuclear Information System (INIS)

    A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided

  5. Fine-mesh deterministic modeling of PWR fuel assemblies: Proof-of-principle of coupled neutronic/thermal–hydraulic calculations

    International Nuclear Information System (INIS)

    Highlights: • We implemented a fine-mesh coupled neutronic/thermal–hydraulic tool. • A CFD approach is used together with the multi-group neutron diffusion approximation. • Temperature-dependent cross-sections are generated with a Monte Carlo method. • We applied the tool to a simplified PWR fuel assembly. • Discrepancies in multiplication factor are seen against radial coarse-mesh averaging. - Abstract: This paper investigates the feasibility of developing a fine mesh coupled neutronic/thermal–hydraulic solver within the same computing platform for selected fuel assemblies in nuclear cores. As a first step in this developmental work, a Pressurized Water Reactor at steady-state conditions was considered. The system being simulated has a finite axial size, but is infinite in the radial direction. The platform used for the modeling is based on the open source C++ library OpenFOAM. The thermal–hydraulics is solved using the built-in SIMPLE algorithm for the mass and momentum fields of the fluid, complemented by an equation for the temperature field applied simultaneously to all the regions (i.e. fluid and solid structures). For the neutronics, a two-group neutron diffusion-based solver was developed, with sets of macroscopic cross-sections generated by the Monte Carlo code SERPENT. The meshing of the system was created by the open source software SALOME. Successful convergence of the neutronic and thermal–hydraulic fields was achieved, thus bringing the solution of the coupled problem to an unprecedented level of details. Most importantly, the true interdependence of the different fields is automatically guaranteed at all scales. In addition, comparisons with a coarse-mesh radial averaging of the thermal–hydraulic variables show that a coarse-mesh fuel temperature identical for all fuel pins can lead to discrepancies of up to 0.5% in pin powers, and of several tens of pcm in multiplication factor

  6. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1997-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement.

  7. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO2 and UO2), typically containing 95% or more UO2. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement

  8. Assembly of a Cost-Effective Anode Using Palladium Nanoparticles for Alkaline Fuel Cell Applications

    Science.gov (United States)

    Feliciano-Ramos, Ileana; Casan~as-Montes, Barbara; García-Maldonado, María M.; Menendez, Christian L.; Mayol, Ana R.; Díaz-Vazquez, Liz M.; Cabrera, Carlos R.

    2015-01-01

    Nanotechnology allows the synthesis of nanoscale catalysts, which offer an efficient alternative for fuel cell applications. In this laboratory experiment, the student selects a cost-effective anode for fuel cells by comparing three different working electrodes. These are commercially available palladium (Pd) and glassy carbon (GC) electrodes, and…

  9. Feasibility study of Large-scale helium GFR employing coated particle fuel. Design study of hexagonal matrix-block fuel assembly cores. Annual report of JFY2004

    International Nuclear Information System (INIS)

    Gas-cooled fast reactor has been taken an interest as a future nuclear reactor power source; we JNC have designed large-scale (1124MW electric power), high-temperature (850degC) with high thermodynamic efficiency (around 47%), helium-cooled fast reactors as a part of feasibility study. Hexagonal-block fuel assembly configuration has been considered as a candidate, where a number of coated particles packed in a SiC matrix are cooled indirectly by ascending flow in penetrating tubes. This manuscript describes, as an annual report of JFY2004, technical keys for enhancing core neutronics/thermal-hydraulics performances of the hex-block concepts and the best-to-date core designs. Technical keys for enhancing are from neutronics viewpoints (radial peaking factor, assembly layout and configuration), thermal-hydraulic viewpoints (heat transfer and film temperature rise, thermal conductivity and fuel-matrix temperature rise, core pressure drop, and coolant temperature), and neutronics/and/thermal-hydraulics combinational viewpoints (effective fuel volume fraction, local temperature decrees for increasing Doppler effects). Two issues (reducing contacting heat resistance in a matrix, and lowering core inflow temperature under depressurization transient) are selected from quantitative pre-evaluations; then are applied to select the reference core designs for achieving reduced Pu fissile inventory and improved average discharge burnup. Two reference cores are designed; one is 'breeding' core, which achieves high breeding ratio and high discharge burnup, the other is 'break-even' core, which brings much higher discharge burnup with a breeding ratio of around unity. [Reference core designs (2400MW thermal/1124MW electric outputs)] Core equiv. diameter/Height/Outermost diameter: 5.42m/1.00m/7.49m. Average Discharge Burnup (Seed region average): 121 GWd/t/123 GWd/t. (Entire core average): 69 GWd/t/89 GWd/t. Breeding Ratio: 1.11/1.03. Initial Pu fissile quantity: 7.0 ton/GWe/7.0 ton

  10. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  11. Design and fabrication of remote welding system for the fuel bundle assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.S.; Lee, J.W.; Park, G.I. [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

    2011-07-01

    Remote fuel bundle welding equipment in a hot-cell was designed and fabricated. To achieve this, a preliminary investigation of hands-on fuel fabrication outside a hot-cell was conducted with a consideration of the constraints caused by welding in a hot-cell. Some basic experiments were also carried out to improve the end-plate welding process for fuel bundle fabrication. The resistance welding equipment using end-plate welding was also improved. It was found that resistance welding was more suitable for joining an end-plate to end caps in a hot-cell. The optimum conditions for end-plate welding for remote operation were also obtained. Preliminary performances to improve the resistance welding process were also examined, and the resistance welding process was determined to be the best in the hot-cell environment for fuel bundle fabrication. The greatest advantage of fuel bundle welding equipment would be a commercialized welding process in which there is extensive production experience. This paper presents an outline of the developed welding equipment for a fuel bundle fabrication and reviews the conceptual design of remote welding equipment using a master-slave manipulator. The design of the remote welding equipment using the Pro-Engineer method was also reviewed. Furthermore the mechanical considerations and a mock-up simulation test were described. Finally, its performance test results were presented for a mock-up of remote fuel bundle welding equipment. (author)

  12. OECD/NEA burnup credit criticality benchmarks phase IIIB. Burnup calculations of BWR fuel assemblies for storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-02-01

    The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

  13. Distribution of fission products in graphite sleeves and blocks of the ninth and tenth OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    Distribution of fission products in graphite sleeves and blocks of the ninth and tenth OGL-1 fuel assemblies was measured by gamma spectrometry with lathe sectioning. The assemblies were loaded with HTGR fuel compacts, which had been produced by a scaled-up facility for the High Temperature Engineering Test Reactor (HTTR) being developed by JAERI; and they were irradiated in an in-pile gas loop, OGL-1. Fission products detected in the sleeves were 137Cs, 134Cs, 155Eu, 154Eu, 144Ce, 125Sb and 110mAg. The last nuclide, however, may have been produced by activation of a stable isotope, 109Ag, contained as impurity. Effective retention capability of the sleeve was observed for 155Eu, 154Eu, 144Ce and 125Sb; while, not for 137Cs and 134Cs. Concentration of 137Cs in the graphite blocks was markedly higher at the downstream side than at the upstream side of the coolant. This was ascribed to migration of the nuclide with the coolant flow and its subsequent sorption on the surface of the block. (author)

  14. Fuel assemblies with inert matrices as reloads of cycle 11 of the Unit 1 of the LVNC; Ensamble combustibles con matrices inertes como recargas del ciclo 11 de la Unidad 1 de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Hernandez M, N.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2005-07-01

    In this work the results that were obtained of the analysis of three different reloads of the cycle 11 with fuel assemblies containing a mixture of UO{sub 2} and plutonium grade armament in an inert matrix. The proposed assemble, consists of an arrangement 10x10 with 42 bars fuels of PuO{sub 2}-CeO{sub 2}, 34 fuel bars with UO{sub 2} and 16 fuel bars with UO{sub 2}-Gd{sub 2O}3. The proposed assemble is equivalent to an it reloadable assemble of the cycle 11. The fuel bars of uranium and gadolinium, are of the same type of those that are used in the reloadable assemble of uranium. The design and generation of the nuclear databases of the fuel cell with mixed fuel, it was carried out with the HELIUMS code. The simulation of operation of the cycle 11, it was carried out with the CM-PRESTO code. The results show that with one reload of 72 assemblies of UO{sub 2} and 32 assemblies with mixed fuel has a cycle length of smaller in 10.5 days to the cycle length with the complete reload of assemblies of UO{sub 2} and a length smaller cycle in 34 days with the complete reload of 104 assemblies with mixed fuel. (Author)

  15. Electron beam welding and laser welding of FRAGEMA fuel assembly components

    International Nuclear Information System (INIS)

    Neutron balance and activity of the primary coolant circuit are improved in PWR if inconel 718 is replaced by a zirconium alloy for fuel element grids. This paper examines laser welding and EB welding of these zirconium alloy grids

  16. Irradiation Test in HANARO for an Evaluation of In-Core Performance of the Parts of Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H.; Choi, M. H.; Cho, J. M.; Son, M. H.; Choi, M. H.; Shin, Y. T.; Park, S. J

    2007-10-15

    For an evaluation of the neutron irradiation properties of the parts of a nuclear fuel assembly requested by KNFC (Korea Nuclear Fuel Co., Ltd.), the 05M-07U instrumented capsule was designed, fabricated, and successfully irradiated at HANARO. The basic structure of the 05M-07U capsule was based on the 04M-17U capsule which had been successfully irradiated in HANARO as part of the 2004 project. However, because of a limited number of specimens and the budget of one university, the remaining space in the capsule was filled with various KAERI specimens. 88 specimens such as 1x1 spacer grid, spring, buckling, growth and tensile specimens of Zirlo and Inconel alloys were inserted in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. The capsule was irradiated in the CT test hole of HANARO of a 30MW thermal output at 270-400 .deg. C up to a fast neutron fluence of 5.7x10{sup 20}(n/cm{sup 2}) (E>1.0MeV). After an irradiation test, the main body of the capsule was cut off at the bottom of the protection tube with a cutting system and it was transported to the IMEF (Irradiated Materials Examination Facility). The irradiated specimens were tested to evaluate the irradiation performance of the parts of a fuel assembly in the IMEF hot cell.

  17. Solid Rocket Booster (SRB) Flight System Integration at Its Best

    Science.gov (United States)

    Wood, T. David; Kanner, Howard S.; Freeland, Donna M.; Olson, Derek T.

    2011-01-01

    The Solid Rocket Booster (SRB) element integrates all the subsystems needed for ascent flight, entry, and recovery of the combined Booster and Motor system. These include the structures, avionics, thrust vector control, pyrotechnic, range safety, deceleration, thermal protection, and retrieval systems. This represents the only human-rated, recoverable and refurbishable solid rocket ever developed and flown. Challenges included subsystem integration, thermal environments and severe loads (including water impact), sometimes resulting in hardware attrition. Several of the subsystems evolved during the program through design changes. These included the thermal protection system, range safety system, parachute/recovery system, and others. Because the system was recovered, the SRB was ideal for data and imagery acquisition, which proved essential for understanding loads, environments and system response. The three main parachutes that lower the SRBs to the ocean are the largest parachutes ever designed, and the SRBs are the largest structures ever to be lowered by parachutes. SRB recovery from the ocean was a unique process and represented a significant operational challenge; requiring personnel, facilities, transportation, and ground support equipment. The SRB element achieved reliability via extensive system testing and checkout, redundancy management, and a thorough postflight assessment process. However, the in-flight data and postflight assessment process revealed the hardware was affected much more strongly than originally anticipated. Assembly and integration of the booster subsystems required acceptance testing of reused hardware components for each build. Extensive testing was done to assure hardware functionality at each level of stage integration. Because the booster element is recoverable, subsystems were available for inspection and testing postflight, unique to the Shuttle launch vehicle. Problems were noted and corrective actions were implemented as needed

  18. A balance procedure for calculating the model fuel assemblies reflooding during design basis accident and its verification on PARAMETER test facility

    Science.gov (United States)

    Bazyuk, S. S.; Ignat'ev, D. N.; Parshin, N. Ya.; Popov, E. B.; Soldatkin, D. M.; Kuzma-Kichta, Yu. A.

    2013-05-01

    A balance procedure is proposed for estimating the main parameters characterizing the process of model fuel assemblies reflooding of a VVER reactor made on different scales under the conditions of a design basis accident by subjecting them to bottom reflooding1. The proposed procedure satisfactorily describes the experimental data obtained on PARAMETER test facility in the temperature range up to 1200°C. The times of fuel assemblies quenching by bottom reflooding calculated using the proposed procedure are in satisfactory agreement with the experimental data obtained on model fuel assemblies of VVER- and PWR-type reactors and can be used in developing measures aimed at enhancing the safety of nuclear power stations.

  19. Computer simulation of thermal-hydraulics of MNSR fuel-channel assembly using LabView

    International Nuclear Information System (INIS)

    A LabView simulator of thermal hydraulics has been developed to demonstrate the temperature profile of coolant flow in the reactor core during normal operation. The simulator could equally be used for any transient behaviour of the reactor. Heat generation, transfer and the associated temperature profile in the fuel-channel elements viz: the coolant, cladding and fuel were studied and the corresponding analytical temperature equations in the axial and radial directions for the coolant, outer surface of the cladding, fuel surface and fuel center were obtained for the simulation using LabView. Tables of values for the equations were constructed by MATLAB and excel software programs. Plots of the equations with LabView were verified and validated with the graphs drawn by the MATLAB. In this thesis, an analysis of the effects of the coolant inlet temperature of 24.5°C and exit temperature of 70.0° on the temperature distribution in fuel-channel elements of the reactor core of cylindrical geometry was carried out. Other parameters, including the total fuel channel power, mass flow rate and convective heat transfer coefficient were varied to study the effects on the temperature profile. The analytical temperature equations in the fuel channel elements of the reactor core were obtained. MATLAB and Excel software were used to construct data for the equations. The plots by MATLAB were used to benchmark the LabVIEW simulation. Excellent agreement was obtained between the MATLAB plots and the LabView simulation results with an error margin of 0.001. The analysis of the results by comparing gradients of inlet temperature, total reactor channel power and mass flow indicated that inlet temperature gradient is one of the key parameters in determining the temperature profile in the MNSR core. (au)

  20. In-pile release behavior of metallic fission products in graphite materials of an HTGR fuel assembly

    International Nuclear Information System (INIS)

    Distribution of metallic fission products in the graphite sleeve and block of the fifth OGL-1 fuel assembly was measured by gamma spectrometry with lathe sectioning. Considerably large release fractions of long-lived fission products with smooth axial profiles were observed in the sleeve due to a large failure fraction of coated fuel particles accompanied with failed silicon carbide layers. Nevertheless, a key nuclide 110mAg, whose large release is suspected at increased burnups for low-enriched uranium fuels, was effectively retained within the graphite sleeve. The retention was also observed for 125Sb, 154Eu and 155Eu up to a burnup of 3.2% fission per initial metal atom, but was limited for 134Cs and 137Cs at high sleeve-temperatures above 9000C. In-pile diffusion coefficients in IG-110 graphite have been evaluated for Cs, Ag and Sb; those for Cs are in reasonable agreement with available in-pile data. (orig.)

  1. Membrane electrode assembly with doped polyaniline interlayer for proton exchange membrane fuel cells under low relative humidity conditions

    Energy Technology Data Exchange (ETDEWEB)

    Cindrella, L. [Fuel Cell Research Lab, Engineering Technology Department, Arizona State University, Mesa, AZ 85212 (United States); Department of Chemistry, National Institute of Technology, Tiruchirappalli, Tamil Nadu 620015 (India); Kannan, A.M. [Fuel Cell Research Lab, Engineering Technology Department, Arizona State University, Mesa, AZ 85212 (United States)

    2009-09-05

    A membrane electrode assembly (MEA) was designed by incorporating an interlayer between the catalyst layer and the gas diffusion layer (GDL) to improve the low relative humidity (RH) performance of proton exchange membrane fuel cells (PEMFCs). On the top of the micro-porous layer of the GDL, a thin layer of doped polyaniline (PANI) was deposited to retain moisture content in order to maintain the electrolyte moist, especially when the fuel cell is working at lower RH conditions, which is typical for automotive applications. The surface morphology and wetting angle characteristics of the GDLs coated with doped PANI samples were examined using FESEM and Goniometer, respectively. The surface modified GDLs fabricated into MEAs were evaluated in single cell PEMFC between 50 and 100% RH conditions using H{sub 2} and O{sub 2} as reactants at ambient pressure. It was observed that the MEA with camphor sulfonic acid doped PANI interlayer showed an excellent fuel cell performance at all RH conditions including that at 50% at 80 C using H{sub 2} and O{sub 2}. (author)

  2. Experimental studies of local coolant hydrodynamics using a scaled model of cassette-type fuel assembly of a KLT-40S reactor

    Science.gov (United States)

    Dmitriev, S. M.; Barinov, A. A.; Varentsov, A. V.; Doronkov, D. V.; Solntsev, D. N.; Khrobostov, A. E.

    2016-08-01

    The results of experimental studies of local hydrodynamic and mass exchange characteristics of the coolant flow behind the spacer grid in the fuel assembly of a KLT-40S reactor are presented. The experiments were aimed at the investigation of representative domains of the fuel assembly with three tracer injection regions. The studies were performed at the aerodynamic test facility using the tracer gas diffusion method. According to the theory of hydrodynamic similarity, the obtained experimental results can be transferred to full-scale coolant flow conditions in standard fuel assemblies. The analysis of the tracer concentration propagation made it possible to determine in detail the flow pattern and find the main regularities and specific features of the coolant flow behind the plate spacer grid of KLT-40S fuel assembly. The hydraulic resistance coefficient of the spacer grid was experimentally determined. The coefficients of mass exchange between cells for representative cells of the displacer region in the KLT-40S fuel assembly were calculated for the first time; these results are presented in the form of the "mixing matrix." The results of studies of local coolant flow hydrodynamics in the KLT-40S fuel assembly are used at AO Afrikantov OKBM for estimation of thermotechnical reliability of active cores for reactors of floating nuclear power stations. The experimental data on hydrodynamic and mass exchange characteristics are included in the database for verification of CDF codes and detailed cell-wise calculation of the active core for KLT-40S reactor installation. The results of these studies can be used at FSUE RFNC-VNIIEF for testing and verification of domestic three-dimensional hydrodynamic CFD codes ("Logos") that are applied for substantiation of newly designed reactor installations. Practical recommendations on the application of the obtained results in thermohydraulic calculations of the active core for the KLT-40S reactor will be worked out. Proposals

  3. Templated assembly of photoswitches significantly increases the energy-storage capacity of solar thermal fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kucharski, TJ; Ferralis, N; Kolpak, AM; Zheng, JO; Nocera, DG; Grossman, JC

    2014-04-13

    Large-scale utilization of solar-energy resources will require considerable advances in energy-storage technologies to meet ever-increasing global energy demands. Other than liquid fuels, existing energy-storage materials do not provide the requisite combination of high energy density, high stability, easy handling, transportability and low cost. New hybrid solar thermal fuels, composed of photoswitchable molecules on rigid, low-mass nanostructures, transcend the physical limitations of molecular solar thermal fuels by introducing local sterically constrained environments in which interactions between chromophores can be tuned. We demonstrate this principle of a hybrid solar thermal fuel using azobenzene-functionalized carbon nanotubes. We show that, on composite bundling, the amount of energy stored per azobenzene more than doubles from 58 to 120 kJ mol(-1), and the material also maintains robust cyclability and stability. Our results demonstrate that solar thermal fuels composed of molecule-nanostructure hybrids can exhibit significantly enhanced energy-storage capabilities through the generation of template-enforced steric strain.

  4. Templated assembly of photoswitches significantly increases the energy-storage capacity of solar thermal fuels

    Science.gov (United States)

    Kucharski, Timothy J.; Ferralis, Nicola; Kolpak, Alexie M.; Zheng, Jennie O.; Nocera, Daniel G.; Grossman, Jeffrey C.

    2014-05-01

    Large-scale utilization of solar-energy resources will require considerable advances in energy-storage technologies to meet ever-increasing global energy demands. Other than liquid fuels, existing energy-storage materials do not provide the requisite combination of high energy density, high stability, easy handling, transportability and low cost. New hybrid solar thermal fuels, composed of photoswitchable molecules on rigid, low-mass nanostructures, transcend the physical limitations of molecular solar thermal fuels by introducing local sterically constrained environments in which interactions between chromophores can be tuned. We demonstrate this principle of a hybrid solar thermal fuel using azobenzene-functionalized carbon nanotubes. We show that, on composite bundling, the amount of energy stored per azobenzene more than doubles from 58 to 120 kJ mol-1, and the material also maintains robust cyclability and stability. Our results demonstrate that solar thermal fuels composed of molecule-nanostructure hybrids can exhibit significantly enhanced energy-storage capabilities through the generation of template-enforced steric strain.

  5. Templated assembly of photoswitches significantly increases the energy-storage capacity of solar thermal fuels.

    Science.gov (United States)

    Kucharski, Timothy J; Ferralis, Nicola; Kolpak, Alexie M; Zheng, Jennie O; Nocera, Daniel G; Grossman, Jeffrey C

    2014-05-01

    Large-scale utilization of solar-energy resources will require considerable advances in energy-storage technologies to meet ever-increasing global energy demands. Other than liquid fuels, existing energy-storage materials do not provide the requisite combination of high energy density, high stability, easy handling, transportability and low cost. New hybrid solar thermal fuels, composed of photoswitchable molecules on rigid, low-mass nanostructures, transcend the physical limitations of molecular solar thermal fuels by introducing local sterically constrained environments in which interactions between chromophores can be tuned. We demonstrate this principle of a hybrid solar thermal fuel using azobenzene-functionalized carbon nanotubes. We show that, on composite bundling, the amount of energy stored per azobenzene more than doubles from 58 to 120 kJ mol(-1), and the material also maintains robust cyclability and stability. Our results demonstrate that solar thermal fuels composed of molecule-nanostructure hybrids can exhibit significantly enhanced energy-storage capabilities through the generation of template-enforced steric strain.

  6. Torque strength of an endplate welding due to process parameters using a fuel assembling welder

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Dae-Seo; Kim, Soo-Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    As fuel bundles in a PHWR core irradiated, inner pressure in the claddings of the fuel rods increases owing to the outer pressure and fission products of the nuclear fissions. Because of a leak possibility from a welding between a cladding and end plug, this welding part is connected with the safety of nuclear fuel rods. Endplug-cladding welding of nuclear fuel rods in a PHWR takes advantage of a resistance upset butt welding. The weldment between a cladding and endplug is to be sound to prevent a leakage of fission products from a cladding as a UO{sub 2} pellet is irradiated. Weld flash was made from a deformation due to a welding heat and increasing the pressure of the resistivity and resistance from a cladding and endplug. Weld line of a welding interface, microstructure of a weldment and a crystallographic structure change were sources of an iodine induced SCC in a reactor. The soundness of a weldment is important because a weld line connects the leakage of fission products from an operational reactor. In this study, welding specimens were fabricated by a resistance welding method using a bundle fuel welder to measure and analyze the torque of an endplug-endplate welding. The torque of a weldment between an endplug and endplate was measured and analyzed with the welding time. The weldability of a weldment between an endplug and endplate was investigated by a metallographic examination.

  7. Accelerating RF cavity of the Booster

    CERN Multimedia

    1983-01-01

    Each of the 4 PS Booster rings has a single accelerating cavity.It consists of 2 quarter-wave ferrite-loaded resonators. 2 figure-of-eight loops tune the frequency throughout the accelerating cycle, from 3 to 8 MHz (from 50 MeV at injection to the original Booster energy of 800 MeV, 2 GeV today). The cavities have a flat design, to fit the ring-to-ring distance of 36 cm, and are forced-air cooled. The 2 round objects in the front-compartments are the final-stage power-tetrodes. See also 8111095.

  8. Accelerating RF cavity of the Booster

    CERN Multimedia

    CERN PhotoLab

    1981-01-01

    Each of the 4 PS Booster rings has a single accelerating cavity. It consists of 2 quarter-wave ferrite-loaded resonators. There are 2 figure-of-eight loops on the ferrite loads for tuning the frequency throughout the acceleration cycle, from 3 to 8 MHz (from 50 MeV at injection to the original Booster energy of 800 MeV, 2 GeV today). The cavities have a flat design, to fit the ring-to-ring distance of 36 cm. The tube for forced-air cooling is visible in the left front. See also 8301084.

  9. Developing the World's Most Powerful Solid Booster

    Science.gov (United States)

    Priskos, Alex S.; Frame, Kyle L.

    2016-01-01

    NASA's Journey to Mars has begun. Indicative of that challenge, this will be a multi-decadal effort requiring the development of technology, operational capability, and experience. The first steps are underway with more than 15 years of continuous human operations aboard the International Space Station (ISS) and development of commercial cargo and crew transportation capabilities. NASA is making progress on the transportation required for deep space exploration - the Orion crew spacecraft and the Space Launch System (SLS) heavy-lift rocket that will launch Orion and large components such as in-space stages, habitat modules, landers, and other hardware necessary for deep-space operations. SLS is a key enabling capability and is designed to evolve with mission requirements. The initial configuration of SLS - Block 1 - will be capable of launching more than 70 metric tons (t) of payload into low Earth orbit, greater mass than any other launch vehicle in existence. By enhancing the propulsion elements and larger payload fairings, future SLS variants will launch 130 t into space, an unprecedented capability that simplifies hardware design and in-space operations, reduces travel times, and enhances two solid propellant five-segment boosters, both based on space shuttle technologies. This paper will focus on development of the booster, which will provide more than 75 percent of total vehicle thrust at liftoff. Each booster is more than 17 stories tall, 3.6 meters (m) in diameter and weighs 725,000 kilograms (kg). While the SLS booster appears similar to the shuttle booster, it incorporates several changes. The additional propellant segment provides additional booster performance. Parachutes and other hardware associated with recovery operations have been deleted and the booster designated as expendable for affordability reasons. The new motor incorporates new avionics, new propellant grain, asbestos-free case insulation, a redesigned nozzle, streamlined manufacturing

  10. Development of improved thermal hydraulics and fuel performance technology; development of turbulence model for flow analysis in nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Joo, W. K.; Kong, D. W.; Park, H. Z. [Yonsei University, Seoul (Korea)

    2001-04-01

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Subchannel flow in a nuclear bundle having vanes to mix flow appears complex turbulent flow. Objective of this study is to develop turbulence model which can predict complex flow. Also, the module will be produced, which can implement the developed turbulence model in the CFX code. The selected turbulence models are k-epsilon model, non-linear k-epsilon model, Reynolds stress model and modified Reynolds stress model to test their performance in the prediction of the flow in nuclear assembly. These models are tested for a 2-D backwise step flow, square duct flow, rod bundle flow and subchannel flow using CFX. The modules, which can implement Reynolds stress model and non-linear k-epsilon odel in CFX code, are produced. The advantages and disadvantages for these turbulence models are described and the limitation of implementation of non-linear model in CFX code is discussed. The results obtained from the research would give a help for the development of turbulence model which can accurately predict the flow through the rod bundles with mixing vanes. 18 refs., 37 figs., 8 tabs. (Author)

  11. Evaluation of the thermal-mechanic performance of fuel rods MOX in fuel assemblies 10 x 10; Evaluacion del desempeno termo-mecanico barras combustibles MOX en ensambles combustible 10 x 10

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H., E-mail: hector.hernandez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    In the Instituto Nacional de Investigaciones Nucleares (Mexico) , we have been working in proposals of fuel assemblies that bear to the reduction of the plutonium inventories that exist a global level, plutonium coming from the dismantlement of the nuclear weapons as of the one used as fuel inside the reactors in operation at the present time. For this reason besides carrying out the evaluation of the neutron performance is necessary to realize the evaluation of the thermal-mechanic behavior of the rods that compose a fuel assembly with the purpose of determining if under the operation conditions to those that are subjected the fuel does not surpass the limit established and this causes a failure in the fuel element. In this sense when carrying out the analysis of an fuel element of mixed oxides in an arrangement 10 x 10 is observed that under the established operation conditions for the proposed cycle values that surpass the limit established for fuel failure are not presented, therefore the proposed assembly can be used as reload element in the nuclear power plant of Laguna Verde. (Author)

  12. Verifying nuclear fuel assemblies in wet storages on a partial defect level: A software simulation tool for evaluating the capabilities of the Digital Cherenkov Viewing Device

    Energy Technology Data Exchange (ETDEWEB)

    Grape, Sophie, E-mail: sophie.grape@physics.uu.se [Department of Physics and Astronomy, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Jacobsson Svärd, Staffan [Department of Physics and Astronomy, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Lindberg, Bo [Lens-Tech AB, Box 733, SE-93127 Skellefteå (Sweden)

    2013-01-11

    The Digital Cherenkov Viewing Device (DCVD) is an instrument that records the Cherenkov light emitted from irradiated nuclear fuels in wet storages. The presence, intensity and pattern of the Cherenkov light can be used by the International Atomic Energy Agency (IAEA) inspectors to verify that the fuel properties comply with declarations. The DCVD is since several years approved by the IAEA for gross defect verification, i.e. to control whether an item in a storage pool is a nuclear fuel assembly or a non-fuel item [1]. Recently, it has also been endorsed as a tool for partial defect verification, i.e. to identify if a fraction of the fuel rods in an assembly have been removed or replaced. The latter recognition was based on investigations of experimental studies on authentic fuel assemblies and of simulation studies on hypothetic cases of partial defects [2]. This paper describes the simulation methodology and software which was used in the partial defect capability evaluations. The developed simulation procedure uses three stand-alone software packages: the ORIGEN-ARP code [3] used to obtain the gamma-ray spectrum from the fission products in the fuel, the Monte Carlo toolkit Geant4 [4] for simulating the gamma-ray transport in and around the fuel and the emission of Cherenkov light, and the ray-tracing programme Zemax [5] used to model the light transport through the assembly geometry to the DCVD and to mimic the behaviour of its lens system. Furthermore, the software allows for detailed information from the plant operator on power and/or burnup distributions to be taken into account to enhance the authenticity of the simulated images. To demonstrate the results of the combined software packages, simulated and measured DCVD images are presented. A short discussion on the usefulness of the simulation tool is also included.

  13. Verifying nuclear fuel assemblies in wet storages on a partial defect level: A software simulation tool for evaluating the capabilities of the Digital Cherenkov Viewing Device

    Science.gov (United States)

    Grape, Sophie; Jacobsson Svärd, Staffan; Lindberg, Bo

    2013-01-01

    The Digital Cherenkov Viewing Device (DCVD) is an instrument that records the Cherenkov light emitted from irradiated nuclear fuels in wet storages. The presence, intensity and pattern of the Cherenkov light can be used by the International Atomic Energy Agency (IAEA) inspectors to verify that the fuel properties comply with declarations. The DCVD is since several years approved by the IAEA for gross defect verification, i.e. to control whether an item in a storage pool is a nuclear fuel assembly or a non-fuel item [1]. Recently, it has also been endorsed as a tool for partial defect verification, i.e. to identify if a fraction of the fuel rods in an assembly have been removed or replaced. The latter recognition was based on investigations of experimental studies on authentic fuel assemblies and of simulation studies on hypothetic cases of partial defects [2]. This paper describes the simulation methodology and software which was used in the partial defect capability evaluations. The developed simulation procedure uses three stand-alone software packages: the ORIGEN-ARP code [3] used to obtain the gamma-ray spectrum from the fission products in the fuel, the Monte Carlo toolkit Geant4 [4] for simulating the gamma-ray transport in and around the fuel and the emission of Cherenkov light, and the ray-tracing programme Zemax [5] used to model the light transport through the assembly geometry to the DCVD and to mimic the behaviour of its lens system. Furthermore, the software allows for detailed information from the plant operator on power and/or burnup distributions to be taken into account to enhance the authenticity of the simulated images. To demonstrate the results of the combined software packages, simulated and measured DCVD images are presented. A short discussion on the usefulness of the simulation tool is also included.

  14. 混合堆增殖钍基燃料组件中子学分析%Neutronics Calculation of Fusion-Fission Hybrid Breeding Thorium Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    马续波; 陈义学; 全国萍; 王悦; 韩静茹; 陆道纲

    2012-01-01

    A preliminary comparative study of the physical properties among 17×17 fuel assembly in PWRs for prototype between uranium assembly and hybrid breeding thorium-based assembly has been investigated respectively using the DRAGON software. The parameters such as fuel temperature coefficient, moderator temperature coefficient and that variation as a function of operation period have been investigated. Results show that the neutron properties of uranium-based assembly and hybrid breeding thorium-based assembly are similitude, but MA mass of hybrid breeding thorium-based assembly is evidently less than those of the uranium assembly.%采用压水堆17×17燃料组件模型,用燃料组件参数计算程序DRAGON分别对混合堆增殖钍燃料组件和全铀组件的中子学特性进行了研究,分析组件的燃料温度系数、慢化剂温度系数及其与燃耗的关系.计算结果表明,混合堆增殖钍燃料组件和全铀组件的中子特性相似,但钍燃料组件中的乏燃料组件中的次锕系核素(MA)的含量明显减少.

  15. Fire safety assessment for a typical hot cell handling failed fuel sub-assembly. Contributed Paper MS-03

    International Nuclear Information System (INIS)

    This paper presents a systematic study of fire hazard potential within a typical hot cell that handles Failed Fuel SubAssemblies (FSA) for cleaning purposes. A hot cell configuration is considered wherein ethyl alcohol is used as the cleaning agent. The potential for generation of ethyl alcohol vapors due to heat load of FSA, hydrogen generation during the cleaning process, possibility of vapour ignition and sustainability of fire within the cell are discussed. Detailed heat transfer and CFD studies were performed using computational tools developed in-house at SRI to address these issues. Based on this, several recommendations and suggestions are provided for safe operating conditions that could preclude the occurrence of fire within the hot cell. (author)

  16. Rational design of lower-temperature solid oxide fuel cell cathodes via nanotailoring of co-assembled composite structures.

    Science.gov (United States)

    Lee, Kang Taek; Lidie, Ashley A; Yoon, Hee Sung; Wachsman, Eric D

    2014-12-01

    A novel in situ co-assembled nanocomposite LSM-Bi1.6 Er0.4 O3 (ESB) (icn-LSMESB) was obtained by conjugated wet-chemical synthesis. It showed an enhancement of the cathode polarization at 600 °C by >140 times relative to conventional LSM-Y0.08 Zr0.84 O1.92 (YSZ) cathodes and exceptional solid oxide fuel cell (SOFC) performance of >2 W cm(-2) below 750 °C. This demonstrates that this novel cost-effective and broadly applicable process provides new opportunities for performance enhancement of energy storage and conversion devices by nanotailoring of composite electrodes. PMID:25287642

  17. Facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Patent for facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies, which are arranged within a building in a horizontal position and are cooled by a gas stream, whereby the building has a storage and a loading zone, characterized by the fact that pallet trucks arranged one above the other in a row and such that an interspace is left for the receiving positions for the containers, the the pallet trucks can be moved along rails that extend between two side walls arranged opposite to one another in the storage zone, that the storage zone can be loaded and unloaded by opening located in these two side walls, and that the gas stream only circulates within the building

  18. The Design Summary of Research Reactor Fuel Assembly%研究堆燃料组件设计综述

    Institute of Scientific and Technical Information of China (English)

    雷涛; 粟敏; 黄春兰

    2014-01-01

    研究堆是核反应堆的一种类型,其主要功能是为研究或其它用途提供中子源,是一种工具堆。燃料组件是研究堆中的重要部件,由于其用途与商用堆存在较大的不同,因此其燃料组件在结构设计上与商用堆组件存在较大差异。本文从燃料组件的整体结构、连接结构以及流道结构等方面对研究堆燃料组件结构设计进行了分析。在此基础上,提出了研究堆燃料组件设计方面的建议,以供类似组件设计参考。%Research reactor is one type of multitudinous nuclear reactors. It is mainly used to research or provide neutrons for others, and is a tool reactor. Fuel assembly is an important component of research reactor, the structure of which is quite different from the one of commercial reactor because of their different uses. The whole structure, the connection and the flow channel of the research reactor are analyzed in this paper. Based on this, the fuel assembly design of the research reactor is proposed in this paper, and it has some reference value for other design.

  19. 78 FR 55648 - Signal Booster Rules

    Science.gov (United States)

    2013-09-11

    ...), published at 78 FR 21555, April 11, 2013, are effective September 11, 2013. FOR FURTHER INFORMATION CONTACT... protect Customer Proprietary Network Information (CPNI). Privacy Act: The information collection... facilitate review of wireless providers' behavior regarding Consumer Signal Boosters, the R&O requires...

  20. Sobol's sensitivity analysis for a fuel cell stack assembly model with the aid of structure-selection techniques

    Science.gov (United States)

    Zhang, Wei; Cho, Chongdu; Piao, Changhao; Choi, Hojoon

    2016-01-01

    This paper presents a novel method for identifying the main parameters affecting the stress distribution of the components used in assembly modeling of proton exchange membrane fuel cell (PEMFC) stack. This method is a combination of an approximation model and Sobol's method, which allows a fast global sensitivity analysis for a set of uncertain parameters using only a limited number of calculations. Seven major parameters, i.e., Young's modulus of the end plate and the membrane electrode assembly (MEA), the contact stiffness between the MEA and bipolar plate (BPP), the X and Y positions of the bolts, the pressure of each bolt, and the thickness of the end plate, are investigated regarding their effect on four metrics, i.e., the maximum stresses of the MEA, BPP, and end plate, and the stress distribution percentage of the MEA. The analysis reveals the individual effects of each parameter and its interactions with the other parameters. The results show that the X position of a bolt has a major influence on the maximum stresses of the BPP and end plate, whereas the thickness of the end plate has the strongest effect on both the maximum stress and the stress distribution percentage of the MEA.

  1. Characterization of spent fuel assemblies for storage facilities using non destructive assay

    International Nuclear Information System (INIS)

    Many non destructive assay (NDA) techniques have been developed by the French Atomic Energy Commission (CEA) for spent fuel characterization and management. Passive and active neutron methods as well as gamma spectrometric methods have been carried out and applied to industrial devices like PYTHONTM and NAJA. Many existing NDA methods can be successfully applied to storage, but the most promising are the neutron methods combined with on line evolution codes. For dry storage applications, active neutron measurements require further R and D to achieve accurate results. Characterization data given by NDA instruments can now be linked to automatic fuel recognition. Both information can feed the storage management software in order to meet the storage operation requirements like: fissile mass inventory, operators declaration consistency or automatic selection of proper storage conditions. (author)

  2. Device for facilitating the insertion and withdrawal of fuel assemblies from a nuclear reactor

    International Nuclear Information System (INIS)

    A device is provided which is installed in a reactor prior to carrying out refueling operations and which accurately locates and isolates a selected core location to permit rapid withdrawal and insertion of fuel subassemblies at that location. A shielded plug designed to cooperate with the refueling apparatus is inserted into an access port in the reactor head. A structural shroud extends down from the plug and carries at its lower end a radially floating, hexagonal spreader tube with mechanisms to rotate it for angular alignment purposes and a linear drive for inserting it into the core. The upper end of the spreader tube serves as a guide for leading the fuel handling apparatus into alignment with the chosen subassembly

  3. NASA's Space Launch System: Developing the World's Most Powerful Solid Booster

    Science.gov (United States)

    Priskos, Alex

    2016-01-01

    NASA's Journey to Mars has begun. Indicative of that challenge, this will be a multi-decadal effort requiring the development of technology, operational capability, and experience. The first steps are under way with more than 15 years of continuous human operations aboard the International Space Station (ISS) and development of commercial cargo and crew transportation capabilities. NASA is making progress on the transportation required for deep space exploration - the Orion crew spacecraft and the Space Launch System (SLS) heavy-lift rocket that will launch Orion and large components such as in-space stages, habitat modules, landers, and other hardware necessary for deep-space operations. SLS is a key enabling capability and is designed to evolve with mission requirements. The initial configuration of SLS - Block 1 - will be capable of launching more than 70 metric tons (t) of payload into low Earth orbit, greater mass than any other launch vehicle in existence. By enhancing the propulsion elements and larger payload fairings, future SLS variants will launch 130 t into space, an unprecedented capability that simplifies hardware design and in-space operations, reduces travel times, and enhances the odds of mission success. SLS will be powered by four liquid fuel RS-25 engines and two solid propellant five-segment boosters, both based on space shuttle technologies. This paper will focus on development of the booster, which will provide more than 75 percent of total vehicle thrust at liftoff. Each booster is more than 17 stories tall, 3.6 meters (m) in diameter and weighs 725,000 kilograms (kg). While the SLS booster appears similar to the shuttle booster, it incorporates several changes. The additional propellant segment provides additional booster performance. Parachutes and other hardware associated with recovery operations have been deleted and the booster designated as expendable for affordability reasons. The new motor incorporates new avionics, new propellant

  4. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design

    International Nuclear Information System (INIS)

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs

  5. End-to-end calculation of the radiation characteristics of VVER-1000 spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Linge, I. I.; Mitenkova, E. F., E-mail: mit@ibrae.ac.ru; Novikov, N. V. [Russian Academy of Sciences, Nuclear Safety Institute (Russian Federation)

    2012-12-15

    The results of end-to-end calculation of the radiation characteristics of VVER-1000 spent nuclear fuel are presented. Details of formation of neutron and gamma-radiation sources are analyzed. Distributed sources of different types of radiation are considered. A comparative analysis of calculated radiation characteristics is performed with the use of nuclear data from different ENDF/B and EAF files and ANSI/ANS and ICRP standards.

  6. End-to-end calculation of the radiation characteristics of VVER-1000 spent fuel assemblies

    Science.gov (United States)

    Linge, I. I.; Mitenkova, E. F.; Novikov, N. V.

    2012-12-01

    The results of end-to-end calculation of the radiation characteristics of VVER-1000 spent nuclear fuel are presented. Details of formation of neutron and gamma-radiation sources are analyzed. Distributed sources of different types of radiation are considered. A comparative analysis of calculated radiation characteristics is performed with the use of nuclear data from different ENDF/B and EAF files and ANSI/ANS and ICRP standards.

  7. Fission Thrust sail as booster for high {\\Delta}v fusion based propulsion

    OpenAIRE

    Ceyssens, Frederik; Wouters, Kristof; Driesen, Maarten

    2014-01-01

    The fission thrust sail as booster for nuclear fusion-based rocket propulsion for future starships is studied. Some required aspects of these systems such as neutron moderation and sail regeneration are discussed. First order calculations are used together with Monte Carlo simulations to assess system performance. When the fusion rocket has relatively low efficiency (~30%) in converting fusion fuel to a directed exhaust, adding a fission sail is shown to be beneficial for obtainable delta-v. ...

  8. Model-Based Control of a Continuous Coating Line for Proton Exchange Membrane Fuel Cell Electrode Assembly

    Directory of Open Access Journals (Sweden)

    Vikram Devaraj

    2015-01-01

    Full Text Available The most expensive component of a fuel cell is the membrane electrode assembly (MEA, which consists of an ionomer membrane coated with catalyst material. Best-performing MEAs are currently fabricated by depositing and drying liquid catalyst ink on the membrane; however, this process is limited to individual preparation by hand due to the membrane’s rapid water absorption that leads to shape deformation and coating defects. A continuous coating line can reduce the cost and time needed to fabricate the MEA, incentivizing the commercialization and widespread adoption of fuel cells. A pilot-scale membrane coating line was designed for such a task and is described in this paper. Accurate process control is necessary to prevent manufacturing defects from occurring in the coating line. A linear-quadratic-Gaussian (LQG controller was developed based on a physics-based model of the coating process to optimally control the temperature and humidity of the drying zones. The process controller was implemented in the pilot-scale coating line proving effective in preventing defects.

  9. The problems of mass transfer and formation of deposits of corrosion products on fuel assemblies of a VVER-1200 reactor

    Science.gov (United States)

    Rodionov, Yu. A.; Kritskii, V. G.; Berezina, I. G.; Gavrilov, A. V.

    2014-03-01

    On the basis of examination of materials published both in Russia and abroad, as well as their own investigations, the authors explain the reasons for the occurrence of such effects as AOA (Axial Offset Anomalies) and an increase in the coolant pressure difference in the core of nuclear reactors of the VVER type. To detect the occurrence of the AOA effect, the authors suggest using the specific activity of 58Co in the coolant. In the VVER-1200 design the thermohydraulic regime for fuel assemblies in the first year of their service life involves slight boiling of the coolant in the upper part of the core, which may induce the occurrence of the AOA effect, intensification of corrosion of fuel claddings, and abnormal increase in deposition of corrosion products. Radiolysis of the water coolant in the boiling section (boiling in pores of deposits) may intensify not only general corrosion but also a localized (nodular) one. As a result of intensification of the corrosion processes and growth of deposits, deterioration of the radiation situation in the rooms of the primary circuit of a VVER-1200 reactor as compared to that at nuclear power plants equipped with reactors of the VVER-1000 type is possible. Recommendations for preventing the AOA effect at nuclear power plants with VVER-1200 reactors on the matter of the direction of further investigations are made.

  10. Development of self-interrogation neutron resonance densitometry (SINRD) to measure U-235 and Pu-239 content in a PWR spent fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Lafleur, Adrienne M [Los Alamos National Laboratory; Charlton, William S [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory

    2009-01-01

    The use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U and {sup 239}Pu content in a PWR spent fuel assembly was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in fission chamber. These simulations utilize the {sup 244}Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be measured using {sup 235}U and {sup 239}Pu fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the sensitivity of the measurements to extraneous material present in fuel. The development of SINRD to measure the fissile content in spent fuel is of great importance to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in FBR spent fuel and heavy metal product from reprocessing methods.

  11. CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

    2011-07-01

    been developed for calculation and synchronization purposes. The data exchange is realized by means of the Parallel Virtual Machine (PVM) software package. In this contribution, steady-state and transient results of a quarter of PWR fuel assembly with cold water injection are presented and compared with obtained results from a RELAP5/PARCS v2.7 coupled calculation. A simplified model for the spacers has been included. A methodology has been introduced to take into account the pressure drop and the turbulence enhancement produced by the spacers. (author)

  12. Pressure-Equalizing Cradle for Booster Rocket Mounting

    Science.gov (United States)

    Rutan, Elbert L. (Inventor)

    2015-01-01

    A launch system and method improve the launch efficiency of a booster rocket and payload. A launch aircraft atop which the booster rocket is mounted in a cradle, is flown or towed to an elevation at which the booster rocket is released. The cradle provides for reduced structural requirements for the booster rocket by including a compressible layer, that may be provided by a plurality of gas or liquid-filled flexible chambers. The compressible layer contacts the booster rocket along most of the length of the booster rocket to distribute applied pressure, nearly eliminating bending loads. Distributing the pressure eliminates point loading conditions and bending moments that would otherwise be generated in the booster rocket structure during carrying. The chambers may be balloons distributed in rows and columns within the cradle or cylindrical chambers extending along a length of the cradle. The cradle may include a manifold communicating gas between chambers.

  13. Optimization of enrichment distributions in nuclear fuel assemblies loaded with Uranium and Plutonium via a modified linear programming technique

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas Vivas, Gabriel Francisco

    1999-12-01

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed Oxide (MOX) isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide range of applicability of the optimization technique. The feature of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactor (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or in the

  14. Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

    Energy Technology Data Exchange (ETDEWEB)

    Bardet, Philippe [George Washington Univ., Washington, DC (United States); Ricciardi, Guillaume [Atomic Energy Commission (CEA) (France)

    2016-01-31

    Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWR fuel bundle behavior during seismic transients.

  15. Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

    International Nuclear Information System (INIS)

    Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging task in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWR fuel bundle behavior during seismic transients.

  16. Operational behaviour of CO{sub 2} booster systems; Betriebsverhalten von CO{sub 2}-Booster-Systemen

    Energy Technology Data Exchange (ETDEWEB)

    Javerschek, Oliver; Hieble, Tobias [BITZER Kuehlmaschinenbau GmbH, Sindelfingen (Germany)

    2011-07-01

    The operating characteristics of booster systems and the resulting operating conditions of CO{sub 2} booster systems in supermarket refrigeration are explained and discussed. Criteria and challenges of different operating and load conditions are gone into. Simulated and measured operating states of a small-scale booster system are compared and evaluated. [German] In der vorliegenden Veroeffentlichung werden unterschiedliche Betriebsverhalten und die daraus resultierenden Betriebsbedingungen von CO{sub 2}-Booster-Systemen in der Supermarktkaelte erlaeutert und diskutiert. Dabei werden wesentliche Kriterien und Herausforderungen bei den unterschiedlichen Betriebs- und Lastbedingungen besprochen. Ausserdem werden simulierte und gemessene Betriebszustaende einer kleinen Booster-Kaelteanlage vergleichend betrachtet und bewertet.

  17. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel