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Sample records for bundles fuel elements

  1. CHF Enhancement of Advanced 37-Element Fuel Bundles

    Directory of Open Access Journals (Sweden)

    Joo Hwan Park

    2015-01-01

    Full Text Available A standard 37-element fuel bundle (37S fuel bundle has been used in commercial CANDU reactors for over 40 years as a reference fuel bundle. Most CHF of a 37S fuel bundle have occurred at the elements arranged in the inner pitch circle for high flows and at the elements arranged in the outer pitch circle for low flows. It should be noted that a 37S fuel bundle has a relatively small flow area and high flow resistance at the peripheral subchannels of its center element compared to the other subchannels. The configuration of a fuel bundle is one of the important factors affecting the local CHF occurrence. Considering the CHF characteristics of a 37S fuel bundle in terms of CHF enhancement, there can be two approaches to enlarge the flow areas of the peripheral subchannels of a center element in order to enhance CHF of a 37S fuel bundle. To increase the center subchannel areas, one approach is the reduction of the diameter of a center element, and the other is an increase of the inner pitch circle. The former can increase the total flow area of a fuel bundle and redistributes the power density of all fuel elements as well as the CHF. On the other hand, the latter can reduce the gap between the elements located in the middle and inner pitch circles owing to the increasing inner pitch circle. This can also affect the enthalpy redistribution of the fuel bundle and finally enhance CHF or dry-out power. In this study, the above two approaches, which are proposed to enlarge the flow areas of the center subchannels, were considered to investigate the impact of the flow area changes of the center subchannels on the CHF enhancement as well as the thermal characteristics by applying a subchannel analysis method.

  2. Study on Unigraphics Drawing Modeling Method for 37-Element and CANFLEX Fuel Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Yu Mi; Park, Joo Hwan

    2010-03-15

    The CANFLEX bundle contains 43 elements of two different diameters. It has two rings of small diameter elements on the outside, and eight elements (with diameter slightly larger than those in the standard 37-Element bundle) in the center. This larger number of small diameter elements on the outside of the CANFLEX bundle enhances thermo-hydraulic capability, resulting in a higher power capability and an improvement in operating safety margins. As a Result of advanced fuel design for CANFLEX fuel bundles, components consisting of fuel bundles are more complicated. Hence, the detailed modeling of components is inevitable in order to analyze the fuel performance by computational fluid dynamics. In this report, the basic design of the advanced fuel for CANDU reactors was carried out and the methodology for the modeling of fuel bundle were described. Firstly, the components consisting of fuel bundles were separately modeled and saved with different file names. The final feature of fuel bundle was accomplished by an assembling process of components. Since this report developed the modeling methodology based on the Unigraphics program, the basic explanations for the software were given first, and the complete modeling of 37-elements and CANFLEX fuel bundles were provided. The components of CANFLEX fuel bundles were also compared with that of 37-elements fuel bundles. Although, in this report, the modeling methodology is applied only to 37-elements and CANFLEX fuel bundles, this methodology may be applicable to the newly designed fuel bundles which are to be developed in the future

  3. Subchannel analysis of CANDU 37-element fuel bundles

    International Nuclear Information System (INIS)

    The subchannel analysis codes COBRA-IV and ASSERT-4 have been used to predict the mass and enthalpy imbalance within a CANDU 37-element fuel channel under various system conditions. The objective of this study was to assess the various capabilities of the ASSERT code and highlight areas where further validation or development may be needed. The investigation indicated that the ASSERT code has all the basic models required to accurately predict the flow and enthalpy imbalance for complex rod bundles. The study also showed that the code modelling of void drift and diffusion requires refinement to some coefficients and that further validation is needed at high flow rate and high void fraction conditions, where ASSERT and COBRA are shown to predict significantly different trends. The results of a recent refinement of ASSERT modelling are also discussed

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

    International Nuclear Information System (INIS)

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

  5. Automation in inspection of PHWR fuel elements & bundles at Nuclear Fuel Complex

    International Nuclear Information System (INIS)

    Nuclear Fuel Complex (NFC), Hyderabad, a constituent of Department of Atomic Energy, India manufactures fuel for all Indian nuclear power reactors. Currently NFC manufactures both 19 element & 37 element bundles for catering to the requirement of 220 MWe & 540 MWe PHWRs. In order to meet the growing needs for the Nuclear Fuel, NFC engaged in expansion of the production facilities. This calls for enhanced throughput at various inspection stages keeping in tandem with the production & for achieving this objective, NFC has chosen automation. This paper deals with automation of the inspection line at NFC. (author)

  6. Manufacturing of 37-element fuel bundles for PHWR 540 - new approach

    Energy Technology Data Exchange (ETDEWEB)

    Arora, U.K.; Sastry, V.S.; Banerjee, P.K.; Rao, G.V.S.H.; Jayaraj, R.N. [Nuclear Fuel Complex, Dept. Atomic Energy, Government of India, Hyderabad (India)

    2003-07-01

    Nuclear Fuel Complex (NFC), established in early seventies, is a major industrial unit of Department of Atomic Energy. NFC is responsible for the supply of fuel bundles to all the 220 MWe PHWRs presently in operation. For supplying fuel bundles for the forthcoming 540 MWe PHWRs, NEC is dovetailing 37-element fuel bundle manufacturing facilities in the existing plants. In tune with the philosophy of self-reliance, emphasis is given to technology upgradation, higher customer satisfaction and application of modern quality control techniques. With the experience gained over the years in manufacturing 19-element fuel bundles, NEC has introduced resistance welding of appendages on fuel tubes prior to loading of UO{sub 2} pellets, use of bio-degradable cleaning agents, simple diagnostic tools for checking the equipment condition, on line monitoring of variables, built-in process control methods and total productive maintenance concepts in the new manufacturing facility. Simple material handling systems have been contemplated for handling of the fuel bundles. This paper highlights the flow-sheet adopted for the process, design features of critical equipment and the methodology for fabricating the 37-element fuel bundles, 'RIGHT FIRST TIME'. (author)

  7. Manufacturing of 37-element fuel bundles for PHWR 540 - new approach

    International Nuclear Information System (INIS)

    Nuclear Fuel Complex (NFC), established in early seventies, is a major industrial unit of Department of Atomic Energy. NFC is responsible for the supply of fuel bundles to all the 220 MWe PHWRs presently in operation. For supplying fuel bundles for the forthcoming 540 MWe PHWRs, NEC is dovetailing 37-element fuel bundle manufacturing facilities in the existing plants. In tune with the philosophy of self-reliance, emphasis is given to technology upgradation, higher customer satisfaction and application of modern quality control techniques. With the experience gained over the years in manufacturing 19-element fuel bundles, NEC has introduced resistance welding of appendages on fuel tubes prior to loading of UO2 pellets, use of bio-degradable cleaning agents, simple diagnostic tools for checking the equipment condition, on line monitoring of variables, built-in process control methods and total productive maintenance concepts in the new manufacturing facility. Simple material handling systems have been contemplated for handling of the fuel bundles. This paper highlights the flow-sheet adopted for the process, design features of critical equipment and the methodology for fabricating the 37-element fuel bundles, 'RIGHT FIRST TIME'. (author)

  8. Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

    Directory of Open Access Journals (Sweden)

    Park Joo Hwan

    2016-01-01

    Full Text Available A CANDU-6 reactor, which has 380 fuel channels of a pressure tube type, is suffering from aging or creep of the pressure tubes. Most of the aging effects for the CANDU primary heat transport system were originated from the horizontal crept pressure tubes. As the operating years of a CANDU reactor proceed, a pressure tube experiences high neutron irradiation damage under high temperature and pressure. The crept pressure tube can deteriorate the Critical Heat Flux (CHF of a fuel channel and finally worsen the reactor operating performance and thermal margin. Recently, the modification of the central subchannel area with increasing inner pitch length of a standard 37-element fuel bundle was proposed and studied in terms of the dryout power enhancement for the uncrept pressure tube since a standard 37-element fuel bundle has a relatively small flow area and high flow resistance at the central region. This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition, the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes. Also, it was discussed how much the crept pressure tubes affected the thermalhydraulic characteristics of the fuel channel as well as the dryout power for the modification of a standard 37-element fuel bundle.

  9. Parametric study of thermo-mechanical behaviour of 19-element PHWR fuel bundle having AHWR fuel material

    International Nuclear Information System (INIS)

    AHWR Th-LEU of 4.3 weight % 235U enrichment is a fuel design option for its trial irradiation in Indian PHWRs. The important component of this option is the large enhancement in the average discharge burn-up from the core. A parametric study of the 19-element fuel bundle, with natural uranium currently is being used in all operating 220 MWe PHWRs, has been carried out for AHWR Th-LEU fuel material by computer code FUDA MOD2. The important fuel parameters such as fuel temperature, fission gas release, fuel swelling and sheath strain have been analyzed for required fuel performance. With Th-LEU, average discharge burnups of about 25,000 MW-d/TeHE can be achieved. The FUDA code (Fuel Design Analysis code) MOD2 version has been used in the fuel element analysis. The code takes into account the inter-dependence of different parameters like fuel pellet temperatures, pellet expansions, fuel-sheath gap heat transfer, sheath strain and stresses, fission gas release and gas pressures, fuel densification etc. Thermo-mechanical analysis of fuel element having AHWR material is carried out for the bundle power histories reaching up to design burn-up 40000 MWd/TeHE. The resultant parameters such as fuel temperature, sheath plastic strain and fission gas pressure for AHWR fuel element were compared with respective thermo-mechanical parameters for similar fuel bundle element with natural uranium as fuel material. (author)

  10. Optimization of thorium-uranium content in a 54-element fuel bundle for use in a CANDU-SCWR

    International Nuclear Information System (INIS)

    A new 54-element fuel bundle design has been proposed for use in a pressure-tube supercritical water-cooled reactor, a pre-conceptual evolution of existing CANDU reactors. Pursuant to the goals of the Generation IV International Forum regarding advancement in nuclear fuel cycles, optimization of the thorium and uranium content in each ring of fuel elements has been studied with the objectives of maximizing the achievable fuel utilization (burnup) and total thorium content within the bundle, while simultaneously minimizing the linear element ratings and coolant void reactivity. The bundle was modeled within a reactor lattice cell using WIMS-AECL, and the uranium and thorium content in each ring of fuel elements was optimized using a weighted merit function of the aforementioned criteria and a metaheuristic search algorithm. (author)

  11. An experimental investigation of the temperature behavior of a CANDU 37-element spent fuel bundle with air backfill

    International Nuclear Information System (INIS)

    As part of the thermal analysis of a CANDU spent fuel dry storage system, a series of experiment has been conducted using a thermal mock-up of a simulated CANDU spent fuel bundle in a dry storage basket. The experimental system was designed to obtain the maximum fuel rod temperature along with the radial and axial temperature distributions within the fuel bundle. The main purpose of these experiments was to characterize the relevant heat transfer mechanisms in a dry, vertically oriented CANDU spent fuel bundle, and to verify the MAXROT code developed for the thermal analysis of a CANDU spent fuel bundle in a dry storage basket. A total of 48 runs were made with 8 different power inputs to the 37-element heater rod bundle ranging from 5 to 40 W, while using 6 different band heaters power inputs from 0 to 250 W to maintain the basket wall at a desired boundary condition temperature at the steady state. The temperature distribution in a heater rod bundle was measured and recorded at the saturated condition for each set of heater rod power and band heaters power. To characterize the heat transfer mechanism involved, the experimental data were corrected analytically for radiation heat transfer and presented as a Nusselt number correlation in terms of the Rayleigh number of the heater rod bundle. The results show that the Nusselt number remains nearly constant and all the experimental dada fall within a conduction regime. The experimental data were compared with the predictions of the MAXROT code to examine the code's accuracy and validity of assumptions used in the code. The MAXROT code explicitly models each representative fuel rod in a CANDU fuel bundle and couples the conductive and radiative heat transfer of the internal gas between rods. Comparisons between the measured and predicted maximum fuel rod temperatures of the simulated CANDU 37-element spent fuel bundle for all 48 tests show that the MAXROT code slightly over-predicts and the agreement is within 2

  12. Behavior of mixed-oxide fuel elements in a tight bundle under duty-cycle conditions

    International Nuclear Information System (INIS)

    The irradiation behavior of the TOB-10 fuel pins was comparable with that obtained in the single pin tests. There was no significant effect that could be directly attributed to tight bundle configuration. The postirradiation examination data provided information on the axial migration of cesium and its effect on cladding strain. Severe fuel/cladding chemical interaction (FCCI), which resulted in substantial cladding thinning and probably restricted venting of fission gas from the fuel column into the pin plena, apparently caused the earlier-than-expected cladding breaches in the D9-clad pins. No such severe FCCI was noted in the 316SS-clad pins. At the time of test termination, the overall cladding strain from creep and swelling was insufficient to cause bundle closure. Consequently, there would have been minimal pin bundle-duct interaction in the subassembly. Neither of the breaches appeared to be induced by pin bundle-duct interaction. (author)

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

    International Nuclear Information System (INIS)

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

  14. Investigation of coolant thermal mixing within 28-element CANDU fuel bundles using the ASSERT-PV thermal hydraulics code

    International Nuclear Information System (INIS)

    This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)

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

    International Nuclear Information System (INIS)

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

  16. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

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

  17. Static stress analysis of CANFLEX fuel bundles

    International Nuclear Information System (INIS)

    The static stress analysis of CANFLEX bundles is performed to evaluate the fuel structural integrity during the refuelling service. The structure analysis is carried out by predicting the drag force, stress and displacements of the fuel bundle. By the comparison of strength tests and analysis results, the displacement values are well agreed within 15%. The analysis shows that the CANFLEX fuel bundle keep its structural integrity. 24 figs., 6 tabs., 12 refs. (Author) .new

  18. Hydraulic characteristics of HANARO fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S.; Chung, H. J.; Chun, S. Y.; Yang, S. K.; Chung, M. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)

  19. Post-irradiation examination of the 37M fuel bundle at Chalk River Laboratories (AECL)

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Daniels, T. [Ontario Power Generation, Pickering, Ontario (Canada); Montin, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-03-15

    The modified (-element (37M) fuel bundle was designed by Ontario Power Generation (OPG) to improve Critical Heat Flux (CHF) performance in ageing pressure tubes. A modification of the conventional 37-element fuel bundle design, the 37M fuel bundle allows more coolant flow through the interior sub-channels by way of a smaller central element. A demonstration irradiation (DI) of thirty-two fuel bundles was completed in 2011 at OPG's Darlington Nuclear Generating Station to confirm the suitability of the 37M fuel bundles for full core implementation. In support of the DI, fuel elements were examined in the Chalk River Laboratories Hot Cells. Inspection activities included: Bundle and element visual examination; Bundle and element dimensional measurements; Verification of bundle and element integrity; and Internal Gas Volume Measurements. The inspection results for 37M were comparable to that of conventional 37-element CANDU fuel. Fuel performance parameters of the 37M DI fuel bundle and fuel elements were within the range observed for similarly operated conventional 37-element CANDU fuel. Based on these Post Irradiation Examination (PIE) results, 37M fuel performed satisfactorily. (author)

  20. LVRF fuel bundle manufacture for Bruce

    Energy Technology Data Exchange (ETDEWEB)

    Pant, A. [Zircatec Precision Industries, Port Hope, Ontario (Canada)

    2005-12-15

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element Bruce LVRF bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37-element fuel over 20 years ago. Introduction of this new line has involved the introduction of significant changes to an environment that is not used to rapid changes with significant impact. At ZPI we have been able to build on our innovative capabilities in new fuel manufacturing, the strength and experience of our core team, and on our prevailing management philosophy of 'support the doer'. The presentation will discuss some of the novel aspects of this fuel introduction and the mix of innovative and classical project management methods that are being used to ensure that project deliveries are being met. Supporting presentations will highlight some of the issues in more detail. (author)

  1. In-pile test of Qinshan PWR fuel bundle

    International Nuclear Information System (INIS)

    In-pile test of Qinshan Nuclear Power Plant PWR fuel bundle has been conducted in HWRR HTHP Test loop at CIAE. The test fuel bundle was irradiated to an average burnup of 25000 Mwd/tU. The authors describe the structure of (3 x 3-2) test fuel bundle, structure of irradiation rig, fuel fabrication, irradiation conditions, power and fuel burnup. Some comments on the in-pile performance for fuel bundle, fuel rod and irradiation rig were made

  2. TRIGA spent fuel bundles safe storage

    International Nuclear Information System (INIS)

    TRIGA-SSR is a steady state research and material test reactor that has been in operation since 1980. The original TRIGA fuel was HEU (highly enriched uranium) with a U235 enrichment of 93 per cent. Almost all TRIGA HEU fuel bundles are now burned-up. Part of the spent fuel was loaded and transferred to US, in a Romania - DOE arrangement. The rest of the TRIGA fuel bundles have to be temporarily stored in the TRIGA facility. As the storage conditions had to be established with caution, neutron and thermal hydraulic evaluations of the storage conditions were required. Some criticality evaluations were made based on the SAR (Safety Analysis Report) data. Fuel constant axial temperature approximation effect is usual for criticality computations. TRIGA-SSR fuel bundle geometry and materials model for SCALE5-CSAS module allows the introduction of a fuel temperature dependency for the entire fuel active height, using different materials for each fuel bundle region. Previous RELAP5 thermal hydraulic computations for an axial and radial power distribution in the TRIGA fuel pin were done. Fuel constant temperature approximation overestimates pin factors for every core operating at high temperatures. From the thermal hydraulic point of view the worst condition of the storage grid occurs when the transfer channel is accidentally emptied of water from the pool, or the bundle is handled accidentally to remain in air. All the residual heat from the bundles has to be removed without fuel overheating and clad failure. RELAP5 computer code for residual heat removal was used in the assessment of residual heat removal. We made a couple of evaluations of TRIGA bundle clad temperatures in air cooling conditions, with different residual heat levels. The criticality computations have shown that the spent TRIGA fuel bundles storage grid is strongly sub-critical with k(eff) = 0.5951. So, there is no danger for a criticality accident for this storage grid type. The assessment is done for

  3. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

  4. Numerical model for thermal and mechanical behaviour of a CANDU 37-element bundle

    International Nuclear Information System (INIS)

    Prediction of transient fuel bundle deformations is important for assessing the integrity of fuel and the surrounding structural components under different operating conditions including accidents. For numerical simulation of the interactions between fuel bundle and pressure tube, a reliable numerical bundle model is required to predict thermal and mechanical behaviour of the fuel bundle assembly under different thermal loading conditions. To ensure realistic representations of the bundle behaviour, this model must include all of the important thermal and mechanical features of the fuel bundle, such as temperature-dependent material properties, thermal viscoplastic deformation in sheath, fuel-to-sheath interactions, endplate constraints and contacts between fuel elements. In this paper, we present a finite element based numerical model for predicting macroscopic transient thermal-mechanical behaviour of a complete 37-element CANDU nuclear fuel bundle under accident conditions and demonstrate its potential for being used to investigate fuel bundle to pressure tube interaction in future nuclear safety analyses. This bundle model has been validated against available experimental and numerical solutions and applied to various simulations involving steady-state and transient loading conditions. (author)

  5. Investigations on flow induced vibration of simulated CANDU fuel bundles in a pipe

    International Nuclear Information System (INIS)

    In this paper, vibration of a two-bundle string consisting of simulated CANDU fuel bundles subjected to turbulent liquid flow is investigated through numerical simulations and experiments. Large eddy simulation is used to solve the three-dimensional turbulent flow surrounding the fuel bundles for determining fluid excitations. The CFD model includes pipe flow, flow through the inlet fuel bundle along with its two endplates, half of the second bundle and its upstream endplate. The fluid excitation obtained from the fluid model is subsequently fed into a fuel bundle vibration code written in FORTRAN. Fluid structure interaction terms for the fuel elements are approximated using the slender body theory. Simulation results are compared to measurements conducted on the simulated fuel bundles in a testing hydraulic loop. (author)

  6. CAT reconstruction and potting comparison of a LMFBR fuel bundle

    International Nuclear Information System (INIS)

    A standard Liquid Metal Fast Breeder Reactor (LMFBR) subassembly used in the Experimental Breeder Reactor II (EBR-II) was investigated, by remote techniques, for fuel bundle distortion by both nondestructive and destructive methods, and the results from both methods were compared. The non-destructive method employed neutron tomography to reconstruct the locations of fuel elements through the use of a maximum entropy reconstruction algorithm known as MENT. The destructive method consisted of ''potting'' (a technique that embeds and permanently fixes the fuel elements in a solid matrix) the subassembly, and then cutting and polishing the individual sections. The comparison indicated that the tomography reconstruction provided good results in describing the bundle geometry and spacer-wire locations, with the overall resolution being on the order of a spacer-wire diameter. A dimensional consistency check indicated that the element and spacer-wire dimensions were accurately reproduced in the reconstruction

  7. The Conflux Fuel bundle: An Economic and Pragmatic Route to the use of Advanced Fuel Cycles in CANDU Reactors

    International Nuclear Information System (INIS)

    The CANFLEX1 bundle is being developed jointly by AECL and KAERI as a vehicle for introducing the use of enrichment and advanced fuel cycles in CANDU2 reactors. The bundle design uses smaller diameter fuel elements in the outer ring of a 43-element bundle to reduce the maximum element ratings in a CANDU fuel bundle by 20% compared to the 37-element bundle currently in use. This facilitates burnups of greater than 21,000 MW d/TAU to optimize the economic benefit available from the use of enrichment and advanced fuel cycles. A combination of this lower fuel rating, plus development work underway at Aecl to enhance the thermalhydraulic characteristics of the bundle (including both CHF3 and bundle. This provides extra flexibility in the fuel management procedures required for fuel bundles with higher fissile contents. The different bundle geometry requires flow tests to demonstrate acceptable vibration and fretting behavior of the Conflux bundle. A program to undertake the necessary range of flow tests has started at KAERI, involving the fabrication of the required bundles, and setting up for the actual tests. A program to study the fuel management requirements for slightly enriched (0.9 wt % 235 in total U) Conflux fuel has been undertaken by both Aecl and KAERI staff, and further work has started for higher enrichments. Irradiation testing of the Conflux bundle started in the NUR reactor in 1989, and a second irradiation test is due to start shortly. This paper describes the program, and reviews the status of key parts of the program

  8. Enthalpy and void distributions in subchannels of PHWR fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. W.; Choi, H.; Rhee, B. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

  9. Spacer for supporting fuel element boxes

    International Nuclear Information System (INIS)

    A spacer plate unit arranged externally on each side and at a predetermined level of a polygonal fuel element box for mutually supporting, with respect to one another, a plurality of the fuel element boxes forming a fuel element bundle, is formed of a first and a second spacer plate part each having the same length and the same width and being constituted of unlike first and second materials, respectively. The first and second spacer plate parts of the several spacer plate units situated at the predetermined level are arranged in an alternating continuous series when viewed in the peripheral direction of the fuel element box, so that any two spacer plate units belonging to face-to-face oriented sides of two adjoining fuel element boxes in the fuel element bundle define interfaces of unlike materials

  10. LVRF fuel bundle manufacture for Bruce - project update

    Energy Technology Data Exchange (ETDEWEB)

    Pant, A. [Zircatec Precision Industries, Port Hope, Ontario (Canada)

    2005-07-01

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element Bruce LVRF bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37 element fuel over 20 years ago. Introduction of this new line has involved the introduction of significant changes to an environment that is not used to rapid changes with significant impact. At ZPI we have been able to build on our innovative capabilities in new fuel manufacturing, the strength and experience of our core team, and on our prevailing management philosophy of 'support the doer'. The presentation will discuss some of the novel aspects of this fuel introduction and the mix of innovative and classical project management methods that are being used to ensure that project deliverables are being met. Supporting presentations will highlight some of the issues in more detail. (author)

  11. LVRF fuel bundle manufacture for Bruce - project update

    International Nuclear Information System (INIS)

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element Bruce LVRF bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37 element fuel over 20 years ago. Introduction of this new line has involved the introduction of significant changes to an environment that is not used to rapid changes with significant impact. At ZPI we have been able to build on our innovative capabilities in new fuel manufacturing, the strength and experience of our core team, and on our prevailing management philosophy of 'support the doer'. The presentation will discuss some of the novel aspects of this fuel introduction and the mix of innovative and classical project management methods that are being used to ensure that project deliverables are being met. Supporting presentations will highlight some of the issues in more detail. (author)

  12. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    International Nuclear Information System (INIS)

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab

  13. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  14. Flow-induced vibration and acoustic behaviour of CANFLEX-LVRF bundles in a Bruce B NGS fuel channel

    International Nuclear Information System (INIS)

    Frequency/temperature sweep tests were performed in a high-temperature/high-pressure test channel to determine the acoustic and flow-induced vibration characteristics of the CANFLEX-LVRF bundle. The vibratory response of CANFLEX-LVRF bundles was compared with that of 37-element fuel bundles under Bruce B NGS fuel channel normal operating conditions. The tests were performed with a 12-bundle string of CANFLEX-LVRF bundles as well as a mixed string for the transition core. The tests showed that the LVRF bundles performed as required without failure or gross geometry changes. The mixed fuel strings behaved in a manner similar to that of a string of CANFLEX-LVRF bundles. (author)

  15. Filler metals for containers holding irradiated fuel bundles

    International Nuclear Information System (INIS)

    One of the procedures being considered for the disposal of Canadian deuterium uranium (CANDU) irradiated fuel bundles is to place the bundles in containers, fill the containers with metal, and place them underground. This investigation deals with the selection of the filler metal with particular reference to the reaction rate with, and bonding of the filler metal to, the fuel sheathing (Zircaloy 4) and potential container materials. Lead, zinc, and aluminium alloys were examined as potential filler metals. (U.K.)

  16. Fuel element design handbook

    Energy Technology Data Exchange (ETDEWEB)

    Merckx, K.R.

    1958-09-01

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

  17. Interactive hypermedia training manual for spent-fuel bundle counters

    International Nuclear Information System (INIS)

    Spent-fuel bundle counters, developed by the Canadian Safeguards Support Program for the International Atomic Energy Agency, provide a secure and independent means of counting the number of irradiated fuel bundles discharged into the fuel storage bays at CANDU nuclear power stations. Paper manuals have been traditionally used to familiarize IAEA inspectors with the operation, maintenance and extensive reporting capabilities of the bundle counters. To further assist inspectors, an interactive training manual has been developed on an Apple Macintosh computer using hypermedia software. The manual uses interactive animation and sound, in conjunction with the traditional text and graphics, to simulate the underlying operation and logic of the bundle counters. This paper presents the key features of the interactive manual and highlights the advantages of this new technology for training

  18. Fuel bundle geometry and composition influence on coolant void reactivity reduction in ACR and CANDU reactors

    International Nuclear Information System (INIS)

    It is very well known that the CANDU reactor has positive Coolant Void Reactivity (CVR), which is most important criticisms about CANDU. The most recent innovations based on using a thin absorbent Hafnium shell in the central bundle element were successfully been applied to the Advanced CANDU Reactor (ACR) project. The paper's objective is to analyze elementary lattice cell effects in applying such methods to reduce the CVR. Three basic fuel designs in their corresponding geometries were chosen to be compared: the ACR-1000TM, the RU-43 (developed in INR Pitesti) and the standard CANDU fuel. The bundle geometry influence on void effect was also evaluated. The WIMS calculations proved the Hafnium absorber suitability (in the latest 'shell design') to achieve the negative CVR target with great accuracy for the ACR-1000 fuel bundle design than for the other two projects. (authors)

  19. Chop-leach fuel bundle residues densification by melting

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, R.G.; Griggs, B.

    1976-11-01

    Two melting processes were investigated for the densification of fuel bundle residues: Industoslag melting and graphite crucible melting. The Industoslag process, with prior decontamination and sorting, can produce ingots of Zircaloy, stainless steel and Inconel of a quality suitable for refabrication and reuse. The process can also melt oxidized mixtures of fuel bundle residues for direct storage. Eutectic mixtures of these materials can be melted in graphite at temperatures of 1300/sup 0/C. Hydrogen absorption experiments with the zirconium-rich alloys show the alloys to be potential tritium reservoirs. 13 figures.

  20. COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    International Nuclear Information System (INIS)

    The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels

  1. Interconnection of bundled solid oxide fuel cells

    Science.gov (United States)

    Brown, Michael; Bessette, II, Norman F; Litka, Anthony F; Schmidt, Douglas S

    2014-01-14

    A system and method for electrically interconnecting a plurality of fuel cells to provide dense packing of the fuel cells. Each one of the plurality of fuel cells has a plurality of discrete electrical connection points along an outer surface. Electrical connections are made directly between the discrete electrical connection points of adjacent fuel cells so that the fuel cells can be packed more densely. Fuel cells have at least one outer electrode and at least one discrete interconnection to an inner electrode, wherein the outer electrode is one of a cathode and and anode and wherein the inner electrode is the other of the cathode and the anode. In tubular solid oxide fuel cells the discrete electrical connection points are spaced along the length of the fuel cell.

  2. Input modelling of ASSERT-PV V2R8M1 for RUFIC fuel bundle

    International Nuclear Information System (INIS)

    This report describes the input modelling for subchannel analysis of CANFLEX-RU (RUFIC) fuel bundle which has been developed for an advanced fuel bundle of CANDU-6 reactor, using ASSERT-PV V2R8M1 code. Execution file of ASSERT-PV V2R8M1 code was recently transferred from AECL under JRDC agreement between KAERI and AECL. SSERT-PV V2R8M1 which is quite different from COBRA-IV-i code has been developed for thermalhydraulic analysis of CANDU-6 fuel channel by subchannel analysis method and updated so that 43-element CANDU fuel geometry can be applied. Hence, ASSERT code can be applied to the subchannel analysis of RUFIC fuel bundle. The present report was prepared for ASSERT input modelling of RUFIC fuel bundle. Since the ASSERT results highly depend on user's input modelling, the calculation results may be quite different among the user's input models. The objective of the present report is the preparation of detail description of the background information for input data and gives credibility of the calculation results

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

    International Nuclear Information System (INIS)

    AECL has extensive experience with thoria-based fuel irradiations as part of an ongoing R&D program on thorium within the Advanced Fuel Cycles Program. The BDL-422 experiment was one component of the thorium program that involved the fabrication and irradiation testing of six Bruce-type bundles fuelled with (Th, Pu)O2 pellets. The fuel was manufactured in the Recycle Fuel Fabrication Laboratories (RFFL) at Chalk River allowing AECL to gain valuable experience in fabrication and handling of thoria fuel. The fuel pellets contained 86.05 wt. % Th and 1.53 wt. % Pu in (Th, Pu)O2. The objectives of the BDL-422 experiment were to demonstrate the ability of 37-element geometry (Th, Pu)O2 fuel bundles to operate to high burnups up to 1000 MWh/kgHE (42 MWd/kgHE), and to examine the (Th, Pu)O2 fuel performance. This paper describes the post-irradiation examination (PIE) results of BDL-422 fuel bundles irradiated to burnups up to 856 MWh/kgHE (36 MWd/kgHE), with power ratings ranging from 52 to 67 kW/m. PIE results for the high burnup bundles (>1000 MWh/kgHE) are being analyzed and will be reported at a later date. The (Th, Pu)O2 fuel performance characteristics were superior to UO2 fuel irradiated under similar conditions. Minimal grain growth was observed and was accompanied by benign fission gas release and sheath strain. Other fuel performance parameters, such as sheath oxidation and hydrogen distribution, are also discussed. (author)

  4. Coupling Systems of Five CARA Fuel Bundles for Atucha I

    International Nuclear Information System (INIS)

    This paper describe the mechanical design of two options for the coupling systems of five CARA fuel bundles, to be used in the Atucha I nuclear power plant. These systems will be hydraulic tested in a low pressure loop to know their hydraulic loss of pressure

  5. Posttest examination of the VVER-1000 fuel rod bundle CORA-W2

    International Nuclear Information System (INIS)

    The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B4C/stainless steel absorber rod was selected by the OECD/CSNI as International Standard Problem (ISP-36). The experimental results of CORA-W2 serve as data base for comparison with analytical predictions of the high-temperature material behavior by various code systems. The first part of the experimental results is described in KfK 5363 (1994), the second part is documented in this report which contains the destructive post-test examination results. The metallographical and analytical (SEM/EDX) post-test examinations were performed in Germany and Russia and are summarized in five individual contributions. The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B4C absorber rod and the stainless steel grid spacers on the ''low-temperature'' bundle damage initiation and progression. The B4C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occurred. (orig.)

  6. The demonstration irradiation of the CANFLEX-NU fuel bundle in Wolsong NGS 1

    International Nuclear Information System (INIS)

    A demonstration irradiation (DI) of 24 CANFLEX-NU fuel bundles in the high power Q07 channel and low power L21 channel of Wolsong Power Generation Station-1 had been successfully conducted jointly by KEPRI/KHNP/KAERI in the period of 2002 July to 2004 January. The tracking of the reactor operation data showed that the reactor has been stably operated during the DI. One CANFLEX bundle irradiated in the Q07 channel had a typical history of high power and high burnup, where the maximum element linear power rating was ∼ 42 kW/m at the burnup of ∼ 50 MWh/kgU and ∼ 35 kW/m at the discharge element burnup of ∼ 210 MWh/kgU. While, another CANFLEX bundles also irradiated in the Q07 channel had a typical history of power ramping, where the maximum element power ramping-up or -down rate was 28 kW/m. The unusual performance and integrity of the CANFLEX elements could not be found in the ELESTRES predictions and also the in-bay visual examinations showed that all the bundles were intact, free of defects and appeared to be in good condition as expected. Therefore, it is concluded that the demonstration irradiation shows the validation of the CANFLEX bundle performance with direct conditions of relevance under the Korean licensing requirements and the KNFC fuel fabrication capability, and provides the rationale for the decision to perform the full-conversion of CANFLEX fuel in WPGS-1. (author)

  7. A study of coolant thermal mixing within CANDU fuel bundles using ASSERT-PV

    International Nuclear Information System (INIS)

    This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles. The approach taken in the present work is to identify the physical mechanisms contributing to coolant mixing, and to systematically assess the importance of each mechanism. Coupled effects were also considered by flow simulation with mixing mechanisms modelled simultaneously. For the limited range of operating conditions considered and when all mixing mechanisms were modelled simultaneously, the flow was found to be very close to fully mixed. A preliminary model of coolant mixing, suitable for use in the fuel and fuel channel code FACTAR, is also presented. (author)

  8. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

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

  9. NEUTRONIC REACTOR FUEL ELEMENT

    Science.gov (United States)

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

    1961-01-01

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

  10. HLM fuel pin bundle experiments in the CIRCE pool facility

    Energy Technology Data Exchange (ETDEWEB)

    Martelli, Daniele, E-mail: daniele.martelli@ing.unipi.it [University of Pisa, Department of Civil and Industrial Engineering, Pisa (Italy); Forgione, Nicola [University of Pisa, Department of Civil and Industrial Engineering, Pisa (Italy); Di Piazza, Ivan; Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy)

    2015-10-15

    Highlights: • The experimental results represent the first set of values for LBE pool facility. • Heat transfer is investigated for a 37-pin electrical bundle cooled by LBE. • Experimental data are presented together with a detailed error analysis. • Nu is computed as a function of the Pe and compared with correlations. • Experimental Nu is about 25% lower than Nu derived from correlations. - Abstract: Since Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of GEN IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to HLM nuclear reactors. In this frame the Integral Circulation Experiment (ICE) test section has been installed into the CIRCE pool facility and suitable experiments have been carried out aiming to fully investigate the heat transfer phenomena in grid spaced fuel pin bundles providing experimental data in support of European fast reactor development. In particular, the fuel pin bundle simulator (FPS) cooled by lead bismuth eutectic (LBE), has been conceived with a thermal power of about 1 MW and a uniform linear power up to 25 kW/m, relevant values for a LFR. It consists of 37 fuel pins (electrically simulated) placed on a hexagonal lattice with a pitch to diameter ratio of 1.8. The FPS was deeply instrumented by several thermocouples. In particular, two sections of the FPS were instrumented in order to evaluate the heat transfer coefficient along the bundle as well as the cladding temperature in different ranks of sub-channels. Nusselt number in the central sub-channel was therefore calculated as a function of the Peclet number and the obtained results were compared to Nusselt numbers obtained from convective heat transfer correlations available in literature on Heavy Liquid Metals (HLM). Results reported in the present work, represent the first set of experimental data concerning fuel pin bundle behaviour in a heavy liquid metal pool, both in forced and

  11. Selection of instruments used for vibration measurement of fuel bundles in a pressure tube under CANDU reactor operating conditions

    International Nuclear Information System (INIS)

    Vibration characteristics of CANDU fuel bundle and fuel elements is a key parameter considered in the design of a fuel bundle. Out-reactor frequency and temperature sweep tests, under reactor operating conditions, are performed to verify vibration characteristics of CANDU fuel bundles. Several options have been considered in the selection of vibration instrumentation to perform out-reactor frequency and temperature sweep tests. This paper compares the benefits and disadvantages of various vibration instruments and summarizes the rationale behind the selection of instruments used for vibration measurements over a range of temperature and pressure pulsation frequencies. The conclusions are presented from the bench tests performed, which confirm the use of the selected instruments. (author)

  12. Development of neural network for analysis of local power distributions in BWR fuel bundles

    International Nuclear Information System (INIS)

    A neural network model has been developed to learn the local power distributions in a BWR fuel bundle. A two layers neural network with total 128 elements is used for this model. The neural network learns 33 cases of local power peaking factors of fuel rods with given enrichment distribution as the teacher signals, which were calculated by a fuel bundle nuclear analysis code based on precise physical models. This neural network model studied well the teacher signals within 1 % error. It is also able to calculate the local power distributions within several % error for the different enrichment distributions from the teacher signals when the average enrichment is close to 2 %. This neural network is simple and the computing speed of this model is 300 times faster than that of the precise nuclear analysis code. This model was applied to survey the enrichment distribution to meet a target local power distribution in a fuel bundle, and the enrichment distribution with flat power shape are obtained within short computing time. (author)

  13. Fuel Element Technical Manual

    Energy Technology Data Exchange (ETDEWEB)

    Burley, H.H. [ed.

    1956-08-01

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

  14. Nuclear fuel element

    Science.gov (United States)

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

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

  15. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

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

  16. Status of the demonstration irradiation program of the new fuel bundle CANFLEX-NU in Korea

    International Nuclear Information System (INIS)

    In the late part of 1999, the Korea Electric Power Corporation has initiated a program CANFLEX-NU (Natural Uranium) fuel in the Wolsong Generating Station (WGS) - no.1 which has been operating since 1983, because the CANFLEX could be used to recover some of a CANDU heat transport system operation margins that had decreased due to The Korea Ministry of Science and Technology (MOST) has recognized the successful demonstration irradiation of 24 CANFLEX bundles at the Pt. Lepreau Generating Station in Canada, as final verification of the CANFLEX design in preparation for full core conversion. Therefore, MOST has pushed and gave a financial support to a KEPRI/KAERI Joint Industrialization Program of CANFLEX-NU Fuel, which will be for 3 years from 2000 November, to validate CANFLEX-NU fuel bundle performance in direct conditions of relevance under the Korean licensing requirements as well as to evaluate the fuel fabrication capability, and to produce a safety analysis report for the full-core implementation. The economic benefits of CANFLEX-NU fuel are directly dependent on the thermalhydraulic performance. Switching from the existing 37-element fuel to the CANFLEX fuel will be largely driven by the economic benefits to be realized. Showing a positive result in the economic evaluation as well as successfully demonstrating the CANFLEX fuel irradiation in WGS-no. 1, the full-core implementation of the fuel at the WGS-no.1 in Korea will proceed by starting the licensing process at around 2003 April because the safety report for the full-core conversion will be ready by 2003 March. This paper describes the status of CANFLEX-NU fuel industrialization program in Korea, as well as the fuel design features. It summarizes the plan of CANFLEX-NU fuel demonstration irradiation at the WGS-no. 1 in Korea and the status of documentation for the demonstration irradiation as well as for the CANFLEX-NU full-core implementation. (author)

  17. Studies of a larger fuel bundle for the ABWR improved evolutionary reactor

    International Nuclear Information System (INIS)

    Studies for an Improved Evolutionary Reactor (IER) based on the Advanced Boiling Water Reactor (ABWR) were initiated in 1990. The author summarizes the current status of the core and fuel design. A core and fuel design based on a BWR K-lattice fuel bundle with a pitch larger than the conventional BWR fuel bundle pitch is under investigation. The core and fuel design has potential for improved core design flexibility and improved reactor transient response. Furthermore, the large fuel bundle, coupled with a functional control rod layout, can achieve improvement of operation and maintenance, as well as improvement of overall plant economy

  18. Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

  19. Laser dismantling of PHWR spent fuel bundles and decladding of fuel pins in the highly radioactive hot cells

    International Nuclear Information System (INIS)

    Full text: For reprocessing of PHWR fuel, fuel bundles are at present chopped mechanically into small pieces of pins using high tonnage mechanical press before dissolution. The existing method of bundle dismantling is purely mechanical using very high force for chopping. A laser based automated bundle dismantling system is developed. In the system, end-plates of bundle, which holds the fuel pins together, are cut using Nd-YAG laser to separate the bundles into pins. In addition to pin separation, the pins are to be chopped into small pieces using a small mechanical chopper. Since the spent fuel is highly radioactive, all these operations are performed remotely in hot cells. Post irradiation examination also requires dismantling of bundles into pins so that they can select the pins for the further examinations. In both these applications laser dismantling remains the most. important step and this system has been developed and tested. This paper describes the experience gained during the development efforts

  20. 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.

  1. MENT reconstruction and potting comparison of a LMFBR fuel bundle

    International Nuclear Information System (INIS)

    Since the advent of computer-assisted-tomography (CAT), the CAT techniques have been rapidly expanded to the nuclear industry. A number of investigators have applied these techniques to reconstruct the fuel bundle configuration inside a subassembly with various degrees of resolution; however, there has been little data available on the accuracy of these reconstructions, and no comparisons have been made with the internal structure of actual irradiated subassemblies. Some efforts have utilized pretest mock-ups to calibrate the CAT algorithms, but the resulting mock-up configurations do not necessarily represent an actual subassembly, so an exact comparison has been lacking. The purpose of this paper is to present the results of a comparison between a CAT reconstruction of an irradiated subassembly and the destructive examination of the same subassembly

  2. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

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

  3. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

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

  4. Experimental study of water flow in nuclear fuel elements

    International Nuclear Information System (INIS)

    This work aims to develop an experimental methodology for investigating the water flow through rod bundles after spacer grids of nuclear fuel elements of PWR type reactors. Speed profiles, with the device LDV (Laser Doppler Velocimetry), and the pressure drop between two sockets located before and after the spacer grid, using pressure transducers were measured

  5. Nuclear reactor fuel element splitter

    International Nuclear Information System (INIS)

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

  6. Spacer for a fuel element

    International Nuclear Information System (INIS)

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

  7. The behaviour of Phenix fuel pin bundle under irradiation

    International Nuclear Information System (INIS)

    An entire Phenix sub-assembly has been mounted and sectioned after irradiation. The examination of cross-sections revealed the effects of mechanical interaction in the bundle (ovalisations and contacts between clads). According to analysis of the sodium channels, cooling of the pin bundle remained uniform. (author)

  8. 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)

  9. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

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

  10. Fuel element for nuclear reactor

    International Nuclear Information System (INIS)

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

  11. Increased burnup of fuel elements

    International Nuclear Information System (INIS)

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

  12. Averaging methods of the gap heat transfer coefficients and the loss form coefficients of nuclear reactor cores loaded with different fuel bundles

    International Nuclear Information System (INIS)

    When performing transient analysis in heterogeneous nuclear reactors loaded with different types of fuel bundles is necessary to model the reactor core by a few representative fuel elements with average properties of a region containing a large number of fuel elements. The properties of these representative fuel bundles are obtained by averaging the thermal-hydraulic properties of the fuel elements contained in each region. In this paper we study the different ways to perform the averaging of the thermal-hydraulic properties that can have an influence on the transient results for licence purposes. Also we study the influence of the different averaging methods on the peak clad temperature (PCT) evolution for a LOCA, and on the critical power ratio (CPR) in the hot channels for a turbine trip transient without bypass credit.

  13. Overview of methods to increase dryout power in CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com [Chalk River Laboratories, AECL, Chalk River (Canada); University of Ottawa, Department of Mechanical Engineering, Ottawa (Canada); Leung, L.K.H. [Chalk River Laboratories, AECL, Chalk River (Canada); Park, J.H. [Korean Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-06-15

    Highlights: • Small changes in bundle geometry can have noticeable effects on the bundle CHF. • Rod spacing devices can results in increases of over 200% in CHF. • CHF enhancement decays exponentially downstream from spacers. • CHF-enhancing bundle appendages also increase the post-CHF heat transfer. - Abstract: In CANDU reactors some degradation in the CCP (critical channel power, or power corresponding to the first occurrence of CHF in any fuel channel) will occur with time because of ageing effects such as pressure-tube diametral creep, increase in reactor inlet-header temperature, increased hydraulic resistance of feeders. To compensate for the ageing effects, various options for recovering the loss in CCP are described in this paper. They include: (i) increasing the bundle heated perimeter, (ii) optimizing the bundle configuration, (iii) optimizing core flow and flux distribution, (iv) reducing the bundle hydraulic resistance, (v) use of CHF-enhancing bundle appendages, (vi) more precise experimentation, and (vii) redefining CHF. The increase in CHF power has been quantified based on experiments on full-scale bundles and subchannel code predictions. The application of several of these CHF enhancement principles has been used in the development of the 43-rod CANFLEX bundle.

  14. System for supporting a bundled tube fuel injector within a combustor

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-21

    A combustor includes an end cover having an outer side and an inner side, an outer barrel having a forward end that is adjacent to the inner side of the end cover and an aft end that is axially spaced from the forward end. An inner barrel is at least partially disposed concentrically within the outer barrel and is fixedly connected to the outer barrel. A fluid conduit extends downstream from the end cover. A first bundled tube fuel injector segment is disposed concentrically within the inner barrel. The bundled tube fuel injector segment includes a fuel plenum that is in fluid communication with the fluid conduit and a plurality of parallel tubes that extend axially through the fuel plenum. The bundled tube fuel injector segment is fixedly connected to the inner barrel.

  15. Fuel element development

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Advances in the manufacture of clad tubes and components for PHWR fuel bundle

    International Nuclear Information System (INIS)

    Fuel bundles for Pressurized Heavy Water Reactors (PHWRs) consists of Uranium di-oxide pellets encapsulated into thin wall Zircaloy clad tubes. Other components such as end caps, bearing pads and spacer pads are the integral elements of the fuel bundle. As the fuel assembly is subjected to severe operating conditions of high temperature and pressure in addition to continual irradiation exposure, all the components are manufactured conforming to stringent specifications with respect to chemical composition, mechanical & metallurgical properties and dimensional tolerances. The integrity of each component is ensured by NDE at different stages of manufacture. The manufacturing route for fuel tubes and components comprise of a combination of thermomechanical processing and each process step has marked effect on the final properties. The fuel tubes are manufactured by processing the extruded blanks in four stage cold pilgering with intermediate annealing and final stress relieving operation. The bar material is produced by hot extrusion followed by multi-pass swaging and intermediate annealing. Spacer pads and bearing pads are manufactured by blanking and coining of Zircaloy sheet which is made by a combination of hot and cold rolling operations. Due to the small size and stringent dimensional requirements of these appendages, selection of production route and optimization of process parameters are important. This paper discusses about various measures taken for improving the recoveries and mechanical and corrosion properties of the tube, sheet and bar materials being manufactured at Nuclear Fuel Complex, Hyderabad For the production of clad tubes, modifications at extrusion stage to reduce the wall thickness variation, introduction of ultrasonic testing of extruded blanks, optimization of cold working and heat treatment parameters at various stages of production etc. were done. The finished bar material is subjected to 100% Ultrasonic and eddy current testing to ensure

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

    International Nuclear Information System (INIS)

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

  19. Validation study of thermal-hydraulic analysis program spiral for fuel pin bundle of sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Full text of publication follows: Japan Nuclear Cycle Development Institute (JNC) has been developing a numerical simulation system in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel subassemblies of sodium-cooled fast reactors under various operating conditions such as normal operation, transient condition or deformed geometry condition from the viewpoint of the assessment of fuel pin structure integrity. This paper describes the validation study of SPIRAL that is one component code of the numerical simulation system and contributes to detailed simulations of local flow and temperature fields in a wire-wrapped fuel pin bundle. SPIRAL is a multi-dimensional finite element method code that can treat complicated geometries like a fuel pin bundle. For numerical stabilization, one can choose Streamline Upwind Petrov Galerkin method and Balancing Tensor Diffusivity method. Semi-implicit solution scheme (fractional step method) developed by Ramaswamy is used for time integration. As the pressure equation matrix solver, ICCG or Gaussian elimination is applied. Energy conservation equations of coolant and structure are also solved and therefore temperature distributions of both coolant and fuel pins can be calculated. Several turbulence models, high/low Reynolds number isotropic/anisotropic models, were incorporated to the code. The code was parallelized using MPI for enhancing simulation efficiency. Pre-processor is also available for numerical grid generation for wire-wrapped fuel pin bundles by curvilinear coordinate system. Fundamental validity related to solving mass, momentum and energy conservation equations and applicability of turbulence models were confirmed by simulating several basic problems. As typical examples, two kinds of simulations using high Re number models, backward facing step flow and 4- fuel-pin bundle in rectangular duct, are introduced in this paper. The simulation results indicate that RNG k-ε model shows relatively

  20. 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

  1. Temperature Distributions in LMR Fuel Pin Bundles as Modeled by COBRA-IV-I

    Science.gov (United States)

    Wright, Steven A.; Stout, Sherry

    2005-02-01

    Most pin type reactor designs for space power or terrestrial applications group the fuel pins into a number of relatively large fuel pin bundles or subassemblies. Fuel bundles for terrestrial liquid metal fast breeders reactors typically use 217 - 271 pins per sub-assembly, while some SP100 designs use up to 331 pins in a central subassembly that was surrounded by partial assemblies. Because thermal creep is exponentially related to temperature, small changes in fuel pin cladding temperature can make large differences in the lifetime in a high temperature liquid metal reactor (LMR). This paper uses the COBRA-IV-I computer code to determine the temperature distribution within LMR fuel bundles. COBRA-IV-I uses the sub-channel analysis approach to determine the enthalpy (or temperature) and flow distribution in rod bundles for both steady-state and transient conditions. The COBRA code runs in only a few seconds and has been benchmarked and tested extensively over a wide range of flow conditions. In this report the flow and temperature distributions for two types of lithium cooled space reactor core designs were calculated. One design uses a very tight fuel pin packing that has a pitch to diameter ratio of 1.05 (small wire wrap with a diameter of 392 μm) as proposed in SP100. The other design uses a larger pitch to diameter ratio of 1.09 with a larger more conventional sized wire wrap diameter of 1 mm. The results of the COBRA pin bundle calculations show that the larger pitch-to-diameter fuel bundle designs are more tolerant to local flow blockages, and in addition they are less sensitive to mal-flow distributions that occur near the edges of the subassembly.

  2. Measurement and CFD calculation of spacer loss coefficient for a tight-lattice fuel bundle

    International Nuclear Information System (INIS)

    Highlights: • Experiment and CFD analysis evaluated the pressure drop in a spacer grid. • The measurement and CFD errors for the spacer loss coefficient were estimated. • The spacer loss coefficient for the dual-cooled annular fuel bundle was determined. • The CFD prediction agrees with the measured spacer loss coefficient within 8%. - Abstract: An experiment and computational fluid dynamics (CFD) analysis were performed to evaluate the pressure drop in a spacer grid for a dual-cooled annular fuel (DCAF) bundle. The DCAF bundle for the Korean optimum power reactor (OPR1000) is a 12 × 12 tight-lattice rod array with a pitch-to-diameter ratio of 1.08 owing to a larger outer diameter of the annular fuel rod. An experiment was conducted to measure the pressure drop in spacer grid for the DCAF bundle. The test bundle is a full-size 12 × 12 rod bundle with 11 spacer grid. The test condition covers a Reynolds number range of 2 × 104–2 × 105 by changing the temperature and flow rate of water. A CFD analysis was also performed to predict the pressure drop through a spacer grid using the full-size and partial bundle models. The pressure drop and loss coefficient of a spacer grid were predicted and compared with the experimental results. The CFD predictions of spacer pressure drop and loss coefficient agree with the measured values within 8%. The spacer loss coefficient for the DCAF bundle is estimated to be approximately 1.50 at a nominal operating condition of OPR1000, i.e., Re = 4 × 105

  3. Study of matrix crack-tilted fiber bundle interaction using caustics and finite element method

    Science.gov (United States)

    Hao, Wenfeng; Zhu, Jianguo; Zhu, Qi; Yuan, Yanan

    2016-02-01

    In this work, the interaction between the matrix crack and a tilted fiber bundle was investigated via caustics and the finite element method (FEM). First, the caustic patterns at the crack tip with different distances from the tilted fiber were obtained and the stress intensity factors were extracted from the geometry of the caustic patterns. Subsequently, the shielding effect of the fiber bundle in front of the crack tip was analyzed. Furthermore, the interaction between the matrix crack and the broken fiber bundle was discussed. Finally, a finite element simulation was carried out using ABAQUS to verify the experimental results. The results demonstrate that the stress intensity factors extracted from caustic experiments are in excellent agreement with the data calculated by FEM.

  4. Apparatus for locating defective nuclear fuel elements

    International Nuclear Information System (INIS)

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

  5. FEED 1.6: modelling of hydrogen diffusion and precipitation in fuel bundle zircaloy components

    International Nuclear Information System (INIS)

    An as-fabricated Zircaloy component in a CANDU® fuel bundle has certain amount of hydrogen. In addition, the Zircaloy component pickups hydrogen during operation, where sheath oxidation occurs on the water side. Hydrogen content in the Zircaloy component will change due to the diffusion under gradients of concentration and temperature. A hydrostatic stress gradient may also have some effect on hydrogen diffusion. When the local concentration of hydrogen exceeds the terminal solid solubility (TSS), hydrides will start to form (i.e., hydride precipitation). Because hydrides have a negative effect on material properties (e.g., lower ductility), the hydrogen content in Zircaloy sheath needs to be limited to ensure that the sheath strength is not affected. The FEED (Finite Element Estimate for Diffusion) code was developed to predict the local hydrogen concentration and formation of hydride. The FEED 1.6 code has the following capabilities: Model transient Hydrogen/Deuterium (H/D) diffusion in Zircaloy components (e.g., fuel sheath, endcap and endcap weld); Model H/D pickup in Zircaloy sheath; Account for the effect of gradients of concentration, temperature and stress; and, Model transient hydride precipitation and re-dissolutions. This paper describes the FEED 1.6 code, including theory, models, and some validation examples. (author)

  6. FEED 1.6: modelling of hydrogen diffusion and precipitation in fuel bundle zircaloy components

    Energy Technology Data Exchange (ETDEWEB)

    Lai, L.; Xu, Z.; Jiang, Q.; Cheng, G. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2010-07-01

    An as-fabricated Zircaloy component in a CANDU® fuel bundle has certain amount of hydrogen. In addition, the Zircaloy component pickups hydrogen during operation, where sheath oxidation occurs on the water side. Hydrogen content in the Zircaloy component will change due to the diffusion under gradients of concentration and temperature. A hydrostatic stress gradient may also have some effect on hydrogen diffusion. When the local concentration of hydrogen exceeds the terminal solid solubility (TSS), hydrides will start to form (i.e., hydride precipitation). Because hydrides have a negative effect on material properties (e.g., lower ductility), the hydrogen content in Zircaloy sheath needs to be limited to ensure that the sheath strength is not affected. The FEED (Finite Element Estimate for Diffusion) code was developed to predict the local hydrogen concentration and formation of hydride. The FEED 1.6 code has the following capabilities: Model transient Hydrogen/Deuterium (H/D) diffusion in Zircaloy components (e.g., fuel sheath, endcap and endcap weld); Model H/D pickup in Zircaloy sheath; Account for the effect of gradients of concentration, temperature and stress; and, Model transient hydride precipitation and re-dissolutions. This paper describes the FEED 1.6 code, including theory, models, and some validation examples. (author)

  7. Compact Fuel Element Environment Test

    Science.gov (United States)

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

    2012-01-01

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

  8. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others

    1995-09-01

    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

  9. Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Najeeb, Umair

    This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

  10. Consideration of subchannel area of a 37-element fuel to enhance CHF

    International Nuclear Information System (INIS)

    CANDU-6 reactor has 380 fuel channels of a pressure tube type, which provides an independent flow passage, and each pressure tube contains 12 fuel bundles horizontally. The CHF of a CANDU fuel bundle in a horizontal fuel channel is one of the important parameters determining the thermalhydraulic safety margin as well as the trip set point of the Regional Overpower Protection (ROP) system. Hence, the CHF enhancement of a CANDU fuel bundle has been an issue for a long time and can be affected by the geometric configuration of the fuel elements as well as several appendages such as the end-plates, bearing pads, and spacers attached to the fuel elements. This paper considers the modification of the inner ring radius of a standard 37-element fuel bundle to enhance the CHF, since the CHFs of a standard 37-element fuel bundle preferably occur at the peripheral subchannels of the center rod, owing to the relative small flow area or high flow resistance under high flow conditions or the normal operating conditions of a CANDU reactor. Subchannel analysis techniques using the ASSERT-PV code were applied to investigate the local CHF characteristics according to the inner ring radius variation for the original diameter of the pressure tube. It was found that the modification of the inner ring radius is very effective in enhancing the dryout power of the fuel bundle under the reactor operating conditions through an enthalpy re-distribution of the subchannels and change in the local locations of the first CHF occurrences. (author)

  11. Prediction of temperature distribution in a fast reactor spent fuel bundle

    International Nuclear Information System (INIS)

    A simple mathematical model is described for predicting temperature distribution in a spent fuel bundle. The model takes into account γ-ray leakage, radiant and conductive heat transports between the various fuel pins arranged in a triangular array and enclosed in a hexagonal shaped tube containing gaseous medium. With the geometry of the fuel bundle the configuration factors between various fuel pins can be calculated. The configuration factors along with the heat generation rates, net γ-ray leakage, surface emissivity, conductivity of the enclosed medium and the temperature of the hexagonal tube can be used to estimate the temperature distribution with the help of the computer code TICOFUSA developed on the basis of this model. (author)

  12. Behavior of AgInCd absorber material in Zry/UO{sub 2} fuel rod simulator bundles tested at high temperatures in the CORA facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-01-15

    The CORA experiments carried out in an out-of-pile facility at the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, are part of the ''Severe Fuel Damage'' (SFD) program. The experimental program is to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in a few cases up to 2400 C. In the CORA experiments two different bundle configurations are tested: PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles. The PWR-type assemblies usually consist of 25 rods with 16 electrically heated fuel rod simulators and nine unheated rods (full-pellet and absorber rods). Bundle CORA-5 contained one Ag/In/Cd - steel absorber rod whereas two absorber rods were used in CORA-12, CORA-15, and CORA-9. The larger bundle CORA-7 contained 5 absorber rods. CORA-12 was terminated by quenching with water from the bottom. In CORA-15 the heated and unheated rods were pressurized to achieve pronounced clad ballooning. Bundle CORA-9 was tested with a system pressure of 1.0 MPa instead of 0.22 MPa. The test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of a LWR. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy (Zry)-steam reaction started at about 1100 C, leading the bundles to maximum temperatures of approximately 2000 C. Rod destruction started with the failure of the absorber rod cladding at about 1200 C, i.e. about 250 K below the melting regime of steel. Penetration of the steel cladding was presumably caused by a eutectic interaction between steel and the zircaloy guide tube. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO{sub 2} interaction

  13. FINITE ELEMENT ANALYSIS OF LAMINAR FLOW AND HEAT TRANSFER IN A BUNDLE OF CYLINDERS

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    In this paper two-dimensional Navier-Stokes equations and energy equation governing incompressible laminar flow past a bundle of cylinders were numerically solved by using the finite element method.The velocity correction method was used for time advancement, and spatial discretization was carried out with the Galerkin weighted residual method.Viscous flows past the cylinder banks arranged in in-line cylinder bundles and staggered cylinder bundles, coupled with heat transfer, were investigated for pitch-diameter ratios of 1.5 and 2.0 and the Reynolds numbers from 50 to 500.Flow structures and heat transfer behavior were discussed.The results obtained agree well with available numerical data.

  14. Calculation of power coefficient in CANFLEX-NU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    Changes in power level affect reactivity due to its dependence on fuel and coolant temperatures. The power coefficient of reactivity is related to the fuel temperature coefficient through the change in fuel temperature per percent change in power. In addition, power level changes are followed by changes in coolant temperature and density which contribute to the reactivity effect. In this report, the power coefficient of CANFLEX-NU was calculated and the result would be compared with that of CANDU-6 reactor which is operating. 8 refs., 43 figs., 2 tabs. (Author)

  15. ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

    International Nuclear Information System (INIS)

    Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles

  16. ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca; Nava-Dominguez, A., E-mail: navadoma@aecl.ca

    2014-08-15

    Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles.

  17. Transportation of irradiated fuel elements

    International Nuclear Information System (INIS)

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

  18. A high temperature fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Sekido, A.; Nakai, M.; Ninomiya, Y.

    1982-12-21

    A solid electrolyte which conducts electricity with heating by oxygen ions and operates at a temperature of 1,000C is used in the element. The cathode, besides the ionic conductivity in oxygen, has an electron conductivity. The anode has electron conductivity. Substances such as Bi203, into which oxides of alkaline earth metals are added, are used for making the cathode. The electrolyte consists of ZrO2 and Y2O3, to which CaO is added. WC, to which an H2 type fuel is fed, serves as the anode. The element has a long service life.

  19. Study on the effect of the CANFLEX-NU fuel element bowing on the critical heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Ho Chun; Cho, Moon Sung; Jeon, Ji Su

    2001-01-01

    The effect of the CANFLEX-NU fuel element bowing on the critical heat flux is reviewed and analyzed, which is requested by KINS as the Government design licensing condition for the use of the fuel bundles in CANDU power reactors. The effect of the gap between two adjacent fuel elements on the critical heat flux and onset-of-dryout power is studied. The reduction of the width of a single inter-rod gap from its nominal size to the minimum manufacture allowance of 1 mm has a negligible effects on the thermal-hydraulic performance of the bundle for the given set of boundary conditions applied to the CANFLEX-43 element bundle in an uncrept channel. As expected, the in-reactor irradiation test results show that there are no evidence of the element bow problems on the bundle performance.

  20. Development of neural network simulating power distribution of a BWR fuel bundle

    International Nuclear Information System (INIS)

    A neural network model is developed to simulate the precise nuclear physics analysis program code for quick scoping survey calculations. The relation between enrichment and local power distribution of BWR fuel bundles was learned using two layers neural network (ENET). A new model is to introduce burnable neutron absorber (Gadolinia), added to several fuel rods to decrease initial reactivity of fresh bundle. The 2nd stages three layers neural network (GNET) is added on the 1st stage network ENET. GNET studies the local distribution difference caused by Gadolinia. Using this method, it becomes possible to survey of the gradients of sigmoid functions and back propagation constants with reasonable time. Using 99 learning patterns of zero burnup, good error convergence curve is obtained after many trials. This neural network model is able to simulate no learned cases fairly as well as the learned cases. Computer time of this neural network model is about 100 times faster than a precise analysis model. (author)

  1. Fuel elements of thermionic converters

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-01-01

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

  2. Resistance factors, two phase multipliers and void fractions for best estimate flow calculations in Dodewaard fuel bundles

    International Nuclear Information System (INIS)

    Values are given for resistance factors, two phase multipliers and core and chimney void fractions in the fuel and chimney to be used in best estimate calculations of the flow in Dodewaard fuel bundles. The resistance factors are based on single phase experimental data for a mockup of the Dodewaard fuel bundle. The two phase multipliers are determined from two phase measurements of mockups of other fuel bundles for nuclear reactors. This is also true for the in bundle void fractions. The void fractions in the chimney have been validated by measured void fractions in large diameter pipes. The recommended changes to the existing input for calculations are somewhat larger than the uncertainties in the measurements. (author). 37 refs.; 48 figs.; 4 tabs

  3. Nuclear reactor fuel element. Kernreaktorbrennelement

    Energy Technology Data Exchange (ETDEWEB)

    Lippert, H.J.

    1985-03-28

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Byung Ok; Hong, K. P.; Park, D. G.; Choo, Y. S.; Baik, S. J.; Kim, K. H.; Kim, H. C.; Jung, Y. H

    2001-05-01

    The post irradiation examination of a HANARO mini-element nuclear fuel, KH96C-004, was done in June 6, 2000. The purpose of this project is to evaluate the in-core performance and reliability of mini-element nuclear fuel for HANARO developed by the project ''The Nuclear Fuel Material Development of Research Reactor''. And, in order to examine the performance of mini-element nuclear fuel in normal output condition, the post irradiation examination of a nuclear fuel bundle composed by 6 mini nuclear fuel rods and 12 dummy fuel rods was performed. Based on these examination results, the safety and reliability of HANARO fuel and the basic data on the design of HANARO nuclear fuel can be ensured and obtained,.

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

    International Nuclear Information System (INIS)

    The post irradiation examination of a HANARO mini-element nuclear fuel, KH96C-004, was done in June 6, 2000. The purpose of this project is to evaluate the in-core performance and reliability of mini-element nuclear fuel for HANARO developed by the project The Nuclear Fuel Material Development of Research Reactor. And, in order to examine the performance of mini-element nuclear fuel in normal output condition, the post irradiation examination of a nuclear fuel bundle composed by 6 mini nuclear fuel rods and 12 dummy fuel rods was performed. Based on these examination results, the safety and reliability of HANARO fuel and the basic data on the design of HANARO nuclear fuel can be ensured and obtained,

  6. A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

    Science.gov (United States)

    Chandraker, Dinesh Kumar; Vijayan, Pallipattu Krishnan; Sinha, Ratan Kumar; Aritomi, Masanori

    The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were

  7. Application of the finite element method in the modelling of coil bundles

    International Nuclear Information System (INIS)

    Three different FEM approaches are presented and evaluated as viable interpretations of an actual coil, each limited for use within specified parameter ranges. One is based on solid elements with correctly defined properties permitting the accurate representation of the global behavior of a coil bundle. The other two are more complex and are based on the combination of various elements each accounting for a different aspect of coil behavior which are best resolved via multi-level substructuring. The choice of the best model for the job rests with the analyst who must first resolve what the goals of the analysis are and given the parameters of the problem, which models can be used. The basic idea behind these models is the application of a systematic modelling technique requiring a close correspondence between the capability of the FE themselves and the true mechanical behavior of that portion of the coil being simulated. In order to have analytical solutions for confirming the bending and torsional capabilities of these coil bundle FEM, their behavior is studied via several basic examples. Laminated beam behavior which categorizes the structural nature of many conventional coil bundles is also examined in some depth. Also discussed is a generalized computer program that was developed to accept the description of any conventional coil section and determine an effective stiffness for it to be used in FEM. The various methodologies described in this paper should be applicable to any bundled coil design. Although only conventional coils are discussed, with the proper modifications the concepts and techniques presented can be applied to other configurations as well, such as superconductors. (orig./HP)

  8. Study of thermal hydraulic behavior of supercritical water flowing through fuel rod bundles

    International Nuclear Information System (INIS)

    Investigations on thermal-hydraulic behavior in Supercritical Water Reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community because of its potential to obtain high thermal efficiency and compact design. Present work deals with CFD analysis to study the flow and heat transfer behavior of supercritical water in 4 metre long 7-pin fuel bundle using commercial CFD package ANSYS CFX for single phase steady state conditions. Considering the symmetric conditions, 1/12th part of the fuel rod bundle is taken as a domain of analysis. RNG K-epsilon model with scalable wall functions is used for modeling the turbulence behavior. Constant heat flux boundary condition is applied at the fuel rod surface. IAPWS equations of state are used to compute thermo-physical properties of supercritical water. Sharp variations in its thermo-physical properties (specific heat, density) are observed near the pseudo-critical temperature causing sharp change in heat transfer coefficient. The pseudo-critical point initially appears in the gaps among heated fuel rods, and then spreads radially outward reaching the adiabatic wall as the flow goes downstream. The enthalpy gain in the centre of the channel is much higher than that in the wall region. Non-uniformity in the circumferential distribution of surface temperature and heat transfer coefficient is observed which is in agreement with published literature. Heat transfer coefficient is high on the rod surface near the tight region and decreases as the distance between rod surfaces increases. (author)

  9. The effect of radial power profile of DUPIC bundle on CHF

    International Nuclear Information System (INIS)

    The axial and ring power profiles of DUPIC bundle are much different from those of reference 37-element fuel bundle since a DUPIC fuel bundle is re-fabricated using spent PWR fuel and 2-bundle shift refuelling scheme is proposed to CANDU-6 reactor. In case that the ring power profile of a fuel bundle is altered, the flow and enthalpy distribution of subchannels and the radial position of CHF occurrence will be changed. Similarly, the axial power profile of a fuel channel affects CHF and axial position of CHF occurrence as well as axial enthalpy, quality and pressure distribution. The ring power profile of the DUPIC bundle as increasing burnup is altered and flattened compared to 37-element bundle and each fuel bundle in a fuel channel has a different ring power profile from the other bundles at different axial position in the same fuel channel. Therefore, how to consider the burnup or ring power effect on CHF is very important to DUPIC thermalhydraulic analysis. At present study, thermalhydraulic analysis of the DUPIC bundle was performed in consideration of ring power profile effect on CHF. The subchannel enthalpy, mass flux and CHF distribution for 0 burnup to discharged burnup (18,000 MWD/THM) of DUPIC bundle were evaluated using ASSERT subchannel code. The results were compared to those of 37-element bundle and the compatability of DUPIC bundle with an existing CANDU-6 was presented in a CHF point of view

  10. Study of heat transfer in a 7-element bundle cooled with the upward flow of supercritical Freon-12

    Science.gov (United States)

    Richards, Graham

    Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are: 1) small number of operating SCW experimental setups and 2) difficulties in testing and experimental costs at very high pressures, temperatures and heat fluxes. However, SuperCritical Water-cooled nuclear Reactor (SCWRs) designs cannot be finalized without such data. Therefore, as a preliminary approach experiments in SCW-cooled bare tubes and in bundles cooled with SC modeling fluids can be used. One of the SC modeling fluids typically used is Freon- 12 (R-12) where the critical pressure is 4.136 MPa and the critical temperature is 111.97ºC. These conditions correspond to a critical pressure of 22.064 MPa and critical temperature of 373.95ºC in water. A set of experimental data obtained in a Freon-12 cooled vertical bare bundle at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia) was analyzed. This set consisted of 20 cases of a vertically oriented 7-element bundle installed in a hexagonal flow channel. To secure the bundle in the flow channel 3 thin spacers were used. The dataset was obtained at equivalent parameters of the proposed SCWR concepts. Data was collected at pressures of about 4.65 MPa for several different combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. Heat fluxes ranged from 9 kW/m2 to 120 kW/m2 and mass fluxes ranged from 440 kg/m2s to 1320 kg/m2s. Also inlet temperatures ranged from 70ºC -- 120ºC. The test section consisted of fuel elements that were 9.5 mm in diameter with the total heated length of 1 m. Bulk-fluid and wall temperature profiles were recorded using a combination of 8 different thermocouples. The data was analyzed with respect to its temperature profile and heat transfer coefficient along the heated length of the test section. In a previous study it was confirmed that there is the existence of three distinct

  11. Automated Fuel Element Closure Welding System

    International Nuclear Information System (INIS)

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

  12. Gamma spectrometry of TRIGA fuel elements

    International Nuclear Information System (INIS)

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

  13. International experience in conditioning spent fuel elements

    International Nuclear Information System (INIS)

    The purpose of this report is to compile and present in a clear form international experience (USA, Canada, Sweden, FRG, UK, Japan, Switzerland) gained to date in conditioning spent fuel elements. The term conditioning is here taken to mean the handling and packaging of spent fuel elements for short- or long-term storage or final disposal. Plants of a varying nature fall within this scope, both in terms of the type of fuel element treated and the plant purpose eg. experimental or production plant. Emphasis is given to plants which bear some similarity to the concept developed in Germany for direct disposal of spent fuel elements. Worldwide, however, relatively few conditioning plants are in existence or have been conceived. Hence additional plants have been included where aspects of the experience gained are also of relevance eg. plants developed for the consolidation of spent fuel elements. (orig./HP)

  14. Nondestructive examination of TRIGA reactor fuel elements

    International Nuclear Information System (INIS)

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

  15. Modelling the oxidation of defected fuel elements

    International Nuclear Information System (INIS)

    Interim dry storage of used fuel is an economical alternative to storage in water pools. The fuel must remain intact during the dry-storage period, otherwise future handling of the fuel will be expensive. Oxidation of defected fuel elements can lead to fuel disintegration. Thus it is important to be able to predict the extent of oxidation of defected fuel elements in a dry-storage facility. In this report, a model is developed for predicting the extent or rate of oxidation of defected fuel elements stored at temperatures up to 170 C. The model employs equivalent porous medium representation of the fuel and described the oxygen concentration in the fuel element using a reaction-diffusion equation. The one- and two-dimensional reaction-diffusion equations are solved on the assumption that the oxygen-fuel reaction is either zeroth or first order in the oxygen concentration. Dimensional analysis of the model equations shows that the solution depends explicitly on a single parameter p. The value of p can be calculated using data from the literature, or it can be estimated from the results of the CEX-1 experiments being carried out at Whiteshell Laboratories. The value of p, estimated from the CEX-1 results, is more than two orders of magnitude larger than the value of p calculated from literature data. Although some reasons for this large difference are suggested, further work is needed to resolve this discrepancy. (author). 16 refs., 2 tabs., 11 figs

  16. Combustor having mixing tube bundle with baffle arrangement for directing fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, Michael John; McConnaughhay, Johnie Franklin

    2016-08-23

    A combustor includes a tube bundle that extends radially across at least a portion of the combustor. The tube bundle includes an upstream surface axially separated from a downstream surface, and a plurality of tubes extend from the upstream surface through the downstream surface to provide fluid communication through the tube bundle. A barrier extends radially inside the tube bundle between the upstream and downstream surfaces, and a baffle extends axially inside the tube bundle between the upstream surface and the barrier.

  17. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  18. MRT fuel element inspection at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J.

    1997-08-01

    To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

  19. Method for inspecting nuclear reactor fuel elements

    International Nuclear Information System (INIS)

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

  20. Fundamental aspects of nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D.R.

    1976-01-01

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

  1. The clearance potential index and hazard factors of CANDU fuel bundle and a comparison of experimental-calculated inventories

    International Nuclear Information System (INIS)

    In the field of radioactive waste management, the radiotoxicity can be characterized by two different approaches: 1) IAEA, 2004 RS-G-1.7 clearance concept and 2) US, 10CFR20 radioactivity concentration guides in terms of ingestion / inhalation hazard expressed in m3 of water/air. A comparison between the two existing safety concepts was made in the paper. The modeled case was a CANDU natural uranium, 37 elements fuel bundle with a reference burnup of 685 GJ/kgU (7928.24 MWd/tU). The radiotoxicity of the light nuclide inventories, actinide, and fission-products was calculated in the paper. The calculation was made using the ORIGEN-S from ORIGEN4.4a in conjunction with the activation-burnup library and an updated decay data library with clearance levels data in ORIGEN format produced by WIMS-AECL/SCALENEA-1 code system. Both the radioactivity concentration expressed in Curie and Becquerel, and the clearance index and ingestion / inhalation hazard were calculated for the radionuclides contained in 1 kg of irradiated fuel element at shutdown and for 1, 50, 1500 years cooling time. This study required a complex activity that consisted of various phases such us: the acquisition, setting up, validation and application of procedures, codes and libraries. For the validation phase of the study, the objective was to compare the measured inventories of selected actinide and fission products radionuclides in an element from a Pickering CANDU reactor with inventories predicted using a recent version of the ORIGEN-ARP from SCALE 5 coupled with the time dependent cross sections library, CANDU 28.lib, produced by the sequence SAS2H of SCALE 4.4a. In this way, the procedures, codes and libraries for the characterization of radioactive material in terms of radioactive inventories, clearance, and biological hazard factors are being qualified and validated, in support for the safety management of the radioactive wastes

  2. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

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

  3. 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

  4. Fuel cladding tubes and fuel elements

    International Nuclear Information System (INIS)

    Purpose: To enable non-destructive measurement for the thickness of zirconium barriers. Constitution: Regions capable of non-destructive inspection are provided at the boundary between a fuel cladding tube made of zirconium alloy and the zirconium barrier lined to the inner circumference surface of the tube. As the regions being capable of distinguishing by ultrasonic wave reflection, solid materials, for example, non-metal materials different from that for the tube and the barrier are placed or gaps are provided at the boundary between the zirconium alloy cladding tube and the zirconium barrier. Since ultrasonic waves are reflected at each of the boundaries by the presence of these regions, thickness of the zirconium barrier can be measured in a non-destructive manner from either the inner or the outer surface of the tube. (Yoshino, Y.)

  5. Spring packed particle bed fuel element

    International Nuclear Information System (INIS)

    This patent describes a gas cooled particle bed nuclear fuel element. It comprises: a porous inner frit; a porous outer frit attached to the inner frit by an end cap t a first end and radially guided by a shoulder at a second end, forming an annulus between the frits; a fuel particle bed in the annulus; a first compressive device at each end of the annulus; and a second compressive device positioned in the annulus within the fuel particle bed

  6. HTGR fuel element size reduction system

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-06-01

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

  7. HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

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

  8. Grids for nuclear fuel elements

    International Nuclear Information System (INIS)

    This invention relates to grids for nuclear fuel assemblies with the object of providing an improved grid, tending to have greater strength and tending to offer better location of the fuel pins. It comprises sets of generally parallel strips arranged to intersect to define a structure of cellular form, at least some of the intersections including a strip which is keyed to another strip at more than one point. One type of strip may be dimpled along its length and another type of strip may have slots for keying with the dimples. (Auth.)

  9. Hydraulic modelling of the CARA Fuel element

    International Nuclear Information System (INIS)

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

  10. Neutronic calculations regarding the new LEU 6 x 6 fuel bundle for 14 MW TRIGA - SSR, in order to increase the reactor power up to 21 MW

    Energy Technology Data Exchange (ETDEWEB)

    Iorgulis, C.; Ciocanescu, M.; Preda, M.; Mladin, M. [Institute of Nuclear Research, Pitesti (Romania)

    1998-07-01

    In order to meet the increasing demands of terminal flux for the experimental devices which will be loaded with CANDU natural uranium pins (or clusters), is necessary to rise the reactor power up to 21 MW. In this respect we consider in our evaluations a new 6x6 TRIGA fuel bundle geometry (the actual fuel bundle contains 5x5 pins). This paper will contain a comparative analysis regarding: flux and power distribution across the 29 fuel bundles standard core, and managements patters, in order to maximize the discharge fuel burnup and core lifetime. (author)

  11. TRIGA - LEU cluster with 36 fuel elements

    International Nuclear Information System (INIS)

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

  12. Spacer for fuel rods in nuclear fuel elements

    International Nuclear Information System (INIS)

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

  13. Structural analysis of reactor fuel elements

    International Nuclear Information System (INIS)

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

  14. Structural analysis of reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Weeks, R.W.

    1977-01-01

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

  15. Improved fuel element for fast breeder reactor

    International Nuclear Information System (INIS)

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

  16. Upgraded HFIR Fuel Element Welding System

    Energy Technology Data Exchange (ETDEWEB)

    Sease, John D [ORNL

    2010-02-01

    The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. In recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.

  17. Thermal-hydraulic analysis of flow blockage in a supercritical water-cooled fuel bundle with sub-channel code

    International Nuclear Information System (INIS)

    Highlights: • COBTA-SC code shows good suitability for the blockage analysis of SCWR fuel bundle. • Several thermal-hydraulic models are incorporated and evaluated for the flow blockage of SCWR-FQT bundle. • The axial/circumferential heat conduction of fuel and heat transfer correlation are identified as the important models. • The peak cladding temperature can be reduced effectively by the safety measures of SCWR-FQT. - Abstract: Sub-channel code is nowadays the most applied method for safety analysis and thermal-hydraulic simulation of fuel assembly. It plays an indispensable role to predict the detail thermal-hydraulic behavior of the supercritical water-cooled reactor (SCWR) fuel assembly because of the strong non-uniformity within the fuel bundle. Since the coolant shows a strong variation of physical thermal property near the pseudo critical line, the local blockage in an assembly of a SCWR is of importance to safety analysis. Due to the low specific heat of supercritical water with high temperature, the blockage and the subsequent flow reduction at the downstream of the blockage will yield particular high cladding temperature. To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage caused by detachment of the wire wrap, the sub-channel code COBRA-SC is unitized. The code is validated by some blockage experiments, and it reveals a good feasibility and accuracy for the SCWR and blockage flow analysis. Some new models, e.g. the axial and circumferential heat conduction model, turbulent mixing models, pressure friction models and heat transfer correlations, are incorporated in COBRA-SC code. And their influence on the cladding temperature and mass flow distribution are evaluated and discussed. Based on the results, the appropriate models for description of the flow blockage phenomenon in SCWR assembly is identified and recommended. A transient analysis of the

  18. Nuclear fuel element with a bond coating

    International Nuclear Information System (INIS)

    The possibility of undesired interactions between the pellets (of UO2 or a mixture of UO2 + PuO2) and the cladding which can cause stress crack corrosion, are to be excluded in particular in the proposed fuel element. The container enclosing the fuel consists according to the invention of a zirconium alloy having a zirconium oxide diffusion barrier on the side facing the fuel and a metal coating on top of this. Cu is best suited, but Ni, Fe or their alloys are named. The treatment of the surfaces to simplify the coating of the individual layers is described. (UWI) 891 HP/UWI 892 CKA

  19. Integrated Planar Solid Oxide Fuel Cell: Steady-State Model of a Bundle and Validation through Single Tube Experimental Data

    Directory of Open Access Journals (Sweden)

    Paola Costamagna

    2015-11-01

    Full Text Available This work focuses on a steady-state model developed for an integrated planar solid oxide fuel cell (IP-SOFC bundle. In this geometry, several single IP-SOFCs are deposited on a tube and electrically connected in series through interconnections. Then, several tubes are coupled to one another to form a full-sized bundle. A previously-developed and validated electrochemical model is the basis for the development of the tube model, taking into account in detail the presence of active cells, interconnections and dead areas. Mass and energy balance equations are written for the IP-SOFC tube, in the classical form adopted for chemical reactors. Based on the single tube model, a bundle model is developed. Model validation is presented based on single tube current-voltage (I-V experimental data obtained in a wide range of experimental conditions, i.e., at different temperatures and for different H2/CO/CO2/CH4/H2O/N2 mixtures as the fuel feedstock. The error of the simulation results versus I-V experimental data is less than 1% in most cases, and it grows to a value of 8% only in one case, which is discussed in detail. Finally, we report model predictions of the current density distribution and temperature distribution in a bundle, the latter being a key aspect in view of the mechanical integrity of the IP-SOFC structure.

  20. 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.

  1. Simulation of the fuel rod bundle test QUENCH-03 using the integral code ASTEC V2

    Energy Technology Data Exchange (ETDEWEB)

    Kruse, Philipp; Koch, Marco K. [Bochum Univ. (Germany). Chair of Energy Systems and Energy Economics

    2010-05-15

    Failure of the main and emergency cooling-systems can lead to an accident with severe core degradation even with core meltdown. To prevent total meltdown of the uncovered and overheated core reflooding with water is an unavoidable accident management measure. The fast supply of water and the resulting increased available amount of steam can lead to crack formations and break up of the oxide layer of the fuel rods. The additionally exposed surface could result in an increased release of hydrogen due to a supplementary exothermal zirconium- steam-oxidation reaction. Within the frame of the QUENCH test-program - realised by FZK - loss of coolant accidents in LWR (Light Water Reactor) are analysed using an experimental reactor core to determine the produced amount of hydrogen, the so-called hydrogen source term. Additionally, the behaviour of the bundle with different absorber rod and cladding materials is being analysed. Based on the post-test calculations of the QUENCH tests with the severe accident code system ASTEC the capability of the code can be established and evaluated. In the following the post-test calculations of the QUENCH-03 test with ASTEC V2 are discussed. (orig.)

  2. Ultrasonic systems for high-accuracy thickness measurement of fuel bundle bearing pads and shield plug crimps

    International Nuclear Information System (INIS)

    The performance of two ultrasonic systems, remotely operated in high radiation environment, are presented. The first system is used to measure the bearing pad height of radioactive fuel bundles located in the irradiated fuel bays, at Darlington NGS. The system was designed and commissioned to achieve an accuracy of ± 20 μm. The repeatability of results is within ± 10 μm uniformity band. The measurements are independent of testing speed, water temperature, bundle temperature, pencil geometry. Possibilities and limitations of the UT system are also presented and some improved alternatives are proposed. The second system was developed for measuring the crimp height of shield plugs (special iron casting) at Bruce B - Mark Ill development. The accuracy of measurements is ± 50 μm, with a repeatability of ± 25 μm. The results are independent of shield plug thickness variation and ovality, crimp off-set and heavy-water temperature. (author)

  3. Determination of transient radial-azimuthal temperature distributions in fuel bundles under loss-of-coolant-accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Saltos, N.T.; Christensen, R.N.; Aldemir, T.

    1988-10-01

    A methodology is presented to determine the transient temperature distributions in fuel bundles under loss-of-coolant accident (LOCA) conditions using a recently developed variational technique for the solution of radial-azimuthal heat conduction in the fuel rods and the modified view factor concept proposed by Uchida and Nakamure to model the radiative heat transfer between the rods. The variational technique is based on the Lebron-Labermont restricted variational principle and represents the temperature distribution in the rods at a given time during the LOCA via parabolic and circular trial functions in the radial and azimuthal directions, respectively. The methodology is implemented to a 4 x 4 boiling water reactor fuel bundle under typical LOCA conditions to investigate the effects of changes in rod heat transfer characteristics and simplifying modeling assumptions on predicted rod temperature distributions. The results show that these effects depend on the rod location in the assembly and LOCA phase under consideration and indicate that same degree of modelling detail may not be necessary for all the rods in the bundle at all times during the LOCA.

  4. Catalogue of fuel elements - 1. addendum October 1958

    International Nuclear Information System (INIS)

    This document contains sheets presenting various characteristics of nuclear fuel elements which are distinguished with respect to their shape: cylinder bar, plate, tube. Each sheet comprises an indication of the atomic pile in which the fuel element is used, dimensions, cartridge data, data related to cooling, to combustion rate, and to fuel handling. A drawing of the fuel element is also given

  5. The 24 CANFLEX-NU bundle demonstration irradiation at Wolsong-1 generating station-bundle manufacture and QA, fuel handling aspects, flasking and shipping and pie for the irradiated fuel, and follow-up documentation

    International Nuclear Information System (INIS)

    Korea Ministry of Science and Technology(MOST) has pushed and given a financial support to a KEPRI/KAERI Joint Industrialization Program of CANFLEX-NU Fuel as one of Korea's National Nuclear Mid- and Long Term R and D Program. The Industrialization Program will be conducted for 3 years from 2000 November to efficiently utilize the CANFLEX fuel technology developed by KAERI and AECL jointly, where the KAERI's works have been conducted under the Korea's national program of the mid- and long-term nuclear R and D programs since 1992. This document is a report to guideline the following activities on the safety assessment for the 24 CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel bundle demonstration irradiation at Wolsong-1 Generating Station: 'bundle manufacture and QA', 'Fuel handling aspects such as loading fuel, de-fuelling and segregation, and visual in-bay examinations', 'Flasking and shipping', 'Post-irradiation examination', and 'Follow-up documentation to be produced'

  6. 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

  7. Development of computational technology on heat transfer and fluid flow in a nuclear fuel bundle of advanced reactor

    International Nuclear Information System (INIS)

    The assessment of the RANS(Reynolds-Averaged Navier-Stokes) based turbulence model was conducted to establish the optimal CFD system for turbulent flow and heat transfer in reactor during the first year of the project. The RANS models used in this project are the two-equation models based on the eddy viscosity assumption and the Second-Moment Closure(SMC) models. Since the nuclear fuel assembly loaded in the nuclear reactor is a rod bundle which is square or triangular array, the predictions using the various turbulence models were compared for turbulent flow in bare square and/or triangular rod bundle and the rod bundle with the flow mixing vane. The study for the second year of the project examined the CFD model and the applicability of the CFD code for the turbulent two-phase flow. The numerical predictions of lateral distributions of void fraction, phasic velocities and turbulent kinetic energy were compared against the experimental results for upward and downward bubbly flow in a vertical tube. The boiling flows in vertical tube and rod bundle were also simulated to verify the CFD results

  8. OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

    Directory of Open Access Journals (Sweden)

    A. Del Nevo

    2012-01-01

    Full Text Available Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters. One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT conducted by the Nuclear Power Engineering Corporation (NUPEC in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions.

  9. Laser assisted decontamination of nuclear fuel elements

    International Nuclear Information System (INIS)

    Laser assisted removal of loosely bound fuel particulates from the clad surface following the process of pellet loading has decided advantages over conventional methods. It is a dry and noncontact process that generates very little secondary waste and can occur inside a glove box without any manual interference minimizing the possibility of exposure to personnel. The rapid rise of the substrate/ particulate temperature owing to the absorption of energy from the incident laser pulse results in a variety of processes that may lead to the expulsion of the particulates. As a precursor to the cleaning of the fuel elements, initial experiments were carried out on contamination simulated on commonly used clad surfaces to gain a first hand experience on the various laser parameters for which as efficient cleaning can be obtained without altering the properties of the clad surface. The cleaning of a dummy fuel element was subsequently achieved in the laboratory by integrating the laser with a work station that imparted simultaneous rotational and linear motion to the fuel element. (author)

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

    International Nuclear Information System (INIS)

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

  11. Nuclear criticality assessment of LEU and HEU fuel element storage

    International Nuclear Information System (INIS)

    Criticality aspects of storing LEU (20%) and HEU (93%) fuel elements have been evaluated as a function of 235U loading, element geometry, and fuel type. Silicide, oxide, and aluminide fuel types have been evaluated ranging in 235U loading from 180 to 620 g per element and from 16 to 23 plates per element. Storage geometry considerations have been evaluated for fuel element separations ranging from closely packed formations to spacings of several centimeters between elements. Data are presented in a form in which interpolations may be made to estimate the eigenvalue of any fuel element storage configuration that is within the range of the data. (author)

  12. 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.

  13. Nuclear fuel element having oxidation resistant cladding

    International Nuclear Information System (INIS)

    This patent describes an improved nuclear fuel element of the type including a zirconium alloy tube, a zirconium barrier layer metallurgically bonded to the inside surface of the alloy tube, and a central core of nuclear fuel material partially filling the inside of the tube so as to leave a gap between the sponge zirconium barrier and the nuclear fuel material. The improvement comprising an alloy layer formed on the inside surface of the zirconium barrier layer. The alloy layer being composed of one or more impurities present in a thin layer region of the zirconium barrier in amounts less than 1% by weight but sufficient to inhibit the oxidation of the inside surface of the zirconium barrier layer without substantially affecting the plastic properties of the barrier layer, wherein the impurities are selected from the group consisting of iron, chromium, copper, nitrogen, and niobium

  14. Investigations of flow and temperature field development in bare and wire-wrapped reactor fuel pin bundles cooled by sodium

    International Nuclear Information System (INIS)

    Highlights: ► We study sodium flow and temperature development in fuel pin bundles. ► Pin diameter, number of pins, wire wrap and ligament gap are varied as parameters. ► Flow development is achieved within ∼30–40 hydraulic diameters. ► Thermal development is attained only for small pin diameter and less number of pins. ► Wire wrap and ligament gap strongly influence Nusselt number. - Abstract: Simultaneous development of liquid sodium flow and temperature fields in the heat generating pin bundles of reactor has been investigated. Development characteristics are seen to be strongly influenced by pin diameter, number of pins, helical wire-wrap, ligament gap between the last row of pins and hexcan wall and Reynolds number. Flow development is achieved within an axial length of ∼125 hydraulic diameters, for all the pin bundle configurations considered. But temperature development is attained only if the pin diameter is small or the number of pins is less. In the case of large pin diameter with more pins, temperature development could not be achieved even after a length of ∼1000 hydraulic diameters. The reason for this behavior is traced to be the weak communication among sub-channels in tightly packed bundles. It is seen that the pin Nusselt number decreases from center to periphery in a bundle. Also, if the ligament gap is narrow, the Nusselt number is large and more uniform. Flow development length is short if the Reynolds number is large and the converse is true for thermal development length. Helical wire-wrap shortens the thermal entry length and significantly enhances the global Nusselt number. But, its influence on hydrodynamic entry length is not significant

  15. Feasibility evaluation of x-ray imaging for measurement of fuel rod bowing in CFTL test bundles

    International Nuclear Information System (INIS)

    The Core Flow Test Loop (CFTL) is a high temperature, high pressure, out-of-reactor helium-circulating system. It is designed for detailed study of the thermomechanical performance, at prototypic steady-state and transient operating conditions, of electrically heated rods that simulate segments of core assemblies in the Gas-Cooled Fast Breeder reactor demonstration plant. Results are presented of a feasibility evaluation of x-ray imaging for making measurements of the displacement (bowing) of fuel rods in CFTL test bundles containing electrically heated rods. A mock-up of a representative CFTL test section consisting of a test bundle and associated piping was fabricated to assist in this evaluation

  16. SEU blending project, concept to commercial operation, Part 3: production of powder for demonstration irradiation fuel bundles

    International Nuclear Information System (INIS)

    The processes for production of Slightly Enriched Uranium (SEU) dioxide powder and Blended Dysprosium and Uranium (BDU) oxide powder that were developed at laboratory scale at Cameco Technology Development (CTD), were implemented and further optimized to supply to Zircatec Precision Industries (ZPI) the quantities required for manufacturing twenty six Low Void Reactivity (LVRF) CANFLEX fuel bundles. The production of this new fuel was a challenge for CTD and involved significant amount of work to prepare and review documentation, develop and approve new analytical procedures, and go through numerous internal reviews and audits by Bruce Power, CNSC and third parties independent consultants that verified the process and product quality. The audits were conducted by Quality Assurance specialists as well as by Human Factor Engineering experts with the objective to systematically address the role of human errors in the manufacturing of New Fuel and confirm whether or not a credible basis had been established for preventing human errors. The project team successfully passed through these audits. The project management structure that was established during the SEU and BDU blending process development, which included a cross-functional project team from several departments within Cameco, maintained its functionality when Cameco Technology Development was producing the powder for manufacturing Demonstration Irradiation fuel bundles. Special emphasis was placed on the consistency of operating steps and product quality certification, independent quality surveillance, materials segregation protocol, enhanced safety requirements, and accurate uranium accountability. (author)

  17. Packaging of spent fuel elements into special containers

    International Nuclear Information System (INIS)

    This report contains detailed description of the procedure for packaging the spent fuel elements from the fuel channels into the special steel containers. The previously cooled fuel elements are packaged into containers by the existing crane and transported later into the spen fuel storage. Instructions for crane operation are included

  18. Heat transfer profiles of a vertical, bare, 7-element bundle cooled with supercritical Freon-12

    International Nuclear Information System (INIS)

    Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. However, SuperCritical Water-cooled nuclear Reactor (SCWRs) designs cannot be optimized without such data. A set of experimental data obtained in Freon-12 (modeling fluid) cooled vertical bare bundle at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia) was analyzed. The existence of three distinct regimes for forced convention with supercritical fluids was experienced. (1) Normal heat transfer; (2) Deteriorated heat transfer, characterized by higher than expected temperatures; and (3) Enhanced heat transfer, characterized by lower than expected temperatures. This work compares the heat transfer coefficient of the experiments to predictions based upon current correlations for heat transfer in super critical fluids where the 1-D correlations are based upon tube data under supercritical water conditions. (author)

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

    International Nuclear Information System (INIS)

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

  20. Searching for a possible fuel element leak

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-01-15

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

  2. Some aspects on security and safety in a potential transport of a CANDU spent nuclear fuel bundle, in Romania

    International Nuclear Information System (INIS)

    Each Member States (MS) is responsible for the security and safety of radioactive material during transport, since radioactive material is most vulnerable during transport. The paper presents some aspects on security and safety related to the potential transport of a CANDU Spent Nuclear Fuel (SNF) bundle from NPP CANDU Cernavoda to INR Pitesti. The possible environmental impact and radiological consequences following a potential event during transportation is analyzed, since the protection of the people and the environment is the essential goal to be achieved. Some testing for the package to be used for transportation will be also given. (author)

  3. Neutron induced activity in fuel element components

    International Nuclear Information System (INIS)

    A thorough investigation of the importance of various nuclides in neutron-induced radioactivity from fuel element construction materials has been carried out for both BWR and PWR fuel assemblies. The calculations were performed with the ORIGEN computer code. The investigation was directed towards the final storage of the assembly components and special emphasis was put to the examination of the sources of carbon-14, cobalt-60, nickel-59, nickel-63 and zirconium-93/niobium-93m. It is demonstrated that the nuclides nickel-59, in Inconel and stainless steel, and zirconium-93/niobium-93m, in Zircaloy, are the ones which constitute the very long term radiotoxic hazard of the irradiated materials. (author)

  4. Thermionic fuel element verification program—overview

    Science.gov (United States)

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

    1991-01-01

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

  5. Thermionic fuel element Verification Program - Overview

    Science.gov (United States)

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

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

  6. Fuel element storage pond for nuclear installations

    International Nuclear Information System (INIS)

    In a fuel element storage pond for nuclear installations, with different water levels, radioactive particles are deposited at the points of contact of the water surface with the pond wall. So that this deposition will not occur, a metal apron is provided in the area of the points of contact of the water surface with the bond wall. The metal apron consists of individual sheets of metal which are suspended by claws in wall hooks. To clean the sheets, these are moved to a position below the water level. The sheets are suspended from the wall hooks during this process. (orig.)

  7. A prediction method of the effect of radial heat flux distribution on critical heat flux in CANDU fuel bundles

    International Nuclear Information System (INIS)

    Fuel irradiation experiments to study fuel behaviors have been performed in the experimental loops of the National Research Universal (NRU) Reactor at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories (CRL) in support of the development of new fuel technologies. Before initiating a fuel irradiation experiment, the experimental proposal must be approved to ensure that the test fuel strings put into the NRU loops meet safety margin requirements in critical heat flux (CHF). The fuel strings in irradiation experiments can have varying degrees of fuel enrichment and burnup, resulting in large variations in radial heat flux distribution (RFD). CHF experiments performed in Freon flow at CRL for full-scale bundle strings with a number of RFDs showed a strong effect of RFD on CHF. A prediction method was derived based on experimental CHF data to account for the RFD effect on CHF. It provides good CHF predictions for various RFDs as compared to the data. However, the range of the tested RFDs in the CHF experiments is not as wide as that required in the fuel irradiation experiments. The applicability of the prediction method needs to be examined for the RFDs beyond the range tested by the CHF experiments. The Canadian subchannel code ASSERT-PV was employed to simulate the CHF behavior for RFDs that would be encountered in fuel irradiation experiments. The CHF predictions using the derived method were compared with the ASSERT simulations. It was observed that the CHF predictions agree well with the ASSERT simulations in terms of CHF, confirming the applicability of the prediction method in fuel irradiation experiments. (author)

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

    International Nuclear Information System (INIS)

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

  9. UNIFRAME interim design report. [Fuel element size reduction plant

    Energy Technology Data Exchange (ETDEWEB)

    Strand, J.B.; Baer, J.W.; Cook, E.J.

    1977-12-01

    A fuel element size reduction system has been designed for the ''cold'' pilot-scale plant for an HTGR Fuel Reference Recycle Facility. This report describes in detail the present design.

  10. Testing device for fuel element samples

    International Nuclear Information System (INIS)

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

  11. Crystal structure of the GTPase domain and the bundle signalling element of dynamin in the GDP state.

    Science.gov (United States)

    Anand, Roopsee; Eschenburg, Susanne; Reubold, Thomas F

    2016-01-01

    Dynamin is the prototype of a family of large multi-domain GTPases. The 100 kDa protein is a key player in clathrin-mediated endocytosis, where it cleaves off vesicles from membranes using the energy from GTP hydrolysis. We have solved the high resolution crystal structure of a fusion protein of the GTPase domain and the bundle signalling element (BSE) of dynamin 1 liganded with GDP. The structure provides a hitherto missing snapshot of the GDP state of the hydrolytic cycle of dynamin and reveals how the switch I region moves away from the active site after GTP hydrolysis and release of inorganic phosphate. Comparing our structure of the GDP state with the known structures of the GTP state, the transition state and the nucleotide-free state of dynamin 1 we describe the structural changes through the hydrolytic cycle.

  12. Fabrication technology of spherical fuel element for HTR-10

    International Nuclear Information System (INIS)

    R and D on the fabrication technology of the spherical fuel elements for the 10 MW HTR Test Module (HTR-10) began from 1986. Cold quasi-isostatic molding with a silicon rubber die is used for manufacturing the spherical fuel elements.The fabrication technology and the graphite matrix materials were investigated and optimized. Twenty five batches of fuel elements, about 11000 of the fuel elements, have been produced. The cold properties of the graphite matrix materials satisfied the design specifications. The mean free uranium fraction of 25 batches was 5 x 10-5

  13. Fuel Element Transfer Cask Modelling Using MCNP Technique

    Science.gov (United States)

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-01

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  14. Reproduction of the RA reactor fuel element fabrication

    International Nuclear Information System (INIS)

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

  15. Subchannel Analysis for enhancing the fuel performance in CANDU reactor

    International Nuclear Information System (INIS)

    The effect of the fuel rod geometry in a fuel bundle using the subchannel code ASSERT has been evaluated to design the fuel bundle having the advanced fuel performance. Based on the configuration of standard 37-element fuel bundle, the element diameter of fuel rods in each ring has been changed while that of fuel rods in other rings is kept as the original size. The dryout power of each element in a fuel bundle has been obtained for the modified fuel bundle and compared with that of a standard fuel bundle. From the calculated mixture enthalpy and void fraction of each subchannel, it was found that the modification of element diameter largely affects to the thermal characteristics of the subchannel on the upper region of a modified element by the buoyancy drift effect. The optimized geometry in a fuel bundle has been suggested from the consideration of the change of void reactivity as well as the dryout power of a bundle. The dependency of the transverse interchange model on the present results has been checked by examining the dryout power of a bundle for the different mixing coefficient and buoyancy drift model

  16. Simultaneous development of flow and temperature fields in wire-wrapped fuel pin bundles of sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Simultaneous development of flow and temperature fields in the entrance region of fast breeder reactor (FBR) fuel pin bundles with helical spacer wires has been investigated by three-dimensional computational simulations. The Reynolds number, pitch of helical spacer wire and number of pins in the bundle are systematically varied. It is found that the magnitude of mean cross-stream velocity in the fully developed region is inversely proportional to the helical pitch length and it is nearly independent of the number of pins. But, there is a strong correlation between the locations of spacer wire and the peak cross-stream velocity. Flow attains full development at an axial length of 70 times hydraulic diameter in all the cases and this length is found to be unaffected by the helical pitch length. Friction factor is seen to fluctuate periodically over a mean value and the fluctuation over each helical pitch corresponds to a specific position of helical wire. The mean value of the friction factor in the entrance region reduces below the mean value in the fully developed region contrary to that seen in ducted flows. The mean fully developed friction factor is inversely proportional to the helical pitch. But, it is independent of the number of pins in the bundle. The Nusselt number passes through multiple minima before attaining fully developed periodic fluctuations and its development is slower than that of friction factor. For larger number of pins thermal development length is larger. Traditionally, the correlations reported for fully developed flow are considered for core design. But, the present study indicates that this approach is not conservative. Further, the entrance region effects and the oscillations in the fully developed region have to be properly accounted in the core design. Nusselt number exhibits a strong dependence on helical pitch similar to that of friction factor. A correlation for Nusselt number is proposed as a function of helical pitch and other

  17. Legal questions concerning the termination of spent fuel element reprocessing

    International Nuclear Information System (INIS)

    The thesis on legal aspects of the terminated spent fuel reprocessing in Germany is based on the legislation, jurisdiction and literature until January 2004. The five chapters cover the following topics: description of the problem; reprocessing of spent fuel elements in foreign countries - practical and legal aspects; operators' responsibilities according to the atomic law with respect to the reprocessing of Geman spent fuel elements in foreign countries; compatibility of the prohibition of Geman spent fuel element reprocessing in foreign countries with international law, European law and German constitutional law; results of the evaluation

  18. Attempt to produce silicide fuel elements in Indonesia

    International Nuclear Information System (INIS)

    After the successful experiment to produce U3Si2 powder and U3Si2-Al fuel plates using depleted U and Si of semiconductor quality, silicide fuel was synthesized using x-Al available at the Fuel Element Production Installation (FEPI) at Serpong, Indonesia. Two full-size U3Si2-Al fuel elements, having similar specifications to the ones of U3O8-Al for the RSG-GAS (formerly known as MPR-30), have been produced at the FEPI. All quality controls required have been imposed to the feeds, intermediate, as well as final products throughout the production processes of the two fuel elements. The current results show that these fuel elements are qualified from fabrication point of view, therefore it is expected that they will be permitted to be tested in the RSG-GAS, sometime by the end of 1989, for normal (∝50%) and above normal burn-up. (orig.)

  19. CFD activities in support of thermal-hydraulic modeling of SFR fuel bundles

    International Nuclear Information System (INIS)

    Extensive testing and validation work is being performed to assess and validate Computational Fluid Dynamics (CFD) applicability to the simulation of SFR fuel assemblies. The demonstrated robustness of the method allows extending the CFD analysis to distorted fuel configurations, which will inevitably occur during extended fuel operation. The subchannel code COBRA-IV-I-MIT is adopted to evaluate the range of applicability of lumped parameter methods. Comparisons of mixing simulations show some intrinsic limitation in the subchannel methods, but allow confirming its overall applicability to nominal and mildly deformed assembly configurations. For significantly deformed geometries CFD is the recommend approach and is applied in this work. Deformed geometries considered include duct swelling, rod swelling, rod bowing, rod twisting, and various combinations of the simple deformations. While not derived from the realistic analysis of the in-core fuel behavior, the distorted geometries have been designed to embrace all conceptual worst case scenarios. The work focuses on the evaluation of the influence of the deformation on the fuel behavior, rather than on the actual fuel performance. Such approach is driven by the objective of deriving general understanding, and evaluating the applicability of subchannel analysis codes to long life fuel design, possibly in combination with distorted-channel factors derived from the CFD analyses. (author)

  20. Simulation of hemp fibre bundle and cores using discrete element method

    Energy Technology Data Exchange (ETDEWEB)

    Al-Amin Sadek, M.; Chen, Y. [Manitoba Univ., Winnipeg, MB (Canada). Dept. of Biosystems Engineering; Lague, C. [Ottawa Univ., Ottawa, ON (Canada). Faculty of Engineering; Landry, H. [Prairie Agricultural Machinery Inst., Humboldt, SK (Canada); Peng, Q. [Manitoba Univ., Winnipeg, MB (Canada). Dept. of Mechanical and Manufacturing Engineering; Zhong, W. [Manitoba Univ., Winnipeg, MB (Canada). Dept. of Textile Sciences

    2010-07-01

    The mechanical behaviour of hemp fibre and core must be well understood in order to obtain high-grade hemp fibre that is currently in high demand for various industrial applications. Modelling by discrete element method can simulate the mechanical behaviour of such materials. A commercial discrete element software called Particle Flow Code was used in this study. In particular, the 3-dimension (PFC3D) was used to simulate hemp fibre and core. Since the basic PFC3D particles are spherical, the individual virtual hemp fibres were defined as strings of balls held together by PFC3D parallel bonds. The study showed that the virtual fibre is flexible and can bend and break by forces. This reflects the characteristics of hemp fibre. Using the clump logic of PFC3D, the virtual hemp core was defined as a rigid and unbreakable body, which reflect the characteristics of the core. The virtual fibre and core were defined with several microproperties, some of which were previously calibrated. The PFC3D bond properties were calibrated in this study. They included normal and shear stiffness; pb{sub k}n and pb{sub k}s; normal and shear strength; and bond disk radius, R of the virtual fibre. The calibration started with developing a PFC3D model to simulate fibre tensile test. The microproperties of virtual fibre and core were calibrated by running the PFC3D model. Literature data from fibre tensile tests was compared with simulation results.

  1. Development of neural network for predicting local power distributions in BWR fuel bundles considering burnable neutron absorber

    International Nuclear Information System (INIS)

    A neural network model is under development to predict the local power distribution in a BWR fuel bundle as a high speed simulator of precise nuclear physical analysis model. The relation between 235U enrichment of fuel rods and local peaking factor (LPF) has been learned using a two-layered neural network model ENET. The training signals used were 33 patterns having considered a line symmetry of a 8x8 assembly lattice including 4 water rods. The ENET model is used in the first stage and a new model GNET which learns the change of LPFs caused by burnable neutron absorber Gadolinia, is added to the ENET in the second stage. Using this two-staged model EGNET, total number of training signals can be decreased to 99. These training signals are for zero-burnup cases. The effect of Gadolinia on LPF has a large nonlinearity and the GNET should have three layers. This combined model of EGNET can predict the training signals within 0.02 of LPF error, and the LPF of a high power rod is predictable within 0.03 error for Gadolinia rod distributions different from the training signals when the number of Gadolinia rods is less than 10. The computing speed of EGNET is more than 100 times faster than that of a precise nuclear analysis model, and EGNET is suitable for scoping survey analysis. (author)

  2. Transient non-boiling heat transfer in a fuel rod bundle during accidental power excursions

    International Nuclear Information System (INIS)

    The physical problem studied is the transient non-boiling heat transfer of a cylindrical fuel rod consisting of fuel, gap, and cladding to a steady, fully developed turbulent flow. The fuel pin is assumed to be located in the interior region of a subassembly with regular triangular or square arrangements. The turbulent velocity field as well as turbulent transport properties are specified as functions of the coordinates normal to the axial flow direction. The heat generation within the fuel may be specified as an arbitrary function of the three spatial coordinates and time. A digital computer program has been developed. On the basis of finite-difference techniques, to solve the governing partial differential equations with their associated subsidiary conditions. Results have been obtained for a series of exponential power transients of interest to safety of liquid-metal and water cooled nuclear reactors. The general physical features of transient convective heat transfer as explored by previous investigators have qualitatively been substantiated by the present analysis. Emphasis has been devoted to investigate the differences of heat-transfer (coefficient) results from multi-region analysis including a realistic fuel rod model and single-region analysis for the coolant region only. A comparison with the engineering relationships for turbulent liquid-metal cooling by Stein, which are an extension of the heat transfer coefficient concept to account for transient heat fluxes, clearly demonstrates that, at the parameters studied, Stein's approach tends to largely overestimate the convective heat transfer at early times

  3. Nonlinear transient deformation of LMFBR fuel elements under impulsive loading

    International Nuclear Information System (INIS)

    Hypothetical reactor accidents are characterized by a sudden release of substantial thermal energy in one fuel element. Presently it cannot be excluded that for instance pressure pulses due to a fuel coolant interaction may have such time scales and impulses as to deform neighboring subassemblies permanently. Additionally coherent fuel element motion may limit control rod scram action and possibly cause untolerable reactivity increases. Therefore LMFBR safety requires to analyse the complex mechanical response of the core structure under typical loading conditions. An important contribution to this problem is to examine the nonlinear structural dynamics of an individual fuel element under prescribed loading and boundary conditions. The subject of this paper is the elastoplastic transient behaviour of one subassembly under given space-and-time dependent pressure loading. The interaction of several colliding fuel elements including coolant dynamics is briefly discussed. (Auth.)

  4. The International Marketing Target of Fuel Element for Research Rectors

    International Nuclear Information System (INIS)

    The International marketing efforts of PT BATAN Teknologi's fuel element for research reactors are out line. These efforts intensively started in third year marketing time since it is commenced on 24 May 1996. The market segmentation told that there are 269 research reactors in the world, I.e. 65 in USA, 27 in Russia, 18 in Japan, and the remaining are in many Countries. Many of those are 78 swimming fool type reactors, and 17 of them, I.e. 4 in Japan, 4 in USA, and each Austria, Germany, Argentina, Iran, Pakistan, Peru, Brazil, Algeria and Indonesia have the similar fuel element specifications with are close related with PT BATAN Teknologi's. It can be predicated that around 38 fuel elements and 84 fuel control can be marketed. The first feasibility study told that for countries such as Peru, Pakistan, Iran, Algeria, became the potential marketing target of the BATAN Teknologi's fuel element, because for those countries the competitors in producing such fuel elements could be minimal. The fuel elements and fuel control which could be presumably marketed in those countries are 83 and 19 respectively. The problem will be facing in near future such as packaging design and nuclear fuel transportation have to be firstly solved by collaborating with foreign companies abroad. Non technical problems including political situation have to be completely studied in order the uranium, transfer to many countries for exporting purpose could easily take place in the future. The government of the Republic of Indonesia (in this case BATAN) and the International Atomic Energy Agency (IAEA) could assist to solve the non technical problems which might be appear in the future as the chance of the exporting the fuel elements and the fuel controls come true. (author)

  5. Experience with TRIGA aluminum-clad fuel elements

    International Nuclear Information System (INIS)

    During 8 years of operation the cumulative heat energy produced in the steady-state TRIGA Mark II 250 kW reactor at Ljubljana reached 4683 MWh. The initial core had Al-clad fuel elements only. The reactivity loss due to the burnup has been compensated by fresh fuel elements with SS-cladding and, lately, by FLIP fuel elements, moving the most irradiated Al-clad fuel elements from B and C rings to the F ring and, lately, to the storage rack. The inspection of the fuel elements during the summer of 1973 revealed excessive elongations of some Al-clad fuel elements, up to 36.8 mm. By the neutronography, performed by indirect methods (In, Dy), and also by direct methods (track detector CA 80-15 B) and by special radiographic procedures on the element, the activity of which decayed sufficiently, it has been demonstrated that the growth is due to the elongation of aluminum cladding only. No growth and/or swelling of the ZrH--U fuel or the graphite plugs has been observed within the accuracy of detection. (U.S.)

  6. The button effect of CANFLEX bundle on the critical heat flux and critical channel power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E.; Inch, W. [Atomic Energy of Canada Limited, Ontario (Canada)

    1997-12-31

    A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

  7. Hermetic seal process for nuclear fuel element

    International Nuclear Information System (INIS)

    The welding of the end plug onto the sheath of the fuel rod is made inside an enclosure filled with inert gas under the same pressure at that needed inside the fuel rod. The welding can be a tungsten arc welding, a laser welding or a micro plasma welding

  8. Structural dynamics studies for single and clustered SNR-300 fuel elements: a comparison of analytical and experimental results

    International Nuclear Information System (INIS)

    Structural damage may occur in liquid metal cooled fast breeder reactors (LMFBR) when local failures escalate to a fast release of excessive thermal energy within one fuel element. Time scales (5 to 50. msec) and peaks of associated pressure pulses (50. to 500. bar) may exceed the burst pressure of the hexagonal wrapper containing the pin-bundle. Though after fuel rupture the pressure level in the vicinity is markedly reduced, still a significant pressure field is propagating through the core. Therefore surrounding subassemblies or control rods are exposed to impulsive differential pressures causing duct flexure as well as local cross section deformation. A theoretical and experimental research program at GfK Karlsruhe investigates the mechanical effects of conservative pressure transients on LMFBR fuel elements. As part of the program this paper describes two computer codes (BEDYN-2, CORE-1) for non-linear structural dynamics of single as well as clustered subassemblies. (Auth.)

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

    International Nuclear Information System (INIS)

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

  10. Repurposing an irradiated instrumented TRIGA fuel element for regular use

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Paulo F.; Souza, Luiz C.A., E-mail: pfo@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    TRIGA IPR-R1 is a research reactor also used for training and radioisotope production, located at the Centro de Desenvolvimento da Tecnologia Nuclear da Comissao Nacional de Energia Nuclear (Nuclear Technology Development Centre, Brazilian National Nuclear Energy Commission - CDTN/CNEN). Its first criticality occurred in November 1960. All original fuel elements were aluminum-clad. In 1971 nine new fuel elements, stainless steel-clad were acquired. One of them was an instrumented fuel element (IFE), equipped with 3 thermocouples. The IFE was introduced into the core only on August 2004, and remained there until July 2007. It was removed from the core after the severing of contacts between the thermocouples and their extension cables. After an unsuccessful attempt to recover electrical access to the thermocouples the IFE was transferred from the reactor pool to an auxiliary spent fuel storage well, with water, in the reactor room. In December 2011 the IFE was transferred to an identical well, dry, where it remains so far. This work is a proposal for recovery of this instrumented fuel element, by removing the cable guide rod and adaptation of a superior terminal plug similar to conventional fuel elements. This will enable its handling through the same tool used for regular fuel elements and its return to the reactor core. This is a delicate intervention in terms of radiological protection, and will require special care to minimize the exposure of operators. (author)

  11. Container for transport of radioactive fuel elements

    International Nuclear Information System (INIS)

    Five or six fuel assemblies may directly be inserted into the bearing cage placed in the storage pool. Later, after decay, it will be possible to put the bearing cage containing the fuel assemblies into the shipping cask for the reprocessing plant. The shipping cask has got a cover filled up for the transport with a sealing compound consisting of salt, a mixture of salt, or bitumen. The wall of the shipping cask has got a sandwich structure. (DG)

  12. Design and Testing of Prototypic Elements Containing Monolithic Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

  13. Thermomechanical analysis of nuclear fuel elements

    International Nuclear Information System (INIS)

    This work presents development of a code to obtain the thermomechanical analysis of fuel rods in the fuel assemblies inserted in the core of BWR reactors. The code uses experimental correlations developed in several laboratories. The development of the code is divided in two parts: a) the thermal part and b) the mechanical part, extending both the fuel and the cladding materials. The thermal part consists of finding the radial distribution of temperatures in the pellet, from the fuel centerline up to the coolant, along the total active length, considering one and two phase flow in the coolant, as a result of the pressure drop in the system. The mechanical part analyzes the effects of temperature gradients, pressure and irradiation, to which the fuel rod is subjected. The strains produced by swelling, creep and thermal stress in the fuel material are analyzed. In the same way the strains in the cladding are analyzed, considering the effects produced by the pressure exerted on the cladding by pellet swelling, by the pressure caused by fission gas release toward the cavities, and by the strain produced on the cladding by the pressure changes of the system. (Author)

  14. Manufacture of nuclear fuel elements for commercial PWR in China

    International Nuclear Information System (INIS)

    Yibin Nuclear Fuel Element Plant (YFP) under the leadership of China National Nuclear Corporation is sole manufacturer in China to specialize in the production of fuel assemblies and associated core components for commercial PWR nuclear power plant. At the early of 1980's, it began to manufacture fuel assemblies and associated core components for the first core of QINSHAN 300 MW nuclear power plant designed and built by China itself. With the development of nuclear power industry in China and the demand for localization of nuclear fuel elements in the early 1990's, YFP cooperated with FRAMATOME France in technology transfer for design and manufacturing of AFA 2G fuel assembly and successfully supplied the qualified fuel assemblies for the reloads of two units of GUANGDONG Da Ya Bay 900 MW nuclear power plant (Da Ya Bay NPP), and has achieved the localization of fuel assemblies and nuclear power plants. Meanwhile, it supplied fuel assemblies and associated core components for the first core and further reloads of Pakistan CHASHMA 300 MW nuclear power plant which was designed and built by China, and now it is manufacturing AFA 2G fuel assemblies and associated core components for the first core of two units of NPQJVC 600 MW nuclear power plant. From 2001 on, YFP will be able to supply Da Ya Bay NPP with the third generation of fuel assembly-AFA 3G which is to realize a strategy to develop the fuel assembly being of long cycle reload and high burn-up

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

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

  17. Detail design of test loop for FIV in fuel bundle and preliminary test

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Woo Gunl; Lee, Wan Young; Kim, Sung Won [Hannam University, Taejeon (Korea)

    2002-04-01

    It is urgent to develop the analytical model for the structural/mechanical integrity of fuel rod. In general, it is not easy to develop a pure analytical model. Occasionally, experimental results have been utilized for the model.Because of this reason, it is required to design proper test loop. Using the optimized test loop, With the optimized test loop, the dynamic behaviour of the rod will be evaluated and the critical flow velocity, which the rod loses the stability in, will be measured for the design of the rod. To verify the integrity of the fuel rod, it is required to evaluate the dynamic behaviour and the critical flow velocity with the test loop. The test results will be utilized to the design of the rod. Generally, the rod has a ground vibration due to turbulence in wide range of flow velocity and the amplitude of vibration becomes larger by the resonance, in a range of the velocity where occurs vortex. The rod loses stability in critical flow velocity caused by fluid-elastic instability. For the purpose of the present work to perform the conceptional design of the test loop, it is necessary (1) to understand the mechanism of the flow-induced vibration and the related experimental coefficients, (2) to evaluate the existing test loops for improving the loop with design parameters and (3) to decide the design specifications of the major equipments of the loop. 35 refs., 14 figs., 4 tabs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  19. Critical heat flux and post-critical heat flux performance of a 6-m, 37-element fully segmented bundle cooled by Freon-12

    International Nuclear Information System (INIS)

    A 6-m, 37-element, electrically heated bundle with full end plate simulation, cooled by Freon-12, has been tested for CHF (critical heat flux) and post-CHF conditions in the MR-3 Freon loop. The bundle was tested in a horizontal attitude and had a uniform axial heat flux distribution and radial heat flux depression. A total of 110 CHF points have been collected over the following range of water equivalent conditions: exit pressure 8.27 - 11.03 MPa, mass flux 1.38 - 8.14 Mg.m-2.s-1, inlet subcooling 0 - 500 kJ.kg-1, outlet quality 10% - 37%. The data have been correlated on both a systems and local conditions basis over a limited mass flux range to within 2.8% rms. Significant CHF increases over smooth bundle results have been observed along with significant CHF improvement over a two end plate bundle simulation in the lower mass flux ranges. A satisfactory axial drypatch spreading correlation has been determined and extensive drypatch wall superheat mapping has been performed

  20. Failed MTR Fuel Element Detect in a Sipping Tests

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Uranium density reduction on fuel element side plates assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rios, Ilka A. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-22

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

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

    Science.gov (United States)

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

    2016-01-01

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

  6. Overview of CEA's R&D on GFR fuel element design: from challenges to solutions

    International Nuclear Information System (INIS)

    Over the period 2002-2012, CEA conducted some extensive R&D on the design of GFR fuel elements (together with related material and core/system studies). This paper reviews the challenges raised by this programme, the solutions proposed to address them, and the remaining issues. Studies were performed on the assembly duct, the pin bundle and the fuel pin. The main issues were related to the challenge of using silicon carbide composites (SiC/SiC) for the pin cladding and the assembly duct, as well as mixed uranium-plutonium carbide (UPuC) for the nuclear fuel. Emphasizing the pin design, key achievements are reviewed in this paper regarding such topics as fission product confinement and high burnup performance, for the sake of which original design options were recently patented. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  8. Finite element simulation of thermal, elastic and plastic phenomena in fuel elements

    International Nuclear Information System (INIS)

    Taking as starting point an irradiation experiment of the first Argentine MOX fuel prototype, performed at the HFR reactor of Petten, Holland, the deformation suffered by the fuel element materials during burning has been numerically studied. Analysis of the pellet-cladding interaction is made by the finite element method. The code determines the temperature distribution and analyzes elastic and creep deformations, taking into account the dependency of the physical parameters of the problem on temperature. (author)

  9. Structural dynamics studies for single and clustered SNR-300 fuel elements: a comparison of analytical and experimental results

    International Nuclear Information System (INIS)

    Structural damage may occur in liquid metal cooled fast breeder reactors (LMFBR) when local failures escalate to a fast release of excessive thermal energy within one fuel element. Time scales (5. to 50. msec) and peaks of associated pressure pulses (50. to 500. bar) may exceed the burst pressure of the hexagonal wrapper containing the pin-bundle. Though after fuel element rupture the pressure level in the vicinity is markedly reduced, still a significant pressure field is propagating through the core. Therefore surrounding subassemblies or control rods are exposed to impulsive differential pressures causing duct flexure as well as local cross section deformation. A theoretical and experimental research program at GfK Karlsruhe investigates the mechanical effects of conservative pressure transients on LMFBR fuel elements. As part of the program this paper describes two computer codes (BEDYN-2, CORE-1) for non-linear structural dynamics of single as well as clustered subassemblies. Important features and idealizations of mechanical models are discussed before the codes are applied to several loading cases. Then two types of subassembly deformation experiments are outlined, and original records of the pressure pulses are used as input for BEDYN-2 to predict and compare deformation histories with experimental findings: (1) Quasistatic plane strain flat-to-flat loading of the hexagonal wrapper with and without the stiffening pin-bundle. (2) Transverse impulsive loading of a complete 1:1 SNR-300 type fuel element model on roller supports. In addition, pressure records from underwater explosion tests on SNR code models are taken as input for the multirow dynamics code CORE-1 and fairly good correlations were found between measured and predicted permanent deformations

  10. Performance and management of IPR-R1 fuel elements

    International Nuclear Information System (INIS)

    The performance of fuel elements during the 23 years of the reactor operation, is presented aiming to introduce improvements in the fuel load distribution and consequent increase of the reactivity. A computer code CORE was developed aiming to calculate the individual burnup of the fuel elements and the value of the reactivity for several core configurations, establishing a routine to control the nuclear material in the IPR-R1. The values calculated were compared with the experimental results. Some alternatives to augment the reactivity of the present core are presented foreseeing the fuel load availability for operation with 100Km and, for angmenting the power reaction in a next stage. (E.G.)

  11. Detection of fuel element vibration at KNK II

    International Nuclear Information System (INIS)

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

  12. Research Progress About Gas-Exhaust-Device for Fuel Element

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wu-ye

    2012-01-01

    <正>UO2-x stack applied in the fuel element has a form of a cylinder with a central hole, where temperature field characterized by high temperature and high gradient is formed due to irradiation. Then nearly all of the gaseous fission products (GFPs) can release into central cavity. However, uranium oxide will evaporate form the fuel stack’s inner surface because of its high temperature (about 1 800-2 000 ℃),

  13. The manufacture of LEU fuel elements at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J.

    1997-08-01

    Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

  14. CONDOR: neutronic code for fuel elements calculation with rods

    International Nuclear Information System (INIS)

    CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author)

  15. Properties of U3Si2-Al dispersion fuel element and its application

    International Nuclear Information System (INIS)

    The properties of U3Si2 fuel and U3Si2-Al dispersion fuel element are introduced, which include U-loading; the banding quality, U-homogeneity and 'dog-bone' phenomenon, the minimum thickness of cladding and the corrosion performances. The fabrication technique of fuel elements, NDT for fuel plates, assemble technique of fuel elements and the application of U3Si2-Al dispersion fuel elements in the world are introduced

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

    Science.gov (United States)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

  18. Three-dimensional porous media based numerical investigation of spatial power distribution effect on advanced nuclear fuel rod bundles critical power

    Energy Technology Data Exchange (ETDEWEB)

    Stosic, Zoran V. [Framatome ANP GmbH . NBTT, Erlangen (Germany)], e-mail: Zoran.Stosic@Framatome-ANP.de; Stevanovic, Vladimir D. [Framatome ANP GmbH, Erlangen (Germany); Iguchi, Tadashi [Japan Atomic Energy Research Institute (JAERI), Ibaraki (Japan)

    2001-07-01

    The influence of spatial power generation shape on thermal-hydraulics behaviour of the fuel rod bundle has been investigated. Particularly, the occurrence of the local Boiling Transition has been analysed, indicating that conditions for the Critical Heat Flux (CHF) are reached somewhere within the boiling water channels in the assembly. The two-phase coolant flow within the bundle is represented with the two-fluid model in 3D space. The porous medium concept is applied in the simulation of the two-phase flow through the rod bundle implying nonequilibrium thermal and flow conditions. The governing equations in three-dimensions are discretized with the control volume method. The 3D numerical simulation and analyses of thermal-hydraulics in a complex geometry of an advanced nuclear fuel assembly are performed for conditions of a partial and/or complete rods uncovering indicating occurrence of high quality CHF - Dryout. The obtained results from numerical simulations are compared with experimental Critical Power data obtained from full scale tests. Employed is an electrically heated test rod bundle with real 1:1 geometry. Different radial and axial power distributions are used with wide range of inlet mass flow rates (2 - 19 kg/s) and coolant inlet subcooling (25 - 185 kJ/kg). The coolant pressure, equal to 6.9 MPa, is typical for BWRs conditions. Comparison of the predicted Critical Power values with measured data shows encouraging agreements for all analysed power distributions and the results completely reflect measured two-phase mixture cross flows, steam void distribution and spatial positions of Dryout onsets. Based on performed numerical investigation, an improvement of Dryout criteria is proposed. Dynamic effects of power shape change on spatial thermal hydraulics and hence on CHF occurrence as well as the influence of transfer function on thermal hydraulics under cyclic power and/or flow rate changes are also being analysed. Experiments for such verifications

  19. Fission product release from defected nuclear reactor fuel elements

    International Nuclear Information System (INIS)

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

  20. CFD analysis of flow and heat transfer in Canadian supercritical water reactor bundle

    International Nuclear Information System (INIS)

    Highlights: • Flow and heat transfer in SCWR fuel bundle design by AECL is studied using CFD. • Bare-rod bundle geometry is tested at 23.5, 25 and 28 MPa using STAR-CCM+ code. • SST k–ω low-Re model was used to study occurrence of heat transfer deterioration. - Abstract: Within the Gen-IV International Forum, AECL is leading the effort in developing a conceptual design for the Canadian SCWR. AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for occurrence of HTD in the supercritical bundle flows. In the current investigation, bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The low-Reynolds number modification of SST k–ω turbulence model along with y+ < 1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5 MPa)

  1. Design evaluation of the HTGR fuel element size reduction system

    Energy Technology Data Exchange (ETDEWEB)

    Strand, J.B.

    1978-06-01

    A fuel element size reduction system for the ''cold'' pilot plant of the General Atomic HTGR Reference Recycle Facility has been designed and tested. This report is both an evaluation of the design based on results of initial tests and a description of those designs which require completion or modification for hot cell use. 11 figures.

  2. Design evaluation of the HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

    A fuel element size reduction system for the ''cold'' pilot plant of the General Atomic HTGR Reference Recycle Facility has been designed and tested. This report is both an evaluation of the design based on results of initial tests and a description of those designs which require completion or modification for hot cell use. 11 figures

  3. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

  5. Experimental analysis of heat flow in simulated fuel elements

    International Nuclear Information System (INIS)

    Since the experimental point of view it has been developed so much thermic simulations of nuclear reactors fuel elements in the laboratory. It is treating to isolate the problem of heat transfer of the complexity of the radioactive materials handling. The simulations starting of electric warming of similar geometric bodies to the real fuel elements. In the Thermo fluids Laboratory of National Institute of Nuclear Research it has been carried out heat transfer experiments in simulated fuel elements using in a first step concentric cylinders, for later to pass to posterior step of direct warming. The purpose of this work is to determine the convective parameters in the refrigerating under the typical prevailing conditions in the experimental reactors. It has been planned to work with isolated bars and groups of bars in convection with water. These works will allow to stablish the infrastructure of laboratory where it can be simulated thermically fuel elements of diverse types of experimental reactors. And specially to observe the solid-fluid effects in vertical surfaces subjected to intense heat fluxes. (Author)

  6. Poolside inspection, repair and reconstitution of LWR fuel elements

    International Nuclear Information System (INIS)

    The purpose of the meeting was to review the state of the art in the area of poolside inspection, repair and reconstitution of light water fuel elements. In the present publication it appears that techniques of inspection, repair and reconstitution of fuel elements have been developed by fuel suppliers and are now routinely and successfully applied in many countries. For the first time, the subject of control rod poolside examination was dealt with, poolside inspection and repair of a MOX assembly were reported and the inspection and repair of WWER assemblies were examined. Compared to the results of the previous meeting, present developments in the area aim now at reaching better economics, better reliability, reduction of personal doses and waste volume. Thirty-six participants representing twelve countries attended the meeting. Fifteen papers were presented in two sessions. An abstract was prepared for each of these papers. Refs, figs, tabs, diagrams, pictures and photos

  7. The technical concept of a temporary store for fuel elements

    International Nuclear Information System (INIS)

    In the German federal government's opinion, interim storage on the sites of nuclear power plants of spent fuel elements is to minimize the number of transports within Germany. As a span of approximately five years must be bridged until the interim stores now planned and filed for will be commissioned and, at the same time, transport activities are to be reduced, a kind of anticipated interim storage, or temporary storage, on power plant sites is unavoidable. The concept for the temporary storage of spent fuel elements is described in the article. On the basis of this concept, the Neckar Joint Nuclear Power Station recently was awarded a storage permit for nuclear fuels under Sec. 6 of the German Atomic Energy Act. Temporary stores following the same concept have been filed for, and are now in the licensing procedure, for another four sites (Philippsburg, Biblis, Kruemmel, Brunsbuettel). (orig.)

  8. CFD simulating the transient thermal–hydraulic characteristics in a 17 × 17 bundle for a spent fuel pool under the loss of external cooling system accident

    International Nuclear Information System (INIS)

    Highlights: • A 3-D CFD is adopted to simulate transient behaviors in an SFP under the accident. • This model realistically simulates a 17 × 17 bundle, rid of porous media approach. • The loss of external cooling system accident for an SFP is assumed in this paper. • Thermal–hydraulic characteristics in a bundle are strongly influenced by grids. • The results confirm temperature rising rate used in Maanshan NPP is conservative. - Abstract: This paper develops a three-dimensional (3-D) transient computational fluid dynamics (CFD) model to simulate the thermal–hydraulic characteristics in a fuel bundle located in a spent fuel pool (SFP) under the loss of external cooling system accident. The SFP located in the Maanshan nuclear power plant (NPP) is selected herein. Without adopting the porous media approach usually used in the previous CFD works, this model uses a real-geometry simulation of a 17 × 17 fuel bundle, which can obtain the localized distributions of the flow and heat transfer during the accident. These distribution characteristics include several peaks in the axial distributions of flow, pressure, temperature, and Nusselt number (Nu) near the support grids, the non-uniform distribution of secondary flow, and the non-uniform temperature distribution due to flow mixing between rods, etc. According to the conditions adopted in the Procedure 597.1 (MNPP Plant Procedure 597.1, 2010) for the management of the loss-of-cooling event of the spent fuel pool in the Maanshan NPP, the temperature rising rate predicted by the present model can be equivalent to 1.26 K/h, which is the same order as that of 3.5 K/h in the this procedure. This result also confirms that the temperature rising rate used in the Procedure 597.1 for the Maanshan NPP is conservative. In addition, after the loss of external cooling system, there are about 44 h for the operator to repair the malfunctioning system or provide the alternative water source for the pool inventory to

  9. Some parametric flow analyses of a particle bed fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.

    1993-05-01

    Parametric calculations are performed, using the SAFSIM computer program, to investigate the fluid mechanics and heat transfer performance of a particle bed fuel element. Both steady-state and transient calculations are included, addressing such issues as flow stability, reduced thrust operation, transpiration drag, coolant conductivity enhancement, flow maldistributions, decay heat removal, flow perturbations, and pulse cooling. The calculations demonstrate the dependence of the predicted results on the modeling assumptions and thus provide guidance as to where further experimental and computational investigations are needed. The calculations also demonstrate that both flow instability and flow maldistribution in the fuel element are important phenomena. Furthermore, results are encouraging that geometric design changes to the element can significantly reduce problems related to these phenomena, allowing improved performance over a wide range of element power densities and flow rates. Such design changes will help to maximize the operational efficiency of space propulsion reactors employing particle bed fuel element technology. Finally, the results demonstrate that SAFSIM is a valuable engineering tool for performing quick and inexpensive parametric simulations addressing complex flow problems.

  10. Gamma scanning of full scale HTR fuel elements

    International Nuclear Information System (INIS)

    Gamma scanning for the determination of burn-up and fission product inventory has been developed at the Dragon Project, suitable for measurements on fuel elements and segments from full-sized integral block elements. This involved the design and construction of a new lead flask with sophisticated collimator design. State-of-the art gamma spectrometric equipment was set up to cope with strong variations of count-rate and high data throughput. Software efforts concentrated on the calculation of the self absorption and absorption corrections in the complicated geometry of multi-hole graphite block segments with a corrugated circumference. The techniques described here are applicable to the non-destructive examination of a wide range of fuel element designs. (author)

  11. Interactions in Zircaloy/UO2 fuel rod bundles with Inconel spacers at temperatures above 1200deg C (posttest results of severe fuel damage experiments CORA-2 and CORA-3)

    International Nuclear Information System (INIS)

    In the CORA experiments test bundles of usually 16 electrically heated fuel rod simulators and nine unheated rods are subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of an LWR. An aim of CORA-2, as a first test of its kind, was also to gain experience in the test conduct and posttest handling of UO2 specimens. CORA-3 was performed as a high-temperature test. The transient phases of CORA-2 and CORA-3 were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1000deg C, leading the bundles to maximum temperatures of 2000deg C and 2400deg C for tests CORA-2 and CORA-3, respectively. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO2 interaction, complete Inconel spacer destruction, and relocation of melts and fragments to lower elevations in the bundle, where extended blockages have formed. In both tests the fuel rod destruction set in together with the formation of initial melts from the Inconel/Zry interaction. The lower Zry spacer acted as a catcher for relocated material. In test CORA-2 the UO2 pellets partially disintegrated into fine particles. This powdering occurred during cooldown. There was no physical disintegration of fuel in test CORA-3. (orig./MM)

  12. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    International Nuclear Information System (INIS)

    IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm3 for U3Si2-Al dispersion-based and 2.3 gU/cm3 for U3O8-Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm3 in U3Si2-Al dispersion and 3.2 gU/cm3 U3O8-Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U3Si2-Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U3O8-Al dispersion fuel plates with 3.2 gU/cm3 showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U3Si2 production at 4.8 gU/cm3, with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

  13. MAW and HTR fuel element test disposal in boreholes

    International Nuclear Information System (INIS)

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

  14. HTGR spent fuel element decay heat and source term analysis

    International Nuclear Information System (INIS)

    Decay heat, gamma dose rates, and neutron source strengths were determined for spent fuel elements from a High-Temperature Gas-Cooled Reactor (HTGR). The calculations were based on curie values reported in General Atomic Report GA-A13886 for the earlier commercial version of a 3000-MW(t) HTGR utilizing the thorium-uranium four-year fuel cycle. The reactor core was designed for an average thermal power density of 8.5 watts per cm3 and a carbon-to-thorium atom ratio which varies between 210:1 and 240:1. Calculations of decay heat, gamma dose rates, and neutron source strengths were made for spent fuel elements from the initial core and from representative nonrecycle and recycle reloads. The study was performed for decay times from 180 days to 10 years. Tables of the isotopic results are given for both the fertile and fissile particles in the fuel elements. In addition, ordered tables of the important isotopic contributors are presented. Graphical presentations of the results are shown and discussed; in addition, comparisons are made with previous determinations

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-03-21

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2011-01-01

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

  19. Single and two-phase flow pressure drop for CANFLEX bundle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E. [Atomic Energy of Canada Limited, Ontario (Canada)

    1998-12-31

    Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-134a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLEX bundle is found to be about 20% higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within {+-} 5% error. 11 refs., 5 figs. (Author)

  20. Installation of an irradiated fuel bundle discharge counter at Bruce NGS-B 3 000 MW(e) CANDU power station

    International Nuclear Information System (INIS)

    Design, manufacture and installation of an irradiated fuel bundle discharge counter for the multi-unit CANDU Bruce NGS-B Generating Station involved contributions from the International Atomic Energy Agency (Agency), designers (AECL), contractors, manufacturers, utility and the regulatory agency. The installation at Bruce NGS-B was the first made by the Agency as a retrofit to a multi-unit CANDU reactor approaching its fist critical operation, where the whole project was the responsibility of the Agency and where the original design of the reactor had not had provision for the Agency equipment. The scheduling and integration of the installation into the normal activities involved in starting up a 3 000 MW(e) multi-unit generating station were successfully achieved. The Agency has demonstrated the capability and performance of the fuel discharge counter

  1. Application of Be-free Zr-based amorphous sputter coatings as a brazing filler metal in CANDU fuel bundle manufacture

    International Nuclear Information System (INIS)

    Amorphous sputter coatings of Be-free multi-component Zr-based alloys were applied as a novel brazing filler metal for Zircaloy-4 brazing. By applying the homogeneous and amorphous-structured layers coated by sputtering the crystalline targets, the highly reliable joints were obtained with the formation of predominantly grown α-Zr grains owing to a complete isothermal solidification, exhibiting high tensile and fatigue strengths as well as excellent corrosion resistance, which were comparable to those of Zircaloy-4 base metal. The present investigation showed that Be-free and Zr-based multi-component amorphous sputter coatings can offer great potential for brazing Zr alloys and manufacturing fuel rods in CANDU fuel bundle system. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-01

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

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

    International Nuclear Information System (INIS)

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

  5. Recent operating experience with 28 element fuel at Pickering NGS

    International Nuclear Information System (INIS)

    A review of 28-element fuel operating experience at Pickering NGS is presented. The following topics are discussed: 1. Recent experience with in-core defects and 131I releases; 2. Operating strategies to minimize defect potential or to mitigate 131I releases to the primary heat transport system; 3. Impact of reduced regulatory limits as well as higher corporate expectations on operating strategies. 3 refs., 3 figs., 2 tabs

  6. Dry store for spent fuel elements from nuclear reactors

    International Nuclear Information System (INIS)

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

  7. CARA CVN: inherently safe fuel element for PHWR power plants

    International Nuclear Information System (INIS)

    This paper presents design alternatives of the CARA fuel element with negative void reactivity coefficient (CVN) enhancing the PHWR safety for L-LOCA sequences. This design enhances the safety and the operation performance in Atucha and Embalse without changes in the operation conditions. This new design balances wide performance margins of CARA SEU 0.9% previous design, with new intrinsic safety requirements without economic penalties. (author)

  8. METHOD AND APPARATUS FOR EXAMINING FUEL ELEMENTS FOR LEAKAGE

    Science.gov (United States)

    Smith, R.R.; Echo, M.W.; Doe, C.B.

    1963-12-31

    A process and a device for the continuous monitoring of fuel elements while in use in a liquid-metal-cooled, argonblanketed nuclear reactor are presented. A fraction of the argon gas is withdrawn, contacted with a negative electrical charge for attraction of any alkali metal formed from argon by neutron reaction, and recycled into the reactor. The electrical charge is introduced into water, and the water is examined for radioactive alkali metals. (AEC)

  9. Convective parameters in fuel elements for research nuclear reactors

    International Nuclear Information System (INIS)

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

  10. The element technology of clean fuel alcohol plant construction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.S; Lee, D.S. [Sam-Sung Engineering Technical Institute (Korea, Republic of); Choi, C.Y [Seoul National University, Seoul (Korea, Republic of)] [and others

    1996-02-01

    The fuel alcohol has been highlighted as a clean energy among new renewable energy sources. However, the production of the fuel alcohol has following problems; (i)bulk distillate remains is generated and (ii) benzene to be used as a entertainer in the azeotropic distillation causes the environmental problem. Thus, we started this research on the ground of preserving the cleanness in the production of fuel alcohol, a clean energy. We examined the schemes of replacing the azotropic distillation column which causes the problems with MSDP(Molecular Sieve Dehydration Process) system using adsorption technology and of treating the bulk distillate remains to be generated as by-products. In addition, we need to develop the continuous yea station technology for the continuous operation of fuel alcohol plant as a side goal. Thus, we try to develop a continuous ethanol fermentation process by high-density cell culture from tapioca, a industrial substrate, using cohesive yeast. For this purpose, we intend to examine the problem of tapioca, a industrial substrate, where a solid is existed and develop a new process which can solve the problem. Ultimately, the object of this project is to develop each element technology for the construction of fuel alcohol plant and obtain the ability to design the whole plant. (author) 54 refs., 143 figs., 34 tabs.

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

    International Nuclear Information System (INIS)

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

  12. A Multi-Dimensional Heat Transfer Model of a Tie-Tube and Hexagonal Fuel Element for Nuclear Thermal Propulsion

    Science.gov (United States)

    Gomez, C. F.; Mireles, O. R.; Stewart, E.

    2016-01-01

    The Space Capable Cryogenic Thermal Engine (SCCTE) effort considers a nuclear thermal rocket design based around a Low-Enriched Uranium (LEU) design fission reactor. The reactor core is comprised of bundled hexagonal fuel elements that directly heat hydrogen for expansion in a thrust chamber and hexagonal tie-tubes that house zirconium hydride moderator mass for the purpose of thermalizing fast neutrons resulting from fission events. Created 3D steady state Hex fuel rod model with 1D flow channels. Hand Calculation were used to set up initial conditions for fluid flow. The Hex Fuel rod uses 1D flow paths to model the channels using empirical correlations for heat transfer in a pipe. Created a 2-D axisymmetric transient to steady state model using the CFD turbulent flow and Heat Transfer module in COMSOL. This model was developed to find and understand the hydrogen flow that might effect the thermal gradients axially and at the end of the tie tube where the flow turns and enters an annulus. The Hex fuel rod and Tie tube models were made based on requirements given to us by CSNR and the SCCTE team. The models helped simplify and understand the physics and assumptions. Using pipe correlations reduced the complexity of the 3-D fuel rod model and is numerically more stable and computationally more time-efficient compared to the CFD approach. The 2-D axisymmetric tie tube model can be used as a reference "Virtual test model" for comparing and improving 3-D Models.

  13. Development of multi-dimensional thermal hydraulic modeling using mixing factors for wire wrapped fuel pin bundles with inter-subassembly heat transfer in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, M.; Kamide, H.; Ohshima, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-10-01

    Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature. The prediction of temperature distribution in the sub-assembly is, therefore one of the important issues for the reactor safety assessment. To treat the complex phenomena in the core, a multi-dimensional thermal hydraulic analysis is the most promising method. From the studies on the multi-dimensional thermal hydraulic modeling for the fuel sub-assemblies, the modeling have been recommended through the analysis of sodium experiments using driver subassembly test rig PLANDTL-DHX and blanket subassembly test rig CCTL-CFR. Computations of steady states experiments in the test rigs using the above modeling showed quite good agreement to the experimental data. In the present study, the use of this modeling was extended to transient analyses, and its applicability was examined. Firstly, non-dimensional parameters used to determine the mixing factors were modified from the ones based on bundle-averaged values to the ones by local values. Secondly, a new threshold function was derived and introduced to cut off the mixing factor of thermal plumes under inertia force dominant conditions. In the results of this validation, the accuracy was comparable between the modeling and the experimental instrumentation. Thus the present modeling is capable of predicting the thermal hydraulic fields of the wire wrapped fuel pin bundles with inter-subassembly heat transfer under the conditions from rated steady operations to transitions toward natural circulation decay heat removal modes. (J.P.N.)

  14. Gamma-ray spectroscopy on irradiated MTR fuel elements

    Science.gov (United States)

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

    2000-08-01

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

  15. Bundling biodiversity

    OpenAIRE

    Heal, Geoffrey

    2002-01-01

    Biodiversity provides essential services to human societies. Many of these services are provided as public goods, so that they will typically be underprovided both by market mechanisms (because of the impossibility of excluding non-payers from using the services) and by government-run systems (because of the free rider problem). I suggest here that in some cases the public goods provided by biodiversity conservation can be bundled with private goods and their value to consumers captured in th...

  16. An experimental and analytical study of fluid flow and critical heat flux in PWR fuel elements

    International Nuclear Information System (INIS)

    This report describes experiments that have been carried out at the Winfrith Establishment of the United Kingdom Atomic Energy Authority to determine the critical heat flux characteristics of pressurized water reactor fuel elements over an unusually wide range of coolant flow conditions that are relevant to both normal and fault conditions of reactor operation. The experiments were carried out in the TITAN loop using an electrically heated bundle of 25 rods of 9.5 mm diameter on a 12.7 mm pitch fitted with plain grids in order to provide a generic base for code validation. The fully tabulated experimental data for critical heat flux, pressure drop and sub-channel mixing are encompassed by ranges of pressure between 20 and 160 Bar, coolant flow between 150 and 3600 Kg/m2s, and coolant inlet temperature between 150 and 3200C. The results of the experiments are compared with predicted data based upon several established critical heat flux correlations. It is concluded that the extrapolation of some correlations to conditions beyond their intended range of application can lead to dangerous over estimates of critical heat flux, but the Winfrith WSC-2 and the EPRI NP-2609 correlations perform well over the whole data range and correlate all data with RMS errors of 9% and 6% respectively. (author)

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

    International Nuclear Information System (INIS)

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

  18. Semiflexible Biopolymers in Bundled Arrangements

    Directory of Open Access Journals (Sweden)

    Jörg Schnauß

    2016-07-01

    Full Text Available Bundles and networks of semiflexible biopolymers are key elements in cells, lending them mechanical integrity while also enabling dynamic functions. Networks have been the subject of many studies, revealing a variety of fundamental characteristics often determined via bulk measurements. Although bundles are equally important in biological systems, they have garnered much less scientific attention since they have to be probed on the mesoscopic scale. Here, we review theoretical as well as experimental approaches, which mainly employ the naturally occurring biopolymer actin, to highlight the principles behind these structures on the single bundle level.

  19. Polarly anisotropic thermoelasticity of cylindrical and spherical fuel elements

    International Nuclear Information System (INIS)

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

  20. Charging machine for the transport of fuel elements

    International Nuclear Information System (INIS)

    Charging machines for the transport of fuel elements for nuclear reactors have got a bridge body supported by two parallel rails via wheels. According to the invention the wheels are fixed to the bridge body by means of guide rods in such a way that at least relative movements in direction of the wheels and transversal to it are possible. Parallel to the guide rods springs and movement attenuators are force-locking by connected. Therefore a stabilizing effect with respect to the transversal forces occurring during earthquakes is achieved. (orig.)

  1. Storage system and method for spent fuel elements

    International Nuclear Information System (INIS)

    The proposal concerns an additional protection against leakage of a FE-transport container for interim storage of spent fuel elements. The gastight container has a second cover placed at a short distance from the first cover. The intermediate hollow space can be connected with a measuring system which indicates if part of the trace gas (mostly helium) added as indicator has escaped from the container due to leakage. The description explains the method and the assembly of required lines and measuring points etc. (UWI)

  2. Liquid films and droplet deposition in a BWR fuel element

    International Nuclear Information System (INIS)

    In the upper part of boiling water reactors (BWR) the flow regime is dominated by a steam-water droplet flow with liquid films on the nuclear fuel rod, the so called (wispy) annular flow regime. The film thickness and liquid flow rate distribution around the fuel rod play an important role especially in regard to so called dryout, which is the main phenomenon limiting the thermal power of a fuel assembly. The deposition of droplets in the liquid film is important, because this process sustains the liquid film and delays dryout. Functional spacers with different vane shapes have been used in recent decades to enhance droplet deposition and thus create more favorable conditions for heat removal. In this thesis the behavior of liquid films and droplet deposition in the annular flow regime in BWR bundles is addressed by experiments in an adiabatic flow at nearly ambient pressure. The experimental setup consists of a vertical channel with the cross-section resembling a pair of neighboring subchannels of a fuel rod bundle. Within this double subchannel an annular flow is established with a gas-water mixture. The impact of functional spacers on the annular flow behavior is studied closely. Parameter variations comprise gas and liquid flow rates, gas density and spacer shape. The setup is instrumented with a newly developed liquid film sensor that measures the electrical conductance between electrodes flush to the wall with high temporal and spatial resolution. Advanced post-processing methods are used to investigate the dynamic behavior of liquid films and droplet deposition. The topic is also assessed numerically by means of single-phase Reynolds-Averaged-Navier-Stokes CFD simulations of the flow in the gas core. For this the commercial code STAR-CCM+ is used coupled with additional models for the liquid film distribution and droplet motion. The results of the experiments show that the liquid film is quite evenly distributed around the circumference of the fuel rods. The

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

    International Nuclear Information System (INIS)

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

  4. Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Trammell, Michael P [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Qualls, A L [ORNL; Harrison, Thomas J [ORNL

    2013-01-01

    ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

  5. Comparative analysis of C A R A fuel element in argentinean PHWR Argentinas

    International Nuclear Information System (INIS)

    This paper presents an analysis of the thermal mechanical behaviour, fuel consumption and economical estimations of the CARA fuel element in the Atucha and Embalse nuclear power plants, compared with the present fuel performance.The present results show that the expect profit by the use of the CARA fuel element in our reactor guaranties the recovery of fund for its development. Likewise it reduces the number of spent fuel to be storage and treated

  6. REBEKA bundle experiments

    International Nuclear Information System (INIS)

    This report is a summary of experimental investigations describing the fuel rod behavior in the refilling and reflooding phase of a loss-of-coolant accident of a PWR. The experiments were performed with 5x5 and 7x7 rod bundles, using indirectly electrically heated fuel rod simulators of full length with original PWR-KWU-geometry, original grid spacers and Zircaloy-4-claddings (Type Biblis B). The fuel rod simulators showed a cosine shaped axial power profile in 7 steps and continuous, respectively. The results describe the influence of the different parameters such as bundle size on the maximum coolant channel blockage, that of the cooling on the size of the circumferential strain of the cladding (azimuthal temperature distribution) a cold control rod guide thimble and the flow direction (axial temperature distribution) on the resulting coolant channel blockage. The rewetting behavior of different fuel rod simulators including ballooned and burst Zircaloy claddings is discussed as well as the influence of thermocouples on the cladding temperature history and the rewetting behavior. All results prove the coolability of a PWR in the case of a LOCA. Therefore, it can be concluded that the ECC-criteria established by licensing authorities can be fulfilled. (orig./HP)

  7. Predictions of Critical Heat Flux Using the ASSERT-PV Subchannel Code for a CANFLEX Variant Bundle

    International Nuclear Information System (INIS)

    The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced CANDU1 reactor fuel bundle. Based primarily on the CANFLEX2 fuel bundle, several geometry changes (such as element sizes and pitchcircle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures

  8. Fabrication of a CANFLEX-RU designed bundle for power ramp irradiation test in NRU

    International Nuclear Information System (INIS)

    The BDL-443 CANFLEX-RU bundle AKW was fabricated at Korea Atomic Energy Research Institute (KAERI) for power ramp irradiation testing in NRU reactor. The bundle was fabricated with IDR and ADU fuel pellets in adjacent elements and contains fuel pellets enriched to 1.65 wt% 235U in the outer and intermediate rings and also contains pellets enriched to 2.00 wt% 235U in the inner ring. This bundle does not have a center element to allow for insertion on a hanger bar. KAERI produced the IDR pellets with the IDR-source UO2 powder supplied by BNFL. ADU pellets were fabricated and supplied by AECL. Bundle kits (Zircaloy-4 end plates, end plugs, and sheaths with brazed appendages) manufactured at KAERI earlier in 1996 were used for the fabrication of the bundle. The CANFLEX bundle was fabricated successfully at KAERI according to the QA provisions specified in references and as per relevant KAERI drawings and technical specification. This report covers the fabrication activities performed at KAERI. Fabrication processes performed at AECL will be documented in a separate report

  9. Graphitic matrix materials for spherical HTR fuel elements

    International Nuclear Information System (INIS)

    The present report comprises the essential results of material development and irradiation testing of graphitic matrix materials for spherical HTR fuel elements and completes the documentation of the irradiation data for 20 matrix materials (Juel-1702). The main emphasis is given to the matrices A3-3 (standard matrix) and A3-27 (matrix synthesized resin), both of which are being used as structural materials for the fuel elements of the AVR and the THTR respectively. In addition, comparisons are made between 18 A3-variants and the standard matrix A3-3. It is shown that three of the variants come into question as a potential for use. The results described were obtained in the framework of the HTR project 'Hochtemperaturreaktor-Brennstoffkreislauf' (HBK), in which are involved the Gesellschaft fuer Hochtemperaturreaktor-Technik mbH, Hochtemperaturreaktor-Brennelemente GmbH, Hochtemperatur-Reaktorbau GmbH, Kernforschungsanlage Juelich GmbH, NUKEM GmbH, and Sigri Elektrographit GmbH/Ringsdorff-Werke GmbH. The project is sponsored by the 'Bundesministerium fuer Forschung und Technologie' and by the state of 'Nordrhein-Westfalen'. (orig.)

  10. Quality control of CANDU6 fuel element in fabrication process

    International Nuclear Information System (INIS)

    To enhance the fine control over all aspects of the production process, improve product quality, fuel element fabrication process for CANDU6 quality process control activities carried out by professional technical and management technology combined mode, the quality of the fuel elements formed around CANDU6 weak links - - end plug , and brazing processes and procedures associated with this aspect of strict control, in improving staff quality consciousness, strengthening equipment maintenance, improved tooling, fixtures, optimization process test, strengthen supervision, fine inspection operations, timely delivery carry out aspects of the quality of information and concerns the production environment, etc., to find the problem from the improvement of product quality and factors affecting the source, and resolved to form the active control, comprehensive and systematic analysis of the problem of the quality management concepts, effectively reducing the end plug weld microstructure after the failure times and number of defects zirconium alloys brazed, improved product quality, and created economic benefits expressly provided, while staff quality consciousness and attention to detail, collaboration department, communication has been greatly improved and achieved very good management effectiveness. (authors)

  11. Triaxial Swirl Injector Element for Liquid-Fueled Engines

    Science.gov (United States)

    Muss, Jeff

    2010-01-01

    A triaxial injector is a single bi-propellant injection element located at the center of the injector body. The injector element consists of three nested, hydraulic swirl injectors. A small portion of the total fuel is injected through the central hydraulic injector, all of the oxidizer is injected through the middle concentric hydraulic swirl injector, and the balance of the fuel is injected through an outer concentric injection system. The configuration has been shown to provide good flame stabilization and the desired fuel-rich wall boundary condition. The injector design is well suited for preburner applications. Preburner injectors operate at extreme oxygen-to-fuel mass ratios, either very rich or very lean. The goal of a preburner is to create a uniform drive gas for the turbomachinery, while carefully controlling the temperature so as not to stress or damage turbine blades. The triaxial injector concept permits the lean propellant to be sandwiched between two layers of the rich propellant, while the hydraulic atomization characteristics of the swirl injectors promote interpropellant mixing and, ultimately, good combustion efficiency. This innovation is suited to a wide range of liquid oxidizer and liquid fuels, including hydrogen, methane, and kerosene. Prototype testing with the triaxial swirl injector demonstrated excellent injector and combustion chamber thermal compatibility and good combustion performance, both at levels far superior to a pintle injector. Initial testing with the prototype injector demonstrated over 96-percent combustion efficiency. The design showed excellent high -frequency combustion stability characteristics with oxygen and kerosene propellants. Unlike the more conventional pintle injector, there is not a large bluff body that must be cooled. The absence of a protruding center body enhances the thermal durability of the triaxial swirl injector. The hydraulic atomization characteristics of the innovation allow the design to be

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

    International Nuclear Information System (INIS)

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

  13. Corrosion studies in fuel element reprocessing environments containing nitric acid

    International Nuclear Information System (INIS)

    Nitric acid is universally used in aqueous fuel element reprocessing plants; however, in the processing scheme being developed by the Consolidated Fuel Reprocessing Program, some of the equipment will be exposed to nitric acid under conditions not previously encountered in fuel element reprocessing plants. A previous report presented corrosion data obtained in hyperazeotropic nitric acid and in concentrated magnesium nitrate solutions used in its preparation. The results presented in this report are concerned with the following: (1) corrosion of titanium in nitric acid; (2) corrosion of nickel-base alloys in a nitric acid-hydrofluoric acid solution; (3) the formation of Cr(VI), which enhances corrosion, in nitric acid solutions; and (4) corrosion of mechanical pipe connectors in nitric acid. The results show that the corrosion rate of titanium increased with the refreshment rate of boiling nitric acid, but the effect diminished rapidly as the temperature decreased. The addition of iodic acid inhibited attack. Also, up to 200 ppM of fluoride in 70% HNO3 had no major effect on the corrosion of either titanium or tantalum. In boiling 8 M HNO3-0.05 M HF, Inconel 671 was more resistant than Inconel 690, but both alloys experienced end-grain attack. In the case of Inconel 671, heat treatment was very important; annealed and quenched material was much more resistant than furnace-cooled material.The rate of oxidation of Cr(III) to Cr(VI) increased significantly as the nitric acid concentration increased, and certain forms of ruthenium in the solution seemed to accelerate the rate of formation. Mechanical connectors of T-304L stainless steel experienced end-grain attack on the exposed pipe ends, and seal rings of both stainless steel and a titanium alloy (6% Al-4% V) underwent heavy attack in boiling 8 M HNO3

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

    International Nuclear Information System (INIS)

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

  15. Development of finite element analysis code SPOTBOW for prediction of local velocity and temperature fields around distorted fuel pin in LMFBR assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-05-01

    A two-dimensional steady-state distributed parameter code SPOTBOW has been developed for predicting the fine structure of cladding temperature in an liquid metal fast breeder reactor (LMFBR) fuel assembly where the deformation of fuel pins is induced by irradiation swelling, creep and thermal distortion under high burn-up operating condition. When the deformed fuel pin approaches adjacent pins and wrapper tube and comes in contact with those, the peak temperature, known as the hot spot temperature, can appear somewhere on the outer surface of the cladding. The temperature rise across the film is an important consideration in the cladding temperature analysis. Fully developed turbulent momentum and heat transfer equations based on the empirical turbulent model are solved by using the Galerkin finite element method which is suitable for the problem of the complicated boundary shape, such as the wire-wrapped fuel pin bundle. A new iteration procedure has been developed for solving the above equations by using the rise in coolant temperature, which is obtained with subchannel analysis codes, as a boundary condition. Calculated results are presented for local temperature distribution in normal and bowing pin bundle geometry, as compared with experiments. (author).

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Determine the homogeneity of UO2 distribution in HTGR fuel element by X-ray tomography

    International Nuclear Information System (INIS)

    The homogeneity of UO2 distribution in HTGR fuel elements is one of the important properties of a fuel element. The X-ray tomography and image processing technology can nondestructively determine the distribution of UO2 particles in a spherical fuel element. A statement was specifically made on tomography ambiguity, otherwise, the technology and point of tomography to a spherical fuel element was described too. Through the computerized image processing, tomography film shows directly homogeneity of UO2 particles. The hardware for image processing and the special software packing were briefly introduced. As the results of measurement, the colour cross-images drawn by a typing machine or taken from from CRT by a camera indicated that determination homogeneity of UO2 particles in a fuel element homogeneity of UO2 particles in a fuel element sphere by tomography was very successful

  19. Burn-up measurements at TRIGA fuel elements containing strong burnable poison

    International Nuclear Information System (INIS)

    The reactivity method of determining the burn-up of research reactor fuel elements is applied to the highly enriched FLIP elements of TRIGA reactors. In contrast to other TRIGA fuel element types, the reactivity of FLIP elements increases with burn-up due to consumption of burnable poison. 33 fuel elements with burn-up values between 3% and 14% were investigated. The experiments showed that variations in the initial fuel composition significantly influence the reactivity and, consequently, increase the inaccuracy of the burn-up measurements. Particularly important are variations in the initial concentration of erbium, which is used as burnable poison in FLIP fuel. A method for reducing the effects of the material composition variations on the measured reactivity is presented. If it is applied, the accuracy of the reactivity method for highly poisoned fuel elements becomes comparable to the accuracy of other methods for burn-up determination. (orig.)

  20. Process for finding defective fuel element cans using ultrasonics

    International Nuclear Information System (INIS)

    As the distance between the test heads can change due to the gassing of fuel elements and as direct ultrasonic echos are assessed as rotary echos, the invention proposes that the expected time range of each can to be tested is determined afresh, before this can is placed between the test heads. The running time of the ultrasonic signals is therefore measured in the intermediate space between the last tested can and the can to be tested next. A constant value is subtracted from this value of running time, which is selected so that the ultrasonic signal received for the running time measurement just does not fall within the expected range. (orig./HP)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

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

  3. The behaviour of fission products in fuel elements of the AVR-core

    International Nuclear Information System (INIS)

    The fuel elements of the THTR-1-type-, THTR-2-type-, and CFB-2-type reactor have an extraordinary retention capacity for all fission products. Pressed-carbide fuel elements, however, at a temperature above 3000C release considerable Sr- and Eu-activities as a result of diffusion through intact PyC-layers. For GLE-1 and GFB-1 fuel elements a considerable release of cesium has been observed which is caused by defective coated particles. (DG)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-19

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-19

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

  6. Damping Properties of the Hair Bundle

    OpenAIRE

    Baumgart, Johannes; Kozlov, Andrei S.; Risler, Thomas; Hudspeth, A. James

    2015-01-01

    The viscous liquid surrounding a hair bundle dissipates energy and dampens oscillations, which poses a fundamental physical challenge to the high sensitivity and sharp frequency selectivity of hearing. To identify the mechanical forces at play, we constructed a detailed finite-element model of the hair bundle. Based on data from the hair bundle of the bullfrog's sacculus, this model treats the interaction of stereocilia both with the surrounding liquid and with the liquid in the narrow gaps b...

  7. Global analysis of bundle behavior in pressurized water reactor specific CORA experiments

    International Nuclear Information System (INIS)

    At Kernforschungszentrum Karlsruhe, out-of-pile bundle experiments are performed in the CORA facility to investigate the behavior of light water reactor fuel elements during severe fuel damage accidents. To analyze the phenomena observed during the tests, such as claddin failure, oxidation, and deformation, as well as their influence on the post test bundle state, four pressurized water reactor specific tests are selected: CORA-2, CORA-3, CORA-5, and CORA-12. From each of these tests, a detailed global analysis using all the measured temperatures, pressures, and fluid compositions as well as videoscope information has been performed. To describe the post test bundle state quantitatively, axial profiles of the bundle cross-section area, the damage state of the rods, the average cladding oxidation, and the damage to the pellets are measured. The effects of CORA-specific components on the bundle melt progression and the measured axial profiles are identified and assessed. Most of the observations during the tests as well as the post test bundle state can be explained by the established common sequence of phenomena. For a better understanding of the melt progression, some physical phenomena, such as the energy release associated with the double-sided oxidation of the cladding, the melt release, or the melt relocation, must be analyzed in detail

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Fuel Temperature Characteristics for Fuel Channels using Burnable Poison in the CANDU reactor

    International Nuclear Information System (INIS)

    Although the CANFLEX RU fuel bundle loaded 11.0 wt% Er2O3 are originally designed focused on the safety characteristics, the fuel temperature characteristics is revealed to be not deteriorated but rather is slightly enhanced by the decreased fuel temperature in the outer ring compared with that of standard 37 fuel bundle. Recently, for an equilibrium CANDU core, the power coefficient was reported to be slightly positive when newly developed Industry Standard Tool set reactor physics codes were used. Therefore, it is required to find a new way to effectively decrease the positive power coefficient of CANDU reactor without seriously compromising the economy. In order to make the power coefficient of the CANDU reactor negative at the operating power, Roh et al. have evaluated the various burnable poison (BP) materials and its loading scheme in terms of the fuel performance and reactor safety characteristics. It was shown that reactor safety characteristics can be greatly improved by the use of the BP in the CANDU reactor. In a view of safety, the fuel temperature coefficient (FTC) is an important safety parameter and it is dependent on the fuel temperature. For an accurate evaluation of the safety-related physics parameters including FTC, the fuel temperature distribution and its correlation with the coolant temperature should be accurately identified. Therefore, we have evaluated the fuel temperature distribution of a CANFLEX fuel bundle loaded with a burnable poison and compared the standard 37 element fuel bundle and CANFELX-NU fuel bundle

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. Nonuniform Oxidation on the Surface of Fuel Element in HTR

    Directory of Open Access Journals (Sweden)

    Peng Liu

    2016-01-01

    Full Text Available The graphite oxidation of fuel element has obtained high attention in air ingress accident analysis of high temperature gas-cooled reactor (HTR. The shape function, defined as the relationship between the maximum and the average of the oxidation, is an important factor to estimate the consequence of the accident. There are no detailed studies on the shape function currently except two experiments several decades ago. With the development of computer technology, CFD method is used in the numerical experiment about graphite oxidation in pebble bed of HTR in this paper. Structured packed beds are used in the calculation instead of random packed beds. The result shows the nonuniform distribution of oxidation on the sphere surface and the shape function in the condition of air ingress accident. Furthermore, the sensitive factors of shape function, such as temperature and Re number, are discussed in detail and the relationship between the shape function and sensitive factors is explained. According to the results in this paper, the shape function ranges from 1.05 to 4.7 under the condition of temperature varying from 600°C to 1200°C and Re varying from 16 to 1600.

  16. Experimental approach and modelling of the mechanical behaviour of graphite fuel elements subjected to compression pulses

    OpenAIRE

    Forquin P.

    2010-01-01

    Among the activities led by the Generation IV International Forum (GIF) relative to the future nuclear systems, the improvement of recycling of fuel elements and their components is a major issue. One of the studied systems by the GIF is the graphite-moderated high-temperature gas cooled reactor (HTGR). The fuel elements are composed of fuel roads half-inch in diameter named compacts. The compacts contain spherical particles made of actinide kernels about 500 m in diameter coated with t...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  1. Fate of Cu, Cr, As and some other trace elements during combustion of recovered waste fuels

    OpenAIRE

    Lundholm, Karin

    2007-01-01

    The increased use of biomass and recovered waste fuels in favor of fossil fuels for heat and power production is an important step towards a sustainable future. Combustion of waste fuels also offers several advantages over traditional landfilling, such as substantial volume reduction, detoxification of pathological wastes, and reduction of toxic leaches and greenhouse gas (methane) formation from landfills. However, combustion of recovered waste fuels emits more harmful trace elements than co...

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

    International Nuclear Information System (INIS)

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

  3. Impact of bundle deformation on CHF: ASSERT-PV assessment of extended burnup Bruce B bundle G85159W

    International Nuclear Information System (INIS)

    This paper presents a subchannel thermalhydraulic analysis of the effect on critical heat flux (CHF) of bundle deformation such as element bow and diametral creep. The bundle geometry is based on the post-irradiation examination (PIE) data of a single bundle from the Bruce B Nuclear Generating Station, Bruce B bundle G85159W, which was irradiated for more than two years in the core during reactor commissioning. The subchannel code ASSERT-PV IST is used to assess changes in CHF and dryout power due to bundle deformation, compared to the reference, undeformed bundle. (author)

  4. Investigations of the behaviour of coated fuel particles and spherical fuel elements at accident temperatures

    International Nuclear Information System (INIS)

    A post irradiation annealing test apparature was constructed for the measurement of fission gas release at temperatures similar to those to be reached in a HTR during a hypothetical accident. From examinations with existing apparatures up to temperatures of 18000C results were available about the load capacity of coated particles as well as knowledges about fission gas release and defect behaviour. These results were used to plan a series of annealing tests with spherical fuel elements up to 25000C. It could be shown that the (U,Th)O2-particles with high burn up will fail during maximum core heat up of a HTR only after some hours at temperatures above 24000C. (orig.)

  5. Strategic Aspects of Bundling

    International Nuclear Information System (INIS)

    The increase of bundle supply has become widespread in several sectors (for instance in telecommunications and energy fields). This paper review relates strategic aspects of bundling. The main purpose of this paper is to analyze profitability of bundling strategies according to the degree of competition and the characteristics of goods. Moreover, bundling can be used as price discrimination tool, screening device or entry barriers. In monopoly case bundling strategy is efficient to sort consumers in different categories in order to capture a maximum of surplus. However, when competition increases, the profitability on bundling strategies depends on correlation of consumers' reservations values. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO2 grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)

  8. Neutron and gamma radiography of UO2 and TRIGA fuel elements

    International Nuclear Information System (INIS)

    The Oregon State TRIGA Reactor neutron radiography facility has been used to produce both neutron and gamma radiographs of reactor fuel. In this paper a comparison of the applicability of neutron and gamma radiography to both UO2 fuel pins and TRIGA fuel elements is made. In the case of UO2 fuel, conventional thermal neutron radiography produces excellent quality radiographs. These radiographs may be used to detect various defects in the fuel such as enrichment differences, cracks, end-capping, inclusions, etc. For TRIGA fuel elements, conventional thermal neutron radiography will not show the internal structure. This is due to the high hydrogen content of the fuel. These elements are typically 8.5 w/o uranium in Zr-H1.7; the density of hydrogen in the fuel being about 80 percent that of water. Further, while epithermal radiography significantly improves the radiographs, defects may go undetected. As an alternative to neutron radiography, high energy gamma radiographs of TRIGA fuel elements have been taken using the same facility. The gamma spectrum emitted by the reactor core is sufficiently high in energy that very good radiographs may be obtained with this technique. These radiographs show excellent detail for the internal structure of the TRIGA fuel. (author)

  9. Neutron and gamma radiography of UO2 and TRIGA fuel elements

    International Nuclear Information System (INIS)

    The Oregon State TRIGA Reactor neutron radiography facility has been used to produce both neutron and gamma radiographs of reactor fuel. In this paper a comparison of the applicability of neutron and gamma radiography to both UO2 fuel pins and TRIGA fuel elements is made. In the case of UO2 fuel, conventional thermal neutron radiography produces excellent quality radiographs. These radiographs may be used to detect various defects in the fuel such as enrichment differences, cracks, end-capping, inclusions, etc. For TRIGA fuel elements, conventional thermal neutron radiography will not show the internal structure. This is due to the high hydrogen content of the fuel. These elements are typically 8.5 w/o uranium in Zr-Hsub(1.7); the density of hydrogen in the fuel being about 80 percent that of water. Further, while epithermal radiography significantly improves the radiographs, defects may go undetected. As an alternative to neutron radiography, high energy gamma radiographs of TRIGA fuel elements have been taken using the same facility. The gamma spectrum emitted by the reactor core is sufficiently high in energy that very good radiographs may be obtained with this technique. These radiographs show excellent detail for the internal structure of the TRIGA fuel. (Auth.)

  10. Assessment of Welding System Modification of The Candu and PWR Fuel Element Types end Plug

    International Nuclear Information System (INIS)

    To anticipate future possibility of a nuclear fuel element industry in Indonesia, research on other types of nuclear fuel element beside Cirene type has to be done. It can be accomplished, one of them, by modifying the already available equipment. Based on the sheath material, the sheath dimension and the welding process parameters such as welding current and welding cycles, the available Magnetic Force Welding can be used for welding end plug of Candu nuclear fuel element by modifying some of its components (tube clamp, plug clamp, etc). The available Pellet drying and element filling furnace with its supporting system with includes helium gas filling, welding chamber, argon gas supply, vacuum system, sheath clamp and sheath driving system can be used for welding end plug with sheath of PWR nuclear fuel element by adding og Tungsten inert Gas (TIG) welding machine in the welding chamber and modifying a few components (seal clamp, sheath clamp)

  11. Single-element coaxial injector for rocket fuel

    Science.gov (United States)

    Larson, L. L.

    1969-01-01

    Improved injector for oxygen difluoride and diborane has better mixing characteristics and is able to project fuel onto the wall of the combustion chamber for better cooling. It produces an essentially conical, diverging, continuous sheet of propellant mixture formed by similarly shaped and continuously impinging sheets of fuel and oxidant.

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

    OpenAIRE

    Casey, William Emerson

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. Bundling in Telecommunications

    OpenAIRE

    Begoña García-Mariñoso; Xavier Martinez-Giralt; Pau Olivella

    2008-01-01

    The paper offers an overview of the literature on bundling in the telecommunications sector and its application in the Spanish market. We argue that the use of bundling in the provision of services is associated to technological reasons. Therefore, there appears no need to regulate bundling activities. However, this is not to say that other related telecom markets should not be scrutinized and regulated, or that the regulator should not pay attention to other bundling-related anticompetitive ...

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

    Energy Technology Data Exchange (ETDEWEB)

    Rudock, E.R.

    1959-08-27

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

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

    Science.gov (United States)

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

    2016-10-01

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

  18. Fuel element gap irregularities determined from infrared scanning

    International Nuclear Information System (INIS)

    It has been hypothesized that the fuel pellet column in a nuclear fuel rod will not be concentric in the cladding. The pellets will assume random offsets, and frequent contact with the tube, resulting in a series of broken spirals. Evidence for the existence and stability of spiralling pellet columns has been observed from crud patterns on fuel rods irradiated for one and two cycles in commercial reactors. Such behavior has important implications on the pellet to cladding thermal conductivity under both normal and accident conditions and on the swelling and burst of cladding under LOCA conditions. An experimental program to verify and quantify the pellet offset behavior in an unirradiated commercial fuel rod was undertaken at ORNL under the sponsorship of the Central Electricity Generating Board and Westinghouse Electric Co. The results of the experimental program were used to develop a model to predict the burst strain behavior of the NRU test rods and were compared with the empirical correlation of the gap conductance inferred from instrumented Halden rods. The tests were performed on a Westinghouse 17 x 17 fuel rod made to the same specifications as commercial fuel except that the fuel pellets were made from depleted UO2

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

    International Nuclear Information System (INIS)

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

  20. CFD simulations of the single-phase and two-phase coolant flow of water inside the original and modified CANDU 37-element bundles

    International Nuclear Information System (INIS)

    Single-phase (inlet temperature of 180° C, outlet pressure of 9 MPa, total power of 2 MW and flow rate of 13.5 Kg/s), and two-phase (inlet temperature of 265° C, outlet pressure of 10 MPa, total power of 7.126 MW and flow rate of 19 Kg/s) water flows inside a CANDU thirty seven element fuel string are simulated using a Computational Fluid Dynamics (CFD) code with parallel processing and results are presented in this paper. The analyses have been performed for the original and modified (11.5 mm center element diameter) designs with skewed cosine axial heat flux distribution and 5.1% diametral creep of the pressure tube. The CFD results are in good agreement with the expected temperature and velocity cross-sectional distributions. The effect of the pressure tube creep on the flow bypass is detected, and the CHF improvement in the core region of the modified design is confirmed. The two-phase flow model reasonably predicted the void distribution and the role of interfacial drag on increasing the pressure drop. In all CFD models, the appendages were shown to enhance the production of cross flows and their corresponding flow mixing and asymmetry. (author)

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

    International Nuclear Information System (INIS)

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

  2. Current status of U3Si2 fuel element fabrication in Brazil

    International Nuclear Information System (INIS)

    IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U3O8-Al dispersion fuel plates with 2.3 gU/cm3. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicide was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)

  3. Thermal-hydraulics performance optimization of Candu fuel using Assert subchannel code

    International Nuclear Information System (INIS)

    An optimization of fuel bundle geometry using the subchannel code ASSERT is performed in support of Candu fuel design to enhance the thermohydraulics performance. The new bundle design is based on a reference CANFLEX bundle with changes to the centre and inner-ring element diameters and pitch-circle diameters (PCDs) of various element rings. Different methods of varying the PCDs for reaching the optimized geometry are considered in an attempt to minimize the optimization effort. The optimized geometry in the present analysis is the one that maximizes the dryout power and that has simultaneous CHF (critical heat flux) initiation involving more than one subchannel rings. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    International Nuclear Information System (INIS)

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

  6. Specialists' meeting on fuel element performance computer modelling, Preston, United Kingdom, 15-19 March 1982

    International Nuclear Information System (INIS)

    The 46 papers of the meeting concerned with computer models of Water Reactor fuel elements cover practically all aspects of behavior of fuel elements in normal operation and in accident condition. Each session of the meeting produced a critical evaluation of one of the 5 topics into which the subject area had been divided. The sessions' report summarize the papers and make recommendations for further work. Separate abstracts were prepared for all the papers presented at this meeting

  7. Distribution of fission products in Peach Bottom HTGR fuel element E11-07

    Energy Technology Data Exchange (ETDEWEB)

    Wichner, R.P.; Dyer, F.F.; Martin, W.J.; Bate, L.C.

    1977-04-01

    This is the second in a projected series of six post-irradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements. Element E11-07, the subject of this report, received an equivalent of 701 full-power days of irradiation prior to scheduled withdrawal. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a /sup 137/Cs inventory of 17 Ci in the graphite sleeve and 8.3 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides /sup 134/Cs, /sup 110m/Ag, /sup 60/Co, and /sup 154/Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the distribution of the beta emitters /sup 3/H, /sup 14/C, and /sup 90/Sr were obtained at six axial locations, four within the fueled region and one each above and below. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. These profiles reveal an increased degree of penetration of /sup 134/Cs, relative to /sup 137/Cs, evidently due to a longer time spent as xenon precursor. In addition to fission product distribution, the appearance of the element components was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed.

  8. Measurement of after-heat production and dose rates of spent AVR fuel elements

    International Nuclear Information System (INIS)

    Data on the afterheat production and dose rate of spent AVR fuel elements prepared by the ORIGEN computer program are verified by measurements. Individual measurements of afterheat and dose rate were implemented on 17 AVR fuel elements with decay periods of 150 days and more than four years, and burnups between 4.1 and 16.4% fima were implemented in the HOT CELLS at the Juelich Nuclear Research Centre. The radiation energy absorbed in the fuel elements and converted into heat was measured with a calorimeter, whereas the emitted radiation fraction was determined via dose rate measurements. The measured results for fuel elements with decay periods of more than one year are in good agreement with the data from ORIGEN. In the case of fuel elements with shorter decay periods (approx. 150 days) in part considerably lower values were measured which can be explained by the fact that the power gradient in time of the fuel elements in the reactor can vary considerably whereas mean are included in the ORIGEN computations assuming full-load operation. (orig./HP))

  9. Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

    Science.gov (United States)

    Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

    2013-01-01

    In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

  10. Radiological measurements during decontamination of PFBR MOX fuel elements using ultrasonic decontamination technique

    International Nuclear Information System (INIS)

    In a fuel fabrication facility fabrication of MOX fuel elements involving various metallurgical processes is carried out in leak tight glove boxes because of high radio toxicity associated with plutonium,. A fuel pin consists of a thin walled tube loaded with cylindrical fuel pellets with plugs welded on both ends. The pellet loading and welding processes result in cross contamination on the tube surface near the edges. It is important that finished fuel pins should not contain any transferable contamination on the surface beyond safe limits applicable for unrestricted release before subjecting the pins to manual handling for quality control checks. Hence it is imperative that thorough decontamination of fuel pins is essential for safe handling. Conventional decontamination methods result in undue personal exposures and generation of solid waste. Though there are number of techniques available for decontamination of non-fuel elements in the nuclear industry, very few of them can be used for decontamination of fuel elements because of possible damage to fuel clad, Ultrasonic cleaning process, using dc-mineralized water as medium does not affect the properties of the clad and is simple to implement and fast to carry out. This paper brings out radiological measurements carried out to study the effectiveness of ultrasonic decontamination technique and the factors involved in achieving required degree of decontamination with reduced individual exposure

  11. Damping Properties of the Hair Bundle

    Science.gov (United States)

    Baumgart, Johannes; Kozlov, Andrei S.; Risler, Thomas; Hudspeth, A. J.

    2011-11-01

    The viscous liquid surrounding a hair bundle dissipates energy and dampens oscillations, which poses a fundamental physical challenge to the high sensitivity and sharp frequency selectivity of hearing. To identify the mechanical forces at play, we constructed a detailed finite-element model of the hair bundle. Based on data from the hair bundle of the bullfrog's sacculus, this model treats the interaction of stereocilia both with the surrounding liquid and with the liquid in the narrow gaps between the individual stereocilia. The investigation revealed that grouping stereocilia in a bundle dramatically reduces the total drag. During hair-bundle deflections, the tip links potentially induce drag by causing small but very dissipative relative motions between stereocilia; this effect is offset by the horizontal top connectors that restrain such relative movements at low frequencies. For higher frequencies the coupling liquid is sufficient to assure that the hair bundle moves as a unit with a low total drag. This work reveals the mechanical characteristics originating from hair-bundle morphology and shows quantitatively how a hair bundle is adapted for sensitive mechanotransduction.

  12. Distribution of fission products in Peach Bottom HTGR fuel element E14-01

    Energy Technology Data Exchange (ETDEWEB)

    Wichner, R.P.; Dyer, F.F.; Martin, W.J.; Fairchild, L.L.

    1977-08-01

    The third in a projected series of six postirradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements is presented. Element E14-01, the subject of the report, was one of the 60 driver elements (out of a total of 804) that contained zirconium boride pellets within a hollow spine. It was also one of the few predimensioned elements, which therefore allowed accurate determination of dimensional change due to irradiation service. The element received an equivalent of 897 full-power days irradiation prior to scheduled termination of Core 2 operation. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a /sup 137/Cs inventory of 0.24 Ci in the graphite sleeve and 0.047 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides /sup 134/Cs, /sup 110m/Ag, /sup 60/Co, and /sup 154/Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the beta-emitters /sup 3/H, /sup 14/C, and /sup 90/Sr were obtained at three axial locations within the fueled region of the element. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. In addition to fission product distributions, the appearance of the component parts of the element was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  14. Dimensional changes in operating UO2 fuel elements: effect of pellet density, burnup and ramp rate

    International Nuclear Information System (INIS)

    We have used an in-reactor diameter measuring rig, in combination with a He-3 coil associated with the X-6 loop of the NRX reactor, to determine the dimensional response of CANDU-type UO2 fuel elements as a function of power, prior irradiation, fuel density and ramp rate. The maximum diametral increase accompanying a ramp was about 1% for high density fuel. No elements failed, despite, in the most severe case, a 1.5 h hold at 56 kW/m, after a ramp from 30 kW/m in less than three minutes. (author)

  15. Development of a new technique for experimental evaluation of the fuel element's subchannel mixing

    International Nuclear Information System (INIS)

    In this work, the development of a new experimental method for the measurement of mixing between the cooling subchannels of nuclear fuel elements by using thermal traces, is presented.The method has been proved on a reduced test section with very positive results, having demonstrated its simplicity and low cost.Because it is suitable for heterogeneous and compact subchannels (asArgentinean fuels) with high water flows in simple and affordable tests at atmospheric pressure, this new method is specially well suited for the design of fuel elements, while it offers advantages over other methods of mixing measurement

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

    International Nuclear Information System (INIS)

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

  17. The clad collapse modelling of Indian PHWR fuel element, an FEM approach

    International Nuclear Information System (INIS)

    The fuel elements for PHWR use a thin, collapsible zircaloy clad design. This design is consistent with essential neutron economy in PHWRs, and also results in better heat transfer between fuel and clad. However, thin clad may give rise to problem of permanent clad collapse under coolant pressure in axial gap and radial gap available during the initial stay of fuel inside the reactor. Present work explores the problem of longitudinal ridges, formed due to permanent circumferential collapse of clad on fuel. The tip of these ridges has the potential to become the site for crack initiation under subsequent cyclic thermal/pressure loading. The collapse behavior of fuel element is studied using FEM modeling of pellet, clad and their contact. This study considers the effects of clad thickness, clad yield strength, clad initial ovality, anisotropy in clad yield strength, and radial gap of fuel element on the collapse behavior. The verification of present model is done for the results of critical buckling pressure required for the longitudinal ridge formation by the available CANDU experimental data, which matched satisfactorily for the yield strength ratio (circumferential YS to longitudinal YS) of 1.5. In addition the longitudinal ridge height and increase in ovality were calculated for the collapse experiments done on the 220 MWe PHWR fuel elements

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-01

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

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

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

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

  20. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Golay, M.W.; Wold, L.

    1981-02-01

    Four tasks are reported on: bundle geometry (wrapped and bare rods), subchannel geometry (bare rods), LMFBR outlet plenum flow mixing, and theoretical determination of local temperature fields in LMFBR fuel rod bundles. (DLC)

  1. Fabrication and Properties of Cylindrical and Tubular UO2-Stainless Steel Fuel Elements

    International Nuclear Information System (INIS)

    The fabrication by extrusion sintering of canned tubes and rods containing a dispersed fuel nucleus of UO2-stainless steel was investigated. The fuel elements may be up to 2 m long; the diameter of the rods is less than 30 mm; the total thickness of the tubes may be as low as 0.8 mm. The fuel cermet contains 20 to 60% UO2 by weight in the form of spherical particles having a diameter greater than 50 μm. The grade of steel for the cermet can be chosen relative to the conditions in which it is to be used. The fuel elements are checked by gammagraphy, macrography and micrography. The thermal conductivity and mechanical tensile properties of the cermets were investigated with respect to UO2 content and particle granulometry. Thermal shock tests and lengthy annealing of the tubular elements caused no cracking or lessening of the can. By this method of extrusion sintering canned fuel elements can be obtained directly; handling of the UO2 powder is reduced to a minimum, so that the losses of enriched uranium in the industrial fabrication of the fuel elements can also be reduced.

  2. A study on end closure of fuel element cladding of PHWRs by GTA welding

    International Nuclear Information System (INIS)

    Zirconium alloys are extensively used in thermal reactors as cladding materials. Fuel material is enclosed in cladding tube and welding with plugs close ends of tube. There are various methods for end closure welding of these fuel elements. Resistance welding (RW) process has been widely used for PHWRs fuel elements. Resistance welds are made rapidly and it is suitable for large-scale production. This process has certain limitations. To explore alternate avenues for superior quality end closure welds of PHWR fuel elements, a study is undertaken of different welding processes. GTAW was chosen for further work since it has well-established practice. New generation GTAW power sources enhance the quality and reliability of welds. A major advantage of GTAW is that the process, equipment and operators can be deployed for any type of fuel elements e.g. BWRs, PWRs, PHWRs as well as AHWRs. Experiments were conducted and the process was studied for Φ15.3 mm OD and 0.4 mm wall thickness tubes. Welding parameters have been optimised. Joint quality has been assessed with respect to specifications of PHWRs fuel elements. Welds were evaluated by NDT as well as destructive methods. The paper describes the GTAW process, selection of power source, design of welding chamber, parameters, problems faced, weld defects observed during standardisation of parameters and their elimination. (author)

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

    International Nuclear Information System (INIS)

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

  4. Automation of remote handling in uranium and mixed oxide fuel element fabrication plants

    International Nuclear Information System (INIS)

    The subject of the analyses are plants for the fabrication or uranium oxide and uranium-plutonium mixed oxide fuel elements. The reference basis of the paper is an overview of the state-of-the-art of manufacturing technologies with regard to automation and remote handling during fuel element fabrication in national and foreign plants, and in comparabel sectors of conventional technologies. Proceeding from ambient dose rates, residence times, and technical conditions or individual doses at typical work-places during fuel element fabrication, work processes are pointed out which, taking into account technical possibilities, should be given priority when automating, and technical solutions for it are sought. Advantages and disadvantages of such measures are outlined, and reduction of radiation exposure is shown (example: mixed oxide fuel fabrication plant at Hanau). (orig./HP)

  5. Early encapsulation of spent fuel elements. Separation of requirements from operation and disposal

    International Nuclear Information System (INIS)

    The design requirements to be met by fuel elements in the light of the trend towards higher burnups are aggravated by the need to take into account the logistics of decommissioning. The criteria applying to operation should be separated from those applying to disposal this can be achieved by early encapsulation of spent fuel elements. This creates the flexibility now for decisions to be taken decades from now. A few relatively simple process step will hermetically seal spent fuel elements in the storage pools of nuclear power plants. Only the capsules are handled until the decision about the spent fuel management pathway is taken. Should direct disposal be chosen, only capsules will be handled right to the repository. This process of early encapsulation not only provides more flexibility, but also offers major economic advantages. (orig.)

  6. Application of replication techniques to fractography of irradiated uranium fuel elements

    International Nuclear Information System (INIS)

    Radiation effects in nuclear fuel elements were studied by fractography. Fuel elements were experimentally irradiated at various doses at the IRR-2. Irradiated, as well as unirradiated fuel elements, were bent till fracture with a bending press. The fractured surfaces were examined by standard replication techniques. The structures reveal the same morphological zones as in standard tensile fractures. The crack origin zone (fibrous zone) appears near the fuel indent. The radial marks zone starts near the origin and covers about 2/3 of the cross-sectional area. The third zone (shear rupture zone) has a bump, sometimes 1 cm high. The crack origin zone is mainly brittle. As the crack propagates in the radial marks zone it becomes more ductile until its entirely ductile at the bump top. Beyond the bump it turns brittle again. (author)

  7. Special equipment for the fabrication and quality control of nuclear fuel elements

    International Nuclear Information System (INIS)

    For the fabrication of LWR fuel elements, columns are packed of up to 4 m in length, consisting of fuel pellets with different uranium enrichment, their weight and total length to be measured prior to further processing to fuel rods. An automated column packing device has been developed for this purpose. The packing jobs and other tasks are computer-controlled, measured data are stored and are printed out for quality documentation. The forces in the springs of fuel spacers of LWR fuel elements are to be measured and compared with the standard requirements, deviations to be documented. For this task, another computer-controlled, automated device has been developed for measuring the spring forces at all required positions after positioning and fixation of the spacers. (orig./DG)

  8. HTGR fuel elements operating conditions during accidents with abrupt power raise

    International Nuclear Information System (INIS)

    The necessity of the investigations for developing of HTGR fuel elements operability criteria, connected with the specific energy release values and the rates of its change in fuel is demonstrated in the paper on the example of the accident with positive reactivity increase at VGM reactor pebble bed compression as a result of seismic impact. It is shown, that the average fuel enthalpy over the core in this accident with the emergency protection failure may reach ∼24 Kj/g U02, and the maximum rate of its increase is about 0.14 Kj/g.s. It considerably exceeds the established limit of fuel enthalpy for LWR fuel elements. (author). 5 refs, 2 figs

  9. Fabrication of U3 Si2 based fuel elements in Brazil

    International Nuclear Information System (INIS)

    IPEN - Brazilian Institute of Nuclear and Energy Research, has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has manufacturing its own fuel element, initially based on U3O8-Al dispersion fuel plates with 2.3 gU/cm3. To support the reactor power increase, higher uranium density had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicide was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and qualification. (author)

  10. 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.)

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

    Science.gov (United States)

    Tondin, Julio Benedito Marin; Filho, Tufic Madi

    2011-08-01

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

  12. Implementation of the utilization program for the fuel elements of the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    The programming operation for the use of the fuel elements in the Atucha-1 nuclear power plant was initially under the responsibility of the KWU Company, as part of the services rendered due for the manufacturing of said elements. This job was done with the help of the TRISIC program, developed in the early seventies by CNEA and SIEMENS staff. From april 21, 1979 on, CNEA took over the responsibility and strategy of the interchange of fuel elements. The several stages carried out for the implementation of this service are detailed. (M.E.L.)

  13. A modelling study for the determination of temperature distribution in a WWER-1000 type fuel element

    International Nuclear Information System (INIS)

    A short description is presented of the fuel element of the WWER-1000 type nuclear reactor. A physical model of the fuel element was developed in order to determine the temperature distribution within the element. The numerical solution of the mathematical model based on the physical model was performed by using the one-dimensional modular program HOVID. The influence of simplifications introduced in the mathematical model on the calculated temperature distribution is discussed. Finally, the results of steady-state calculations are analyzed. (R.P.) 11 refs.; 7 figs.; 4 tabs

  14. A CFD numerical model for the flow distribution in a MTR fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Andrade, Delvonei Alves de; Santos, Pedro Henrique Di Giovanni; Oliveira, Fabio Branco Vaz de; Torres, Walmir Maximo; Umbehaun, Pedro Ernesto; Souza, Jose Antonio Batista de; Belchior Junior, Antonio; Sabundjian, Gaiane; Prado, Adelk de Carvalho, E-mail: acprado@ipen.br, E-mail: delvonei@ipen.br, E-mail: dpedro_digiovanni_s@hotmail.com, E-mail: fabio@ipen.br, E-mail: wmtorres@ipen.br, E-mail: umbehaun@ipen.br, E-mail: jasouza@ipen.br, E-mail: abelchior@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Angelo, Edvaldo, E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, Sao Paulo, SP (Brazil); Angelo, Gabriel, E-mail: gangelo@fei.edu.br [Fundacao Educacional Inaciana (FEI), Sao Bernardo do Campo, SP (Brazil)

    2015-07-01

    Previously, an instrumented dummy fuel element (DMPV-01), with the same geometric characteristics of a MTR fuel element, was designed and constructed for pressure drop and flow distribution measurement experiments at the IEA-R1 reactor core. This dummy element was also used to measure the flow distribution among the rectangular flow channels formed by element fuel plates. A CFD numerical model was developed to complement the studies. This work presents the proposed CFD model as well as a comparison between numerical and experimental results of flow rate distribution among the internal flow channels. Numerical results show that the model reproduces the experiments very well and can be used for the studies as a more convenient and complementary tool. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  16. AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

    International Nuclear Information System (INIS)

    Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700 C. This temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelength dispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700 C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction

  17. An Evaluation of Potential Liner Materials for Eliminating FCCI in Irradiated Metallic Nuclear Fuel Elements

    International Nuclear Information System (INIS)

    Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting long live transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700 deg. C. This temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelength dispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700 deg. C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction. (authors)

  18. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    International Nuclear Information System (INIS)

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles

  19. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Z., E-mail: chengz@aecl.ca; Rao, Y.F., E-mail: raoy@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-10-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles.

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

    International Nuclear Information System (INIS)

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

  1. Oxide fuel element and blanket element development programs. Quarterly progress report, April-June 1978

    International Nuclear Information System (INIS)

    Approval-in-principle has been granted for run beyond breach experiment XY-2, which will incorporate an F11A series rod. Fuel microstructures and operating parameters have been tabulated for 118 specimens from the F20 power to melt experiment. Retained gas measurements have been compiled indicating 36-50 μl/gm in this high power fuel. Topical report GEFR-00367 was prepared describing F20 results. Preparation of the Test Design Description for axial blanket experiment AB-1 is proceeding on schedule (for Cycle 2 irradiation). The safety analysis calculations, showing no fuel melting nor sodium boiling in design-basis upsets, have been completed

  2. Mechanical separation process for decladding of LWR fuel elements

    International Nuclear Information System (INIS)

    A comparison of the advantages and disadvantages of known methods of decladding led to cavitation erosion being used as a decladding mechanism. This process attacks not the jacket of the fuel rod but the fuel itself. Cavitation erosion is the consequence of imploding vapour bubbles entailing dynamic stress of a high frequency and high amplitude. The separation effect is due to the different material properties. Ductile materials as a rule are much more resistant to dynamic stress than brittle materials. Systematic experiments at varying pressures, volume flow, nozzle geometries and distances between nozzle and sample led to optimized parameters. There was a conspicuous rise in the relations pressure to depth of erosion and volume flow to depth of erosion. This considered, p=700 bar and d=1.6 mm were found to be useful parameters. The relation of the distance from nozzle to sample and the erosion obtained also has an optimum at s=50 mm. This distance can be shortened in the course of the operation. A great entrance angle combined with a nozzle outlet channel of the length l=1/2 D improves the erosion result considerably. The attack of the cavitating water jet on the jacket of the fuel rod causes a weight loss of <=2per mille. (orig./HP)

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

    International Nuclear Information System (INIS)

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

  4. Choices of canisters and elements for the first fuel shipment from K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J.

    1995-03-01

    Twenty-two canisters (10 prime and 12 backup candidates) in the K West Basin have been identified as containing fuel which, when examined, will satisfy the Data Quality Objectives for the first fuel shipment from this basin. These were chosen as meeting criteria such as containing relatively long fuel elements, locking bar integrity, and the availability of gas/liquid interface level measurements for associated canister gas traps. Two canisters were identified as having reported broken fuel on initial loading. Usage and interpretation of canister cesium concentration measurements have also been established and levels of maximum and minimum acceptable cesium concentration (from a data optimization point of view) for decapping have been determined although other operational cesium limits may also apply. Criteria for picking particular elements, once a canister is opened, are reviewed in this document. A pristine, a slightly damaged, and a badly damaged element are desired. The latter includes elements with end caps removed but does not include elements which have large amounts of swelling or split cladding that might interfere with handling tools. Finally, operational scenarios have been suggested to aid in the selections of canisters and elements in a way that utilizes anticipated canister gas sampling and leads to a correct and quick choice of elements which will supply the desired data.

  5. Trace elements retained in washed nuclear fuel reprocessing solvents

    International Nuclear Information System (INIS)

    Analysis of purified TBP extractant from solvent extraction processes at Savannah River Plant showed several stable elements and several long-lived radioisotopes. Stable elements Al, Na, Br, Ce, Hg, and Sm are found in trace quantities in the solvent. The only stable metallic element consistently found in the solvent was Al, with a concentration which varies from about 30 ppM to about 10 ppM. The halogens Br and Cl appear to be found in the solvent systems as organo halides. Radionuclides found were principally 106Ru, 129I, 3H, 235U, and 239Pu. The 129I concentration was about 1 ppM in the first solvent extraction cycle of each facility. In the other cycles, 129I concentration varied from about 0.1 to 0.5 ppM. Both 129I and 3H appear to be in the organic solvent as a result of exchange with hydrogen

  6. Oxide fuel element and blanket element development programs. Quarterly progress report, January-February-March, 1979

    International Nuclear Information System (INIS)

    Fuel pin profilometry of some 9% burnup F20-F5 pins showed small diameter increases at the fuel-insulator interface at the top of the core. Neither these secondary peaks nor the larger diameter increases near the core midplane exhibited any relationship to the local presence of once-molten fuel in any F20 fuel pin. Augmented safety analysis computations for experiment AB-1 (additional transients suggested by HEDL) showed that cumulative damage fractions from the additional transients were in every case less than 10-4. Mechanical tests have been performed that confirm previous computations for the removal end plugs to be used in a characterizer subassembly for AB-1. The resulting pin removal forces are well within the design envelope

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW(th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post- irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Microstructural characterization by metallographic analyses; (iii) Determination of mechanical properties; (iv) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

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

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW (th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Gamma scanning and tomography; (iii) Measurement of pressure, volume and isotopic composition of fission gas; (iv) Microstructural characterization by metallographic analyses; (v) Determination of mechanical properties; amd (vi) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  11. The nuclear fuel elements' world market and the position of the Argentine Republic as producer

    International Nuclear Information System (INIS)

    The development of the nuclear fuel elements' industry is analyzed, both in the present and projected world market, up to the year 2000, in the light of the situation affecting the nucleoelectric industry. By means of the offer/demand function, an analysis is made of the behaviour of the fuel elements' market throughout the fuel cycle structure. The regional unbalances between availability and demand of uranium resources are considered, as well as the factors having an unfavorable incidence on the fuel cycle's economic equation. The economic structure to be used for the calculation of the nucleoelectric generating cost is presented, in order to situate, within said nuclear economy, the component corresponding to the fuel cycle cost. Emphasis is placed on the 'front end' stages of the fuel cycle, but also considering those stages belonging to the 'back end'. Argentina's fuel elements market and its present and projected nucleoelectric park are analyzed, indicating their relative position in the world market. (R.J.S.)

  12. 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

  13. Automatic X-ray inspection for the HTR-PM spherical fuel elements

    International Nuclear Information System (INIS)

    Highlights: • An automatic X-ray inspection method is established to characterize HTR pebbles. • The method provides physical characterization and the inner structure of pebbles. • The method can be conducted non-destructively, quickly and automatically. • Sample pebbles were measured with this AXI method for validation. • The method shows the potential to be applied in situ. - Abstract: Inefficient quality assessment and control (QA and C) of spherical fuel elements for high temperature reactor-pebblebed modules (HTR-PM) has been a long-term problem, since conventional methods are labor intensive and cannot reveal the inside information nondestructively. Herein, we proposed a nondestructive, automated X-ray inspection (AXI) method to characterize spherical fuel elements including their inner structures based on X-ray digital radiography (DR). Briefly, DR images at different angles are first obtained and then the chosen important parameters such as spherical diameters, geometric and mass centers, can be automatically extracted and calculated via image processing techniques. Via evaluating sample spherical fuel elements, we proved that this AXI method can be conducted non-destructively, quickly and automatically. This method not only provides accurate physical characterization of spherical fuel elements but also reveals their inner structure with good resolution, showing great potentials to facilitate fast QA and C in HTM-PM spherical fuel element development and production

  14. Standard laboratory hydraulic pressure drop characteristics of various solid and I&E fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Waters, E.D.; Horn, G.R.

    1958-01-20

    The purpose of this report is to present a set of standard pressure-drop curves for various fuel elements in process tubes of Hanford reactors. The flow and pressures within a process tube assembly under normal conditions are dependent to a large extent on the magnitude of the pressure drop across the fuel elements. The knowledge of this pressure drop is important in determination of existing thermal conditions within the process tubes and in predicting conditions for new fuel element designs or changes in operating conditions. The pressure-flow relations for the different Hanford fuel element-process tube assemblies have all been determined at one time or another in the 189-D Hydraulics Laboratory but the data had never been collected into a single report. Such a report is presented now in the interest of establishing a set of ``standard curves`` as determined by laboratory investigations. It must be recognized that the pressure drops of fuel elements in actual process tubes in the reactors may be slightly different than those reported here. The data presented here were obtained in new process tubes while reactor process tubes are usually either corroded or filmed, depending on their past history.

  15. Parabolic k-ample bundles

    CERN Document Server

    Biswas, Indranil

    2011-01-01

    We construct projectivization of a parabolic vector bundle and a tautological line bundle over it. It is shown that a parabolic vector bundle is ample if and only if the tautological line bundle is ample. This allows us to generalize the notion of a k-ample bundle, introduced by Sommese, to the context of parabolic bundles. A parabolic vector bundle $E_*$ is defined to be k-ample if the tautological line bundle ${\\mathcal O}_{{\\mathbb P}(E_*)}(1)$ is $k$--ample. We establish some properties of parabolic k-ample bundles.

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

    International Nuclear Information System (INIS)

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

  17. Performance analysis of covered conductors and aerial bundled cables by the finite elements method; Analise de desempenho de cabos cobertos e pre-reunidos pelo metodo dos elementos finitos

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Joao Jose dos Santos

    1996-12-31

    The covered conductors and the aerial bundled cables (abc) for overhead distribution systems 15 kV systems, has been a good solution for the main brazilian electrical utilities when applied in areas with to limited space on poles and to satisfy the performance claimed in tree areas or narrow streets, focusing people safety and electrical service continuity. Otherwise, in this work it is presented that these cables show failures that can be related to the solid isolants used. It is showed the laboratory tests results to characterize its thermal and electric behaviour and an application of the Finite Element Method through the software Flux 2 d was done. The results obtained showed the requirements to be demanded through the specification and proposals elements to change the standard. (author) 57 refs., 20 figs., 73 tabs.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The design of modern HTRs is based on high quality fuel. A research and development programme has demonstrated the satisfactory performance in fuel manufacturing, irradiation testing and accident condition testing of irradiated fuel elements. This report describes the fuel particles with their low-enriched UO2 kernels and TRISO coating, i.e. a sequence of pyrocarbon, silicon carbide, and pyrocarbon coating layers, as well as the spherical fuel element. Testing was performed in a generic programme satisfying the requirements of both the HTR-MODUL and the HTR 500. With a coating failure fraction less than 2x10-5 at the 95% confidence level, the results of the irradiation experiments surpassed the design targets. Maximum accident temperatures in small, modular HTRs remain below 1600deg C, even in the case of unrestricted core heatup after depressurization. Here, it was demonstrated that modern TRISO fuels retain all safety-relevant fission products and that the fuel does not suffer irreversible changes. Isothermal heating tests have been extended to 1800deg C to show performance margins. Ramp tests to 2500deg C demonstrate the limits of present fuel materials. A long-term programm is planned to improve the statistical significance of presently available results and to narrow remaining uncertainty limits. (orig.)

  1. FEM analysis of the rolled bottom nozzle of 16NGF fuel element

    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 Rolled Bottom Nozzle (previously casting) used in the fuel element 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 BEAM4. 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 Bottom Nozzle is assured when the design criteria defined in the ASME Boiler and Pressure Vessel Code Section III is satisfied. The results of these analyses were used to prove that the Rolled Bottom 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 Elements. 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)

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

    Institute of Scientific and Technical Information of China (English)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

  4. RA3: Application of a calculation model for fuel management with SEFE (Slightly Enriched Fuel Elements)

    International Nuclear Information System (INIS)

    The RA-3 (5 MW, MTR) reactor is mainly utilized to produce radioisotopes (Mo-99, I-131, etc.). It started operating with Low Enrichment Uranium (LEU) in 1990, and spends around 12 fuels per year. Although this consumption is small compared to a nuclear power station. It is important to do a good management of them. The present report describes: - A reactor model to perform the Fuel Shuffling. - Results of fuel management simulations for 2 and a half years of operation. Some features of the calculations can be summarized as follows: 1) A 3D calculation model is used with the code PUMA. It does not have experimental adjustments, except for some approximations in the reflector representation and predicts: power, flux distributions and reactivity of the core in an acceptable way. 2) Comparisons have been made with the measurements done in the commissioning with LEU fuels, and it has also been compared with the empirical method (the previous one) which had been used in the former times of operation with LEU fuel. 3) The number of points of the model is approximately 13500, an it can be run in 80386 personal computer. The present method has been verified as a good tool to perform the simulations for the fuel management of RA-3 reactor. It is expected to produce some economic advantages in: - Achieving a better utilization of the fuels. - Leaving more time of operation for radioisotopes production. The activation measurements through the whole core required by the previous method can be significantly reduced. (author)

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

    International Nuclear Information System (INIS)

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

  6. Determination of the fuel element burn-up for mixed TRIGA core by measurement and calculation with new TRIGLAV code

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, T.; Ravnik, M.; Persic, A. (J.Stefan Institute, Ljubljana (Slovenia))

    1999-12-15

    Results of fuel element burn-up determination by measurement and calculation are given. Fuel element burn-up was calculated with two different programs TRIGLAV and TRIGAC using different models. New TRIGLAV code is based on cylindrical, two-dimensional geometry with four group diffusion approximation. TRIGAC program uses one-dimensional cylindrical geometry with twogroup diffusion approximation. Fuel element burn-up was measured with reactivity method. In this paper comparison and analysis of these three methods is presented. Results calculated with TRIGLAV show considerably better alignment with measured values than results calculated with TRIGAC. Some two-dimensional effects in fuel element burn-up can be observed, for instance smaller standard fuel element burn-up in mixed core rings and control rod influence on nearby fuel elements. (orig.)

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

    International Nuclear Information System (INIS)

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

  8. Contact fiber bundles

    OpenAIRE

    Lerman, Eugene

    2003-01-01

    We define contact fiber bundles and investigate conditions for the existence of contact structures on the total space of such a bundle. The results are analogous to minimal coupling in symplectic geometry. The two applications are construction of K-contact manifolds generalizing Yamazaki's fiber join construction and a cross-section theorem for contact moment maps

  9. Principal noncommutative torus bundles

    DEFF Research Database (Denmark)

    Echterhoff, Siegfried; Nest, Ryszard; Oyono-Oyono, Herve

    2008-01-01

    action) and give necessary and sufficient conditions for any non-commutative principal torus bundle being RKK-equivalent to a commutative one. As an application of our methods we shall also give a K-theoretic characterization of those principal torus-bundles with H-flux, as studied by Mathai...... and Rosenberg which possess "classical" T-duals....

  10. Additional materials on the safety assessment for the 24 CANFLEX-NU bundle demonstration irradiation at Wolsong-1 generating station as the answers on the KINS's second questions

    International Nuclear Information System (INIS)

    This report describes the additional materials on the Safety Assessment for the 24 CANFLEX-NU Bundle Demonstration Irradiation at Wolsong-1 Generating Station, which is the answers on the KINS's first questions as a result of the review of the Safety Assessment report. The additional materials cover : 1) Subjects on CANFLEX-NU fuel channel and the selection criteria, 2) Subjects on the evaluation of the discharge fuel with respect to the purpose of the DI, 3) Subjects on the planning of report of the DI results, 4) Subjects on the integrity evaluation of the CANFLEX-NU fuel bundles in the reactor, 5) Subjects on the thermal-hydraulic compatibility of the CANFLEX-NU fuel with the existing 37-element bundles, 6) Subjects on the Input variables of CATHENA code used in the safety analyses, 7) Subjects on the loss of forced circulation, 8) Subjects on the CANFLEX-NU fuel element power under condition of non-LOCA, 9) Subjects on the loss of reactivity control, 10) Subjects on the LOCA. The additional materials are made on the basis of the Safety Assessment for the 24 CANFLEX-NU Bundle Demonstration Irradiation at Wolsong-1 Generating Station(H. C. Suk et. al, KAERI/TR-1864/2001, 2001.6) and Operational Procedure of the Wolsong-1 Generating Station, and so on

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

    International Nuclear Information System (INIS)

    The design, fabrication, irradiation results, and post-irradiation examination for three superheater test fuel elements are described. During the spring of 1966 these clusters, each consisting of six fuel rods, were successfully exposed in the superheater loop No. 5 in the R2 reactor for a maximum of 24 days at a maximum outer cladding surface temperature of ∼ 650 deg C. During irradiation the linear heat rating of the rods was in the range 400-535 W/cm. The diameter of the UO2 pellets was 11.5 and 13.0 mm; the wall thickness of the 20/25 Nb and 20/35 cladding was in every case 0.4 mm. The diametrical gap between fuel and cladding was one of the main parameters and was chosen to be 0.05, 0.07 and 0.10 mm. These experiments, to be followed by one high cladding temperature irradiation (∼ 750 deg C) and one long time irradiation (∼ 6000 MWd/tU), were carried out to demonstrate the operational capability of short superheater test fuel rods at steady and transient operational environments for the Marviken superheater fuel elements and also to provide confirmation of design criteria for the same fuel elements

  12. New Nuclear Materials Including Non Metallic Fuel Elements. Vol. I. Proceedings of the Conference on New Nuclear Materials Technology, Including Non Metallic Fuel Elements

    International Nuclear Information System (INIS)

    One of the major aims of the International Atomic Energy Agency in furthering the peaceful uses of atomic energy is to encourage the development of economical nuclear power. Certainly, one of the more obvious methods of producing economical nuclear power is the development of economical fuels that can be used at high temperatures for long periods of time, and which have sufficient strength and integrity to operate under these conditions without permitting the release of fission products. In addition it is desirable that after irradiation these new fuels be economically reprocessed to reduce further the cost of the fuel cycle. As nuclear power becomes more and more competitive with conventional power the interest in new and more efficient higher-temperature fuels naturally increases rapidly. For these reasons, the Agency organized a Conference on New Nuclear Materials Technology, Including Non-Metallic Fuel Elements, which was held from 1 to 5 July 1963 at the International Hotel, Prague, with the assistance and co-operation of the Government of the Czechoslovak Socialist Republic. A total of 151 scientists attended, from 23 countries and 4 international organizations. The participants heard and discussed more than 60 scientific papers

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  15. Grain growth and temperature distribution in the irradiated fuel elements designed for Qinshan Power Plant

    International Nuclear Information System (INIS)

    The micro-analysis results of UO2 fuel core of irradiated test elements designed for Qinshan Power Plant are presented. The temperature of element centers is quantitatively calculated by using the grain size measurement results according to the principle of grain growth dynamics, and the results are in accordance with that of the FRAP-CON calculation. Based on the post-irradiation examination data, the dynamic parameters, K and Q which are particularly suitable to description of irradiation behaviours of UO2 fuel made in China are deduced

  16. Reactivity change measurements on plutonium-uranium fuel elements in hector experimental techniques and results

    International Nuclear Information System (INIS)

    The techniques used in making reactivity change measurements on HECTOR are described and discussed. Pile period measurements were used in the majority of oases, though the pile oscillator technique was used occasionally. These two methods are compared. Flux determinations were made in the vicinity of the fuel element samples using manganese foils, and the techniques used are described and an error assessment made. Results of both reactivity change and flux measurements on 1.2 in. diameter uranium and plutonium-uranium alloy fuel elements are presented, these measurements being carried out in a variety of graphite moderated lattices at temperatures up to 450 deg. C. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-11-01

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

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

    International Nuclear Information System (INIS)

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

  19. Determination of the burn-up of TRIGA fuel elements by calculation and reactivity experiments

    International Nuclear Information System (INIS)

    The burnup of 17 fuel elements of the TRIGA Mark-II reactor in Vienna was measured. Different types of fuel elements had been simultaneously used for several years. The measured burnup values are compared with those calculated on the basis of core configuration and reactor operation history records since the beginning of operation. A one-dimensional, two-group diffusion computer code TRIGAP was used for the calculations. Comparison with burnup values determined by γ-scanning is also made. (orig./HP)

  20. Study of causes of can leakage in original WWER and RBMK fuel elements

    International Nuclear Information System (INIS)

    The post-irradiation examination is accomplished for four WWER-1000 spent fuel assemblies (SFA) and three RBMK-1000 SFAs accepted to be failed in the course of inspection. From the investigation results and in accordance with the literature data the following explanation of the reasons for clad leakage is given: local mechanical damages (primary defects) of fuel cans arise from the interaction with debris entered the coolant circuit to the extent that through holes can be formed. The penetration of the coolant into a fuel element results in hydrogenation and embrittlement of various sections of fuel cans (secondary defects). Primary defects are noted to be of frequent occurrence near spacer grids. To reveal defective WWER-1000 fuel assemblies in operation it is necessary that the data on the primary circuit coolant activity should be taken into account

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

    Science.gov (United States)

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

    2010-02-23

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

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

    Science.gov (United States)

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

    2013-09-03

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

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

    Science.gov (United States)

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

    2011-03-01

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

  4. Feasibility study of modeling a CANDU fuel element using a multiphysics object-oriented simulation environment

    International Nuclear Information System (INIS)

    The first phase of the feasibility study of using a Multiphysics Object-Oriented Simulation Environment (MOOSE) for modeling a CANDU fuel element is presented. A two-dimensional model of a fuel pellet sheath was created to examine the contact algorithm within MOOSE. The results obtained show the expected behaviour of contact pressure and penetration in 2D. Preliminary results for a 3D model of a quarter fuel pellet and sheath are provided but at present contain anomalies currently being investigated. The next steps in the feasibility study are outlined. (author)

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

    International Nuclear Information System (INIS)

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

  6. Feasibility study of modeling a CANDU fuel element using a multiphysics object-oriented simulation environment

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K., E-mail: Kyle.Gamble@rmc.ca [Royal Military College of Ontario, Kingston, Ontario (Canada); Williams, A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Chan, P.K. [Royal Military College of Ontario, Kingston, Ontario (Canada)

    2013-07-01

    The first phase of the feasibility study of using a Multiphysics Object-Oriented Simulation Environment (MOOSE) for modeling a CANDU fuel element is presented. A two-dimensional model of a fuel pellet sheath was created to examine the contact algorithm within MOOSE. The results obtained show the expected behaviour of contact pressure and penetration in 2D. Preliminary results for a 3D model of a quarter fuel pellet and sheath are provided but at present contain anomalies currently being investigated. The next steps in the feasibility study are outlined. (author)

  7. Deformations of the generalised Picard bundle

    International Nuclear Information System (INIS)

    Let X be a nonsingular algebraic curve of genus g ≥ 3, and let Mξ denote the moduli space of stable vector bundles of rank n ≥ 2 and degree d with fixed determinant ξ over X such that n and d are coprime. We assume that if g = 3 then n ≥ 4 and if g = 4 then n ≥ 3, and suppose further that n0, d0 are integers such that n0 ≥ 1 and nd0 + n0d > nn0(2g - 2). Let E be a semistable vector bundle over X of rank n0 and degree d0. The generalised Picard bundle Wξ(E) is by definition the vector bundle over Mξ defined by the direct image pMξ *(Uξ x pX*E) where Uξ is a universal vector bundle over X x Mξ. We obtain an inversion formula allowing us to recover E from Wξ(E) and show that the space of infinitesimal deformations of Wξ(E) is isomorphic to H1(X, End(E)). This construction gives a locally complete family of vector bundles over Mξ parametrised by the moduli space M(n0,d0) of stable bundles of rank n0 and degree d0 over X. If (n0,d0) = 1 and Wξ(E) is stable for all E is an element of M(n0,d0), the construction determines an isomorphism from M(n0,d0) to a connected component M0 of a moduli space of stable sheaves over Mξ. This applies in particular when n0 = 1, in which case M0 is isomorphic to the Jacobian J of X as a polarised variety. The paper as a whole is a generalisation of results of Kempf and Mukai on Picard bundles over J, and is also related to a paper of Tyurin on the geometry of moduli of vector bundles. (author)

  8. Final disposal of dissolver sludges, claddings, fuel hardware, and spent HTGR fuel elements in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    In the Federal Republic of Germany, radioactive waste of marked decay heat production shall be disposed of in vertical 300-m-deep boreholes in a salt dome formation. The disposal technique for heat generating intermediate-level wastes arising from LWR fuel reprocessing and for spent HTGR pebble bed fuel is referred to as the ILW borehole technique. This technique is under development since 1983 in a R and D project for intermediate-level waste and spent HTGR fuel element test disposal in boreholes. This paper reports on the main objectives of this project, to establish the ILW borehole technique by developing the emplacement and plugging concept up to design and testing of prototype components and testing of the multicomponent technique under real conditions in the ASSE salt mine

  9. FRANCO, Finite Element Method (FEM) Fuel Rod Analysis for Solid and Annular Configurations

    International Nuclear Information System (INIS)

    1 - Description of program or function: The FRANCO code is a quasi- static two-dimensional fuel rod analysis code, that calculates the fuel temperature and material deformation as a function of heat generation rate. Both solid and annular fuel configurations are modeled. 2 - Method of solution: FRANCO uses two-dimensional finite element theory and applications for mechanical deformation and heat conduction, and determines the temperature distribution from the fuel center to the coolant adjacent to the clad at a position along the fuel rod axis. FRANCO calculates the average temperature of each radial division, the nodal displacement, and strain and stress within the fuel pellet and clad. The principal stresses, which represent maximum and minimum stresses within an element, result from Mohr's circle relationship between normal stresses. FRANCO is capable of predicting the thermo-mechanical behavior in the radial direction of a single fuel rod for both boiling water reactors (BWR's) and pressurized water reactors (PWR's). The cross sectional plane geometry of fuel rod is modeled using three-node constant strain triangular finite elements, and both thermal and mechanical solutions are computed with the same finite element configurations. The local linear heat generation rate is modeled as a uniform heat source in a fuel pellet, and the coolant temperature and heat transfer coefficient are applied as known boundary conditions at the boundary of the cladding surface. The total load to form the global force vector consists of the thermal load that results from thermal expansion of the material and the mechanical load exerted by pressure. FRANCO assumes the fuel-cladding gap region to be conductive material in order to simplify the analysis, and this gap is simulated by either an open gap or a closed gap model. A time- dependent problem can be simulated by FRANCO using quasi-static analysis when time-dependent parameters are provided. FRANCO can treat a steady-state or

  10. Current research projects at the Austrian TRIGA Mark II. Location of failed fuel elements in Austrian TRIGA Mark II

    International Nuclear Information System (INIS)

    The system developed at the Atominstitut monitors the radioactive Krypton- and Xenon nuclides in the primary water circuitry and allows selective control of any fuel element for its fission gas release. A suspected fuel element is enclosed in an underwater capsule attached in the reactor tank. Water is pumped along the fuel element to a vacuum degasser where the gases are separated from the tank water. The degassed water is returned to the reactor pool while the gases are pumped to a very sensitive proportional counter. The fuel elements of the TRIGA core were checked by the described procedure

  11. Annular burnout data from rod-bundle experiments

    International Nuclear Information System (INIS)

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident

  12. Trace elements in co-combustion of solid recovered fuel and coal

    DEFF Research Database (Denmark)

    Wu, Hao; Glarborg, Peter; Jappe Frandsen, Flemming;

    2013-01-01

    Trace element partitioning in co-combustion of a bituminous coal and a solid recovered fuel (SRF) was studied in an entrained flow reactor. The experiments were carried out at conditions similar to pulverized coal combustion, with SRF shares of 7.9 wt.% (wet basis), 14.8 wt.% and 25.0 wt.%. In ad......Trace element partitioning in co-combustion of a bituminous coal and a solid recovered fuel (SRF) was studied in an entrained flow reactor. The experiments were carried out at conditions similar to pulverized coal combustion, with SRF shares of 7.9 wt.% (wet basis), 14.8 wt.% and 25.0 wt......-based additives increased the volatility of Cd, Pb and As, whereas addition of ammonium sulphate generally decreased the volatility of trace elements. Addition of kaolinite reduced the volatility of Pb, while the influence on other trace elements was insignificant. The results from the present work imply...

  13. Annular burnout data from rod bundle experiments

    International Nuclear Information System (INIS)

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident. Level average fluid conditions within the test section were calculated using steady-state mass and energy conservation considerations for the steady-state tests and a transient, homogeneous, equilibrium computer code for the transient tests. Unlike tube dryout, burnout within a rod bundle does not necessarily occur at one distinct axial level. The location of individual rod dryout was determined by scanning rods axially and locating the position where rod superheat increased from approx. =0 to 30 K or greater. Thermocouple instrumentation within the bundle allows the location of dryout to be determined to within approximately +.5 cm for many of the tests

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

    International Nuclear Information System (INIS)

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

  15. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    International Nuclear Information System (INIS)

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles

  16. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-09-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles.

  17. Postirradiation examination of a low enriched U3Si2-Al fuel element manufactured and irradiated at Batan, Indonesia

    International Nuclear Information System (INIS)

    The first low-enriched U3Si2-Al dispersion plate-type fuel element produced at the Nuclear Fuel Element Center, BATAN, Indonesia, was irradiated to a peak 235U burnup of 62%. Postirradiation examinations performed to data shows the irradiation behavior of this element to be similar to that of U3Si2-Al plate-type fuel produced and tested at other institutions. The main effect of irradiation on the fuel plates is a thickness increase of 30--40 μm (2.5-3.0%). This thickness increase is almost entirely due to the formation of a corrosion layer (Boehmite). The contribution of fuel swelling to the thickness increase is rather small (less than 10 μm) commensurate with the burnup of the fuel and the relatively moderate as-fabricated fuel volume fraction of 27% in the fuel meat

  18. 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)

  19. Acceptance of spent nuclear fuel in multiple element sealed canisters 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: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs

  20. New Nuclear Materials Including Non Metallic Fuel Elements. Vol. II. Proceedings of the Conference on New Nuclear Materials Technology, Including Non Metallic Fuel Elements

    International Nuclear Information System (INIS)

    One of the major aims of the International Atomic Energy Agency in furthering the peaceful uses of atomic energy is to encourage the development of economical nuclear power. Certainly, one of the more obvious methods of producing economical nuclear power is the development of economical fuels that can be used at high temperatures for long periods of time, and which have sufficient strength and integrity to operate under these conditions without permitting the release of fission products. In addition it is desirable that after irradiation these new fuels be economically reprocessed to reduce further the cost of the fuel cycle. As nuclear power becomes more and more competitive with conventional power the interest in new and more efficient higher-temperature fuels naturally increases rapidly. For these reasons, the Agency organized a Conference on New Nuclear Materials Technology, Including Non-Metallic Fuel Elements, which was held from 1 to 5 July 1963 at the International Hotel, Prague, with the assistance and co-operation of the Government of the Czechoslovak Socialist Republic. A total of 151 scientists attended, from 23 countries and 4 international organizations. The participants heard and discussed more than 60 scientific papers. The Agency wishes to thank the scientists who attended this Conference for their papers and for many spirited discussions that truly mark a successful meeting. The Agency wishes also to record its gratitude for the assistance and generous hospitality accorded the Conference, the participants and the Agency's staff by the Government of the Czechoslovak Socialist Republic and by the people of Prague. The scientific information contained in these Proceedings should help to quicken the pace of progress in the fabrication of new and m ore economical fuels, and it is hoped that these proceedings will be found useful to all workers in this and related fields

  1. Post irradiation examination of thermal reactor fuels

    Science.gov (United States)

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

    2008-12-01

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

  2. Moving towards sustainable thorium fuel cycles

    International Nuclear Information System (INIS)

    The CANDU reactor has an unsurpassed degree of fuel-cycle flexibility as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle design. These features facilitate the introduction and full exploitation of thorium fuel cycles in CANDU reactors in an evolutionary fashion. Thoria (ThO2) based fuel offers both fuel performance and safety advantages over urania (UO2) based fuel, due its higher thermal conductivity which results in lower fuel-operating temperatures at similar linear element powers. Thoria fuel has demonstrated lower fission gas release than UO2 under similar operating powers during test irradiations. In addition, thoria has a higher melting point than urania and is far less reactive in hypothetical accident scenarios owing to the fact that it has only one oxidation state. This paper examines one possible strategy for the introduction of thorium fuel cycles into CANDU reactors. In the short term, the initial fissile material would be provided in a heterogeneous bundle of low-enriched uranium and thorium. The medium term scenario uses homogeneous Pu/Th bundles in the CANDU reactor, further increasing the energy derived from the thorium. In the long term, the full energy potential from thorium would be realized through the recycle of the U-233 in the used fuel. With U-233 recycle in CANDU reactors, plutonium would then only be required to top up the fissile content to achieve the desired burnup. (author)

  3. Thermomechanical evaluation of BWR fuel elements for procedures of preconditioned with FEMAXI-V

    International Nuclear Information System (INIS)

    The limitations in the burnt of the nuclear fuel usually are fixed by the one limit in the efforts to that undergo them the components of a nuclear fuel assembly. The limits defined its provide the direction to the fuel designer to reduce to the minimum the fuel failure during the operation, and they also prevent against some thermomechanical phenomena that could happen during the evolution of transitory events. Particularly, a limit value of LHGR is fixed to consider those physical phenomena that could lead to the interaction of the pellet-shirt (Pellet Cladding Interaction, PCI). This limit value it is related directly with an PCI limit that can be fixed based on experimental tests of power ramps. This way, to avoid to violate the PCI limit, the conditioning procedures of the fuel are still required for fuel elements with and without barrier. Those simulation procedures of the power ramp are carried out for the reactor operator during the starting maneuvers or of power increase like preventive measure of possible consequences in the thermomechanical behavior of the fuel. In this work, the thermomechanical behavior of two different types of fuel rods of the boiling water reactor is analyzed during the pursuit of the procedures of fuel preconditioning. Five diverse preconditioning calculations were carried out, each one with three diverse linear ramps of power increments. The starting point of the ramps was taken of the data of the cycle 8 of the unit 1 of the Laguna Verde Nucleo electric Central. The superior limit superior of the ramps it was the threshold of the lineal power in which a fuel failure could be presented by PCI, in function of the fuel burnt. The analysis was carried out with the FEMAXI-V code. (Author)

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

    International Nuclear Information System (INIS)

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

  5. Finite difference analysis of the transient temperature profile within GHARR-1 fuel element

    International Nuclear Information System (INIS)

    Highlights: • Transient heat conduction for GHARR-1 fuel was developed and simulated by MATLAB. • The temperature profile after shutdown showed parabolic decay pattern. • The recorded temperature of about 411.6 K was below the melting point of the clad. • The fuel is stable and no radioactivity will be released into the coolant. - Abstract: Mathematical model of the transient heat distribution within Ghana Research Reactor-1 (GHARR-1) fuel element and related shutdown heat generation rates have been developed. The shutdown heats considered were residual fission and fission product decay heat. A finite difference scheme for the discretization by implicit method was used. Solution algorithms were developed and MATLAB program implemented to determine the temperature distributions within the fuel element after shutdown due to reactivity insertion accident. The simulations showed a steady state temperature of about 341.3 K which deviated from that reported in the GHARR-1 safety analysis report by 2% error margin. The average temperature obtained under transient condition was found to be approximately 444 K which was lower than the melting point of 913 K for the aluminium cladding. Thus, the GHARR-1 fuel element was stable and there would be no release of radioactivity in the coolant during accident conditions

  6. Apparatus for the storage of transport- and storage-containers containing radioactive fuel elements

    International Nuclear Information System (INIS)

    The invention concerns an apparatus for the storage of transport and storage containers containing radioactive fuel elements. For each transport or storage container there is a separate silo-type container of steel, concrete, prestressed concrete or suchlike breakproof and fireproof material, to be placed in the open, that can be opened for removal and placing of the transport or storage container respectively. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

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

  10. End plug welding of FBTR fuel element by Nd-YAG CW laser

    International Nuclear Information System (INIS)

    Currently thin walled SS-316 FBTR fuel element is closed at both ends with SS solid plug using TIG welding at Radiometallurgy Division, Bhabha Atomic Research Centre. This welding is involved with material of non uniform thickness. The tube is thinner whereas solid plug is thicker material. Precise control of welding parameters and positioning of torch is necessary during welding to get satisfactory penetration without any major defect. In order to meet the objective of achieving good quality weld heat input during welding has to be kept minimal. Of late in Radiometallurgy Division, Bhabha Atomic Research Centre Nd-YAG CW laser welding has been introduced for end plug welding of the fuel element. Necessary experiments have been done to evaluate the performance of laser welding process for the welding of FBTR fuel elements. Laser welding of fuel element has been done at different power level and speeds. The depth of penetration, fusion zone shape, overall weld profile and weld bead thickness have been evaluated. The amount of liquid metal formed during welding has been calculated theoretically by fitting the overall weld profile in mathematical function. The weld profile area has been estimated by image analysis technique. From these data, the heat absorption by the weld pool have been calculated and thermal efficiency has been estimated. Attempt has been made to optimize laser welding parameters. (author)

  11. Prospect of Uranium Silicide fuel element with hypostoichiometric (Si ≤3.7%)

    International Nuclear Information System (INIS)

    An attempt to obtain high uranium-loading in silicide dispersion fuel element using the fabrication technology applicable nowadays can reach Uranium-loading slightly above 5 gU/cm3. It is difficult to achieve a higher uranium-loading than that because of fabricability constraints. To overcome those difficulties, the use of uranium silicide U3Si based is considered. The excess of U is obtained by synthesising U3Si2 in Si-hypostoichiometric stage, without applying heat treatment to the ingot as it can generate undesired U3Si. The U U will react with the matrix to form U alx compound, that its pressure is tolerable. This experiment is to consider possibilities of employing the U3Si2 as nuclear fuel element which have been performed by synthesising U3Si2-U with the composition of 3.7 % weigh and 3 % weigh U. The ingot was obtained and converted into powder form which then was fabricated into experimental plate nuclear fuel element. The interaction between free U and Al-matrix during heat-treatment is the rolling phase of the fuel element was observed. The study of the next phase will be conducted later

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

    International Nuclear Information System (INIS)

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

  13. R and D activities for ultimate disposal of HTR spent fuel elements

    International Nuclear Information System (INIS)

    For the direct ultimate disposal of HTR spent fuel, the LWR MAW storage method is examined, with a view to the following aspects: Activity released under rock pressure, and data of the design basis accidents: drop of waste package, underground fire, and inflow of lye in the post-operational phase of the repository. Strength tests with model packages containing HTR spent fuel in loose packing, or backfilled, have shown that backfilling with cement mortar or sand prevents damage to the spent fuel caused by rock pressure. Particle cracking in damaged, not backfilled fuel elements was at a maximum of 1 p.c. In leaching experiments with spent, complete fuel elements (900C - 2000C and 130/300 bar) only contaminations of the graphite matrix were released. Nuclide release after some 100 days of leaching time is determined by diffusion processes, but even after 1250 days of leaching, only trace amounts of Cs, Sr, Ce, Ba, Eu, Co, actinides or heavy metals could be found. High burn-up UO2 fuel kernels without coating and graphite matrix under the effects of leaching at 900C, 130 bar already after 60 days exhibited strong corrosion damage, and after about 100 days, cesium was released almost completely. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Tolosa, S.C.; Marajofsky, A.

    2004-10-06

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

  15. Velocity and turbulence distributions in wall subchannels of a road bundle in three axial planes downstream of a spacer grid

    Science.gov (United States)

    Rehme, K.

    1987-03-01

    The velocity, turbulence, and temperature distributions in nuclear fuel element bundles of nuclear reactors were investigated. The mean velocity, the wall shear stresses, and the turbulence were measured in two wall subchannels of a rod bundle of four parallel rods, arranged in a rectangular channel, for three axial planes. A spacer grid was inserted in the rod bundle, for ratios between the distance spacer grid/measuring plane and the hydraulic diameter (LIDh) of 40.4, 32.8 and 16.9. The Reynolds number was 145,000. The results show that the distributions of the velocity and the turbulence are affected by the spacer grid, already for LIDh = 40.4. The effects of the spacer grid increase with decreasing distance to the spacer grid.

  16. Fluidized-bed gasification under pressure of fuel element graphite in an industrial-scale reprocessing plant for HTR fuel elements

    International Nuclear Information System (INIS)

    In the head end of nuclear fuel reprocessing, the graphite cladding of spent HTR fuel elements is separated from the fissible material. Fluidized-bed combustion has some advantages over fixed-bed combustion. It is the method of choice in the design of a large-scale plant of 50.000 MWe HTR power. By means of an excess pressure of about 5 bar, a threefold increase in efficiency of a fluidized-bed ractor can be achieved. For an optimum layout of a prototype combustion plant, jacket cooling and internal heat exchangers are required. For an assessment of fluidized-bed combustion under pressure as a process step in the head end of a reprocessing plant, the author presents heat transfer calculations on the basis of a varying specific combustion load and investigations of the necessary peripheral equipment (reactor vessel, dust removal systems, gas supply and distribution, etc.) in several model set-ups. (RB)

  17. A review of trace element emissions from the combustion of refuse-derived fuel with coal

    International Nuclear Information System (INIS)

    The effects of cocombusting refuse-derived fuel (RDF) with coal on stack emissions of trace elements in the ash stream were reviewed. The large number of variables and uncertainties involved precluded drawing definitive conclusions regarding many of the trace elements. However, it is evident that cocombustion resulted in increased emissions of Cd, Cu, Hg, Pb, and Zn. Emissions of As and Ni tended to decrease when RDF was fired with coal. Modeling studies indicated that ambient levels of trace elements during cocombustion should be within acceptable limits. However, periodic monitoring of Cd, Hg, and Pb may be warranted in some instances

  18. High-Uranium-Loaded U3O8-Al fuel element development program. Part 1

    International Nuclear Information System (INIS)

    The High-Uranium-Loaded U3O8-Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U3O8-Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U3O8-Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U3O8). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % U involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U3O8 and aluminum

  19. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

  20. Restrictions of stable bundles

    CERN Document Server

    Balaji, V

    2011-01-01

    The Mehta-Ramanathan theorem ensures that the restriction of a stable vector bundle to a sufficiently high degree complete intersection curve is again stable. We improve the bounds for the "sufficiently high degree" and propose a possibly optimal conjecture.

  1. TRISO-Fuel Element Performance Modeling for the Hybrid LIFE Engine with Pu Fuel Blanket

    Energy Technology Data Exchange (ETDEWEB)

    DeMange, P; Marian, J; Caro, M; Caro, A

    2010-02-18

    A TRISO-coated fuel thermo-mechanical performance study is performed for the hybrid LIFE engine to test the viability of TRISO particles to achieve ultra-high burnup of a weapons-grade Pu blanket. Our methodology includes full elastic anisotropy, time and temperature varying material properties for all TRISO layers, and a procedure to remap the elastic solutions in order to achieve fast fluences up to 30 x 10{sup 25} n {center_dot} m{sup -2} (E > 0.18 MeV). In order to model fast fluences in the range of {approx} 7 {approx} 30 x 10{sup 25} n {center_dot} m{sup -2}, for which no data exist, careful scalings and extrapolations of the known TRISO material properties are carried out under a number of potential scenarios. A number of findings can be extracted from our study. First, failure of the internal pyrolytic carbon (PyC) layer occurs within the first two months of operation. Then, the particles behave as BISO-coated particles, with the internal pressure being withstood directly by the SiC layer. Later, after 1.6 years, the remaining PyC crumbles due to void swelling and the fuel particle becomes a single-SiC-layer particle. Unrestrained by the PyC layers, and at the temperatures and fluences in the LIFE engine, the SiC layer maintains reasonably-low tensile stresses until the end-of-life. Second, the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Obtaining more reliable measurements, especially at higher fluences, is an imperative for the fidelity of our models. Finally, varying the geometry of the TRISO-coated fuel particles results in little differences in the scope of fuel performance. The mechanical integrity of 2-cm graphite pebbles that act as fuel matrix has also been studied and it is concluded that they can reliable serve the entire LIFE burnup cycle without failure.

  2. Experimental verification of structural models to analyze the nonlinear dynamics of LMFBR fuel elements

    International Nuclear Information System (INIS)

    Local fault situations in LMFBR cores may produce severe pressure pulses within one fuel element. The fact cannot be ignored that these pressures can have peaks and impulses that may expand and rupture the wrapper around the element. This will impulsively load the surrounding subassemblies and possibly the control rods due to extreme coolant pressure gradients and/or subassembly collision forces. Fast reactor safety requires this mechanical propagation process through the core to be analyzed, and therefore appropriate models and solution methods are needed to simulate the nonlinear structural dynamics of one typical hexagonal fuel element. The aim of this paper is to outline one- and two-dimensional structural models and discuss their capabilities and suitability for multirow core calculations. For this purpose static and impulsive single subassembly loading experiments are described and typical results are reported and compared with numerical predictions. (Auth.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-13

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

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

    International Nuclear Information System (INIS)

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

  5. 17506 - Order of 20 july 1989 on the storage period for fuel elements for spanish nuclear power plants

    International Nuclear Information System (INIS)

    The Order was made in furtherance of Decree No. 813/1988 amending a Decree of 1985 on the reorganisation of activities in the nuclear fuel cycle. It establishes new requirements regarding fuel elements for PWRs and BWRs, namely by providing that their operators should stock enough fuel elements for one load at least two months prior to the planned loading. Other plants should have the number of fuel elements necessary for their continuous operation for four months at 80 per cent of their nominal power

  6. Legal questions concerning the termination of spent fuel element reprocessing; Rechtsfragen der Beendigung der Wiederaufarbeitung abgebrannter Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    John, Michele

    2005-07-01

    The thesis on legal aspects of the terminated spent fuel reprocessing in Germany is based on the legislation, jurisdiction and literature until January 2004. The five chapters cover the following topics: description of the problem; reprocessing of spent fuel elements in foreign countries - practical and legal aspects; operators' responsibilities according to the atomic law with respect to the reprocessing of Geman spent fuel elements in foreign countries; compatibility of the prohibition of Geman spent fuel element reprocessing in foreign countries with international law, European law and German constitutional law; results of the evaluation.

  7. Mechanical Property Evaluation of High Burnup PHWR Fuel Clads

    International Nuclear Information System (INIS)

    Assurance of clad integrity is of vital importance for the safe and reliable extension of fuel burnup. In order to study the effect of extended burnup of 15,000 MW∙d/tU on the performance of Pressurised Heavy Water Reactor (PHWR) fuel bundles of 19-element design, a couple of bundles were irradiated in Indian PHWR. The tensile property of irradiated cladding from one such bundle was evaluated using the ring tension test method. Using a similar method, claddings of mixed oxide (MOX) fuel elements irradiated in the pressurized water loop (PWL) of CIRUS to a burnup of 10,000 MW∙d/THM were tested. The tests were carried out both at ambient temperature and at 300°C. The paper will describe the test procedure, results generated and discuss the findings. (author)

  8. Development of TUF-ELOCA - a software tool for integrated single-channel thermal-hydraulic and fuel element analyses

    International Nuclear Information System (INIS)

    The TUF-ELOCA tool couples the TUF and ELOCA codes to enable an integrated thermal-hydraulic and fuel element analysis for a single channel during transient conditions. The coupled architecture is based on TUF as the parent process controlling multiple ELOCA executions that simulate the fuel elements behaviour and is scalable to different fuel channel designs. The coupling ensures a proper feedback between the coolant conditions and fuel elements response, eliminates model duplications, and constitutes an improvement from the prediction accuracy point of view. The communication interfaces are based on PVM and allow parallelization of the fuel element simulations. Developmental testing results are presented showing realistic predictions for the fuel channel behaviour during a transient. (author)

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

    International Nuclear Information System (INIS)

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

  10. Romanian irradiation experiment on AHWR type fuel elements containing mixed oxide of thorium and uranium pellets

    International Nuclear Information System (INIS)

    One of the main objectives of the Institute for Nuclear Research (ICN) - Nuclear Fuel R and D Program is the development of new types of fuel based on: Slightly Enriched Uranium (SEU), Recycled Uranium (RU) and Thorium. Two experimental fuel elements (A23 and A24) were irradiated in TRIGA Research Reactor of ICN-Romania (C1 device) in a power ramp conditions. Element A23, contained mixed oxide of thorium and uranium pellets, achieved a maximum linear power of 51 KW/m and has reached a discharge burn-up around 189.2 MWh/KgHE; element A24, contained dioxide of uranium pellets, achieved a maximum linear power of 63 kW/m and has reached a discharge burn-up of around 207.8 MWh/KgHE. The experiments simulation has been performed using an improved version of ROFEM Code, version developed through the efforts of researchers from ICN - Nuclear Fuel Performance Department. The simulation results are in good agreement with experimental data. (author)

  11. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, September 1, 1980-November 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1981-02-01

    Four tasks are reported: bundle geometry (wrapped and bare rods), subchannel geometry (bare rods), subchannel geometry (bare rods), LMFBR outlet plenum flow mixing, and theoretical determination of local temperature fields in LMFBR fuel rod bundles. (DLC)

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

    International Nuclear Information System (INIS)

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

  13. Subtleties Concerning Conformal Tractor Bundles

    CERN Document Server

    Graham, C Robin

    2012-01-01

    The realization of tractor bundles as associated bundles in conformal geometry is studied. It is shown that different natural choices of principal bundle with normal Cartan connection corresponding to a given conformal manifold can give rise to topologically distinct associated tractor bundles for the same inducing representation. Consequences for homogeneous models and conformal holonomy are described. A careful presentation is made of background material concerning standard tractor bundles and equivalence between parabolic geometries and underlying structures.

  14. The isolation and analysis of critical stress areas in graphite fuel elements

    International Nuclear Information System (INIS)

    This paper presents a summary of methods being used at General Atomic for the analysis of irradiation-induced stresses in the graphite fuel elements of a high-temperature gas-cooled reactor (HTGR). The main topic of the paper is a systematic strategy for the isolation and detailed analysis of critical stress areas. The first program in the hierarchy, named SURVEY/STRESS, uses a simplified beam theory model to sort and identify potentially critical fuel elements among the thosands in the core. SURVEY/STRESS employs an idealized geometry consisting of an array of parallel fibers, each one of which is an accurate representation of the symmetrical web between the fuel and coolant holes. The program automatically superimposes the localized web stress field on the average stress fiel of the block. A more refined sorting is done with the aid of a general purpose beam theory code named GBEAM. GBEAM employs an array of triangular elements to accurately model the geometry. The calculated axial stresses are very close to those of a two-dimensional analysis, but they are obtained at a considerable cost saving. The third program in the series is SAFIRE, a two-dimensional, viscoelastic, finite element program. SAFIRE employs a generalized plane strain (GPS) element in which three global bending degrees of freedom are introduced at the element level. The GPS element correctly considers the stiffness and Poisson coupling effects of out-of-plane deformations. This is a significant improvement over earlier programs that superimposed reversed reaction forces on a plane str

  15. Conceptual design of a multicell thermionic fuel element for a 40-KWE space nuclear power system

    International Nuclear Information System (INIS)

    This paper addresses the conceptual design of a Russian thermionic fuel element (TFE) in support of the S-PRIME 40 kWe in-core thermionic space power reactor studies sponsored by the U.S. Department of Energy TI-SNPS program. The design responds to requirements specified by the U.S. component of the S-PRIME team and is based on the multicell ''flashlight'' TFE approach. Following a general description of the TFE design, the considerations leading to several key design decisions are discussed. These include nuclear and thermionic performance, cesium management, fission product management, materials selection, fuel system, and lifetime. copyright American Institute of Physics 1995

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

    International Nuclear Information System (INIS)

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

  17. Development of the CANDU high-burnup fuel design/analysis technology

    International Nuclear Information System (INIS)

    This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs

  18. A two-dimensional finite element method for analysis of solid body contact problems in fuel rod mechanics

    International Nuclear Information System (INIS)

    Two computer codes for the analysis of fuel rod behavior have been developed. Fuel rod mechanics is treated by a two-dimensional, axisymmetric finite element method. The program KONTAKT is used for detailed examinations on fuel rod sections, whereas the second program METHOD2D allows instationary calculations of whole fuel rods. The mechanical contact of fuel and cladding during heating of the fuel rod is very important for it's integrity. Both computer codes use a Newton-Raphson iteration for the solution of the nonlinear solid body contact problem. A constitutive equation is applied for the dependency of contact pressure on normal approach of the surfaces which are assumed to be rough. If friction is present on the contacting surfaces, Coulomb's friction law is used. Code validation is done by comparison with known analytical solutions for special problems. Results of the contact algorithm for an elastic ball pressing against a rigid surface are confronted with Hertzian theory. Influences of fuel-pellet geometry as well as influences of discretisation of displacements and stresses of a single fuel pellet are studied. Contact of fuel and cladding is calculated for a fuel rod section with two fuel pellets. The influence of friction forces between fuel and cladding on their axial expansion is demonstrated. By calculation of deformations and temperatures during an instationary fuel rod experiment of the CABRI-series the feasibility of two-dimensional finite element analysis of whole fuel rods is shown. (orig.)

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

    International Nuclear Information System (INIS)

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

  20. Fusion option to dispose of spent nuclear fuel and transuranic elements

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.

    2000-02-10

    The fusion option is examined to solve the disposition problems of the spent nuclear fuel and the transuranic elements. The analysis of this report shows that the top rated solution, the elimination of the transuranic elements and the long-lived fission products, can be achieved in a fusion reactor. A 167 MW of fusion power from a D-T plasma for sixty years with an availability factor of 0.75 can transmute all the transuranic elements and the long-lived fission products of the 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. The operating time can be reduced to thirty years with use of 334 MW of fusion power, a system study is needed to define the optimum time. In addition, the fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future. Fusion blankets with a liquid carrier for the transuranic elements can achieve a transmutation rate for the transuranic elements up to 80 kg/MW.y of fusion power with k{sub eff} of 0.98. In addition, the liquid blankets have several advantages relative to the other blanket options. The energy from this transmutation is utilized to produce revenue for the system. Molten salt (Flibe) and lithium-lead eutectic are identified as the most promising liquids for this application, both materials are under development for future fusion blanket concepts. The Flibe molten salt with transuranic elements was developed and used successfully as nuclear fuel for the molten salt breeder reactor in the 1960's.

  1. Bundling in semiflexible polymers: A theoretical overview.

    Science.gov (United States)

    Benetatos, Panayotis; Jho, YongSeok

    2016-06-01

    Supramolecular assemblies of polymers are key modules to sustain the structure of cells and their function. The main elements of these assemblies are charged semiflexible polymers (polyelectrolytes) generally interacting via a long(er)-range repulsion and a short(er)-range attraction. The most common supramolecular structure formed by these polymers is the bundle. In the present paper, we critically review some recent theoretical and computational advances on the problem of bundle formation, and point a few promising directions for future work. PMID:26813628

  2. Fabrication of simulated plate fuel elements: Defining role of stress relief annealing

    Science.gov (United States)

    Kohli, D.; Rakesh, R.; Sinha, V. P.; Prasad, G. J.; Samajdar, I.

    2014-04-01

    This study involved fabrication of simulated plate fuel elements. Uranium silicide of actual fuel elements was replaced with yttria. The fabrication stages were otherwise identical. The final cold rolled and/or straightened plates, without stress relief, showed an inverse relationship between bond strength and out of plane residual shear stress (τ13). Stress relief of τ13 was conducted over a range of temperatures/times (200-500 °C and 15-240 min) and led to corresponding improvements in bond strength. Fastest τ13 relief was obtained through 300 °C annealing. Elimination of microscopic shear bands, through recovery and partial recrystallization, was clearly the most effective mechanism of relieving τ13.

  3. Finite element simulation of fission gas release and swelling in UO2 fuel pellets

    International Nuclear Information System (INIS)

    A fission gas release model is presented, which solves the atomic diffusion problem with xenon and krypton elements tramps produced by uranium fission during UO2 nuclear fuel irradiation. The model considers intra and intergranular precipitation bubbles, its re dissolution owing to highly energetic fission products impact, interconnection of intergranular bubbles and gas sweeping by grain border in movement because of grain growth. In the model, the existence of a thermal gradient in the fuel pellet is considered, as well as temporal variations of fission rate owing to changes in the operation lineal power. The diffusion equation is solved by the finite element method and results of gas release and swelling calculation owing to gas fission are compared with experimental data. (author)

  4. Conceptual Engineering of CARA Fuel Element with Negative Void Coefficient for Atucha II

    Directory of Open Access Journals (Sweden)

    H. Lestani

    2011-01-01

    Full Text Available Experimentally validated void reactivity calculations were used to study the feasibility of a change in the design basis of Atucha II Nuclear Power Plant including the Large LOCA event. The use of CARA fuel element with burnable neutronic absorbers and enriched uranium is proposed instead of the original fuel. The void reactivity, refuelling costs, and power peaking factors are analysed at conceptual level to optimize the burnable neutronic absorber, the enrichment grade, and their distribution inside the fuel. This work concludes that, for the considered plant conditions, either a void reactivity coefficient granting no prompt critical excursion on Large LOCA or negative void reactivity is achievable, with advantages on refuelling cost and linear power density.

  5. Fabrication of simulated plate fuel elements: Defining role of out-of-plane residual shear stress

    Energy Technology Data Exchange (ETDEWEB)

    Rakesh, R., E-mail: rakesh.rad87@gmail.com [DAE Graduate Fellows, IIT Bombay, Powai, Mumbai 400076 (India); Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Kohli, D. [DAE Graduate Fellows, IIT Bombay, Powai, Mumbai 400076 (India); Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Sinha, V.P.; Prasad, G.J. [Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Samajdar, I. [Department of Metallurgical Engineering and Materials Science, IIT Bombay, Powai, Mumbai 400076 (India)

    2014-02-01

    Bond strength and microstructural developments were investigated during fabrication of simulated plate fuel elements. The study involved roll bonding of aluminum–aluminum (case A) and aluminum–aluminum + yttria (Y{sub 2}O{sub 3}) dispersion (case B). Case B approximated aluminum–uranium silicide (U{sub 3}Si{sub 2}) ‘fuel-meat’ in an actual plate fuel. Samples after different stages of fabrication, hot and cold rolling, were investigated through peel and pull tests, micro-hardness, residual stresses, electron and micro-focus X-ray diffraction. Measurements revealed a clear drop in bond strength during cold rolling: an observation unique to case B. This was related to significant increase in ‘out-of-plane’ residual shear stresses near the clad/dispersion interface, and not from visible signatures of microstructural heterogeneities.

  6. Fabrication of simulated plate fuel elements: Defining role of out-of-plane residual shear stress

    Science.gov (United States)

    Rakesh, R.; Kohli, D.; Sinha, V. P.; Prasad, G. J.; Samajdar, I.

    2014-02-01

    Bond strength and microstructural developments were investigated during fabrication of simulated plate fuel elements. The study involved roll bonding of aluminum-aluminum (case A) and aluminum-aluminum + yttria (Y2O3) dispersion (case B). Case B approximated aluminum-uranium silicide (U3Si2) 'fuel-meat' in an actual plate fuel. Samples after different stages of fabrication, hot and cold rolling, were investigated through peel and pull tests, micro-hardness, residual stresses, electron and micro-focus X-ray diffraction. Measurements revealed a clear drop in bond strength during cold rolling: an observation unique to case B. This was related to significant increase in 'out-of-plane' residual shear stresses near the clad/dispersion interface, and not from visible signatures of microstructural heterogeneities.

  7. A study on the laser welding of aluminum for end-capping of KMRR fuel elements

    International Nuclear Information System (INIS)

    The selecting of proper welding process for aluminum end-capping of KMRR nuclear fuel elements is considered important in respect to the soundness of weldments and the improvement of the performance of nuclear fuels during the operation in reactor. The probability of leakage of the fission products is mostly apt to occur at the weldments, and it is connected directly with the safety and life prediction of the nuclear reactor in operation. This study is to investigate the influence of Ti and Zr micro-additions on the grain refinement in Al weld metal and to prevent weld cracks in pure Al. In addition, mechanical properties of laser welded specimen were observed and the technical data for KMRR nuclear fuels manufacturing was provided. In the future, this study will also present the improvement of KMRR nuclear end-caps welding. (Author)

  8. Analysis of gamma ray spectrum of Al-Pu plate fuel elements

    International Nuclear Information System (INIS)

    Gamma ray spectrometry of aluminium (0.5mm thick) clad Al-Pu plate fuel elements containing ∼10 g plutonium per plate, was carried out using a high purity germanium detector (HPGe) and a 4K multichannel analyser. The detector to fuel plate distance was kept 750 mm to have a full view of the plate and to reduce dead time corrections. The data was analysed using peak fitting program SAMPO. Typical spectral shape obtained in this study is presented. Such spectral calibration avoids errors based on extrapolation from point source geometries and could be useful in evaluating (i) plutonium contents in similar plate fuels and in turn checking the final plutonium inventory and (ii)radiation fields at different working distances. (author). 3 refs., 1 fig., 1 tab

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  12. Right bundle branch block

    DEFF Research Database (Denmark)

    Bussink, Barbara E; Holst, Anders Gaarsdal; Jespersen, Lasse;

    2013-01-01

    AimsTo determine the prevalence, predictors of newly acquired, and the prognostic value of right bundle branch block (RBBB) and incomplete RBBB (IRBBB) on a resting 12-lead electrocardiogram in men and women from the general population.Methods and resultsWe followed 18 441 participants included.......5%/2.3% in women, P Right bundle branch block was associated with significantly...... increased all-cause and cardiovascular mortality in both genders with age-adjusted hazard ratios (HR) of 1.31 [95% confidence interval (CI), 1.11-1.54] and 1.87 (95% CI, 1.48-2.36) in the gender pooled analysis with little attenuation after multiple adjustment. Right bundle branch block was associated...

  13. Cna 1 spent fuel element interim dry storage system thermal analysis

    International Nuclear Information System (INIS)

    At the moment, the Atucha I Nuclear Power Plant (Cnea-I) located in the city of Lima, has enough room to store its spent fuel (Sf) in their two pools spent fuel until about 2015.In case of life extension a spend fuel element interim dry storage system is needed.Nucleolectrica Argentina S.A. (N A-S A) and the Comision Nacional de Energia Atomica (Cnea), have proposed different interim dry storage systems.These systems have to be evaluated in order to choose one of them.The present work's objective is the thermal analysis of one dry storage alternative for the Sf element of Cna 1.In this work a simple model was developed and used to perform the thermal calculations corresponding to the system proposed by Cnea.This system considers the store of sealed containers with 37 spent fuels in concrete modules.Each one of the containers is filled in the pool houses and transported to the module in a transference cask with lead walls.Fulfill the maximum cladding temperature requirement (<200 degrade C) for a total 2140 W thermal power

  14. Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

    2010-04-01

    The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

  15. An advancement in iterative solution schemes for three-dimensional, two-fluid modeling of two-phase flow in PWR fuel bundles

    International Nuclear Information System (INIS)

    Highlights: • A fully three-dimensional two-fluid model coupled with heat conduction was outlined. • Two-fluid numerical scheme capability was evaluated against NUPEC PSBT Benchmark. • GMRES, FGMRES, DQGMRES, CGNR, BCG, and TFQMR solvers were tested as iterative schemes. • Candidate Krylov solvers do not introduce deviations to the two-phase flow results. • GMRES, FGMRES, and DQGMRES have a more efficient and stable convergence performance. - Abstract: This paper outlines a fully three-dimensional two-fluid one-pressure model with a semi-implicit finite difference scheme coupled with heat conduction which can be applicable to thermal non-equilibrium two-phase flow field in subchannel geometry of Pressurized Water Reactors (PWR). The system of equations was linearized using the Newton–Raphson method and was collapsed into the pressure equations forming a system of the Poisson type. Then, two-phase flow modeling was combined with Krylov methods as advanced computing techniques to investigate the feasibility of implementing preconditioned Krylov subspace solvers as the numerical scheme to solve pressure equations. Six popular Krylov subspace solvers were considered: GMRES, FGMRES, DQGMRES, CGNR, BCG, and TFQMR combined with the block incomplete LU factorization with a dual truncation strategy (BILUT) preconditioner. These proposed iterative solvers were applied to the constructed linear pressure equations in the inner iteration in combination with the outer-Raphson iteration loop. Evaluation was performed in two stages. First, two-fluid numerical scheme capability was evaluated against OECD/NRC NUPEC PWR Bundle tests (PSBT Benchmark). The results for steady-state (PSBT) bundle show that an overall agreement can be found. At the second stage, convergency, stability, and accuracy of the proposed schemes were studied based on PSBT steady-state data through a comparison of utilized Krylov solvers and the direct inversion method as the pressure solution

  16. Thermal energy management and chemical reaction investigation of micro-proton exchange membrane fuel cell and fuel cell system using finite element modelling

    OpenAIRE

    McGee, Seán

    2015-01-01

    Fuel cell systems are becoming more commonplace as a power generation method and are being researched, developed, and explored for commercial use, including portable fuel cells that appear in laptops, phones, and of course, chargers. This thesis examines a model constructed on inspiration from the myFC PowerTrekk, a portable fuel cell charger, using COMSOL Multiphysics, a finite element analysis software. As an educational tool and in the form of zero-dimensional, two-dimensional, and three-d...

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

    International Nuclear Information System (INIS)

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

  18. Principal -bundles on Nodal Curves

    Indian Academy of Sciences (India)

    Usha N Bhosle

    2001-08-01

    Let be a connected semisimple affine algebraic group defined over . We study the relation between stable, semistable -bundles on a nodal curve and representations of the fundamental group of . This study is done by extending the notion of (generalized) parabolic vector bundles to principal -bundles on the desingularization of and using the correspondence between them and principal -bundles on . We give an isomorphism of the stack of generalized parabolic bundles on with a quotient stack associated to loop groups. We show that if is simple and simply connected then the Picard group of the stack of principal -bundles on is isomorphic to ⊕ , being the number of components of .

  19. Bundles of Banach algebras

    Directory of Open Access Journals (Sweden)

    J. W. Kitchen

    1994-01-01

    Full Text Available We study bundles of Banach algebras π:A→X, where each fiber Ax=π−1({x} is a Banach algebra and X is a compact Hausdorff space. In the case where all fibers are commutative, we investigate how the Gelfand representation of the section space algebra Γ(π relates to the Gelfand representation of the fibers. In the general case, we investigate how adjoining an identity to the bundle π:A→X relates to the standard adjunction of identities to the fibers.

  20. The FPIN2 code - an application of the finite element method to the analysis of the transient response of oxide and metal fuel elements

    International Nuclear Information System (INIS)

    The FPIN2 code simulates the thermal-mechanical response of fast reactor fuel pins to transient events. Temperatures of the fuel pin and coolant are calculated using a simple pin-in-a-pipe geometry. The mechanical analysis uses an implicit finite element formulation with linear shape functions which allow for general material behavior in the fuel and cladding including cracking and melting. This formulation provides a very convenient structure for implementing different models and improvements in algorithms. The paper summarizes the FPIN2 methodology and presents results for the transient response of both oxide and metallic fuel pins under similar overpower transients. (author)

  1. On projective space bundle with nef normalized tautological line bundle

    CERN Document Server

    Yasutake, Kazunori

    2011-01-01

    In this paper, we study the structure of projective space bundles whose relative anti-canonical line bundle is nef. As an application, we get a characterization of abelian varieties up to finite etale covering.

  2. Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

    2009-08-01

    Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

  3. On the possibility of reprocessing of fuel elements of dispersion type with copper matrix by pyrochemical methods

    International Nuclear Information System (INIS)

    A consideration is given to pyrochemical processes suitable for separation of uranium dioxide from structural materials when reprocessing cermet type fuel elements. The estimation of the possibility to apply liquid antimony and bismuth, potassium and copper chlorides melts is made. The specimens compacted of copper and uranium dioxide powders in a stainless steel can are used as simulators of fuel element sections. It is concluded that the dissolution of structural materials in molten salts at the stage of uranium dioxide concentration is the process of choice for reprocessing of dispersion type fuel elements

  4. Thorium-Based Fuels Preliminary Lattice Cell Studies for Candu Reactors

    International Nuclear Information System (INIS)

    The choice of nuclear power as a major contributor to the future global energy needs must take into account acceptable risks of nuclear weapon proliferation, in addition to economic competitiveness, acceptable safety standards, and acceptable waste disposal options. Candu reactors offer a proven technology, safe and reliable reactor technology, with an interesting evolutionary potential for proliferation resistance, their versatility for various fuel cycles creating premises for a better utilization of global fuel resources. Candu reactors impressive degree of fuel cycle flexibility is a consequence of its channel design, excellent neutron economy, on-power refueling, and simple fuel bundle. These features facilitate the introduction and exploitation of various fuel cycles in Candu reactors in an evolutionary fashion. The main reasons for our interest in Thorium-based fuel cycles have been, globally, to extend the energy obtainable from natural Uranium and, locally, to provide a greater degree of energy self-reliance. Applying the once through Thorium (OTT) cycle in existing and advanced Candu reactors might be seen as an evaluative concept for the sustainable development both from the economic and waste management points of view. Two Candu fuel bundles project will be used for the proposed analysis, namely the Candu standard fuel bundle with 37 fuel elements and the CANFLEX fuel bundle with 43 fuel elements. Using the Canadian proposed scheme - loading mixed ThO2-SEU CANFLEX bundles in Candu 6 reactors - simulated at lattice cell level led to promising conclusions on operation at higher fuel burnups, reduction of the fissile content to the end of the cycle, minor actinide content reduction in the spent fuel, reduction of the spent fuel radiotoxicity, presence of radionuclides emitting strong gamma radiation for proliferation resistance benefit. The calculations were performed using the lattice codes WIMS and Dragon (together with the corresponding nuclear data

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

    International Nuclear Information System (INIS)

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

  6. Stainless steel sheathed UO2 fuel elements irradiated in the X-7 organic loop

    International Nuclear Information System (INIS)

    The X-7 organic loop of the NRX reactor was successfully commissioned by irradiating nine stainless steel sheathed UO2 fuel elements to a maximum burnup of about 1450 MWd/tonne U. The loop conditions and coolant chemistry for the tests are reported, The sheath thermocouples all failed early in the Commissioning test irradiation, while the bulk coolant thermocouples with enclosed junctions operated satisfactorily. All elements were covered with an organic fouling film whose thickness and appearance seemed to vary with sheath temperature. Dimensions of the elements had not changed from pre-irradiation values, The behavior of the UO2 was consistent with other Chalk River tests at similar heat ratings. (author)

  7. Hybrid bundle divertor design

    International Nuclear Information System (INIS)

    A hybrid bundle divertor design is presented that produces <0.3% magnetic ripple at the center of the plasma while providing adequate space for the coil shielding and structure for a tokamak fusion test reactor similar to the International Tokamak Reactor and the Engineering Test Facility (with R = 5 m, B = 5 T, and a /SUB wall/ = 1.5 m, in particular). This hybrid divertor consists of a set of quadrupole ''wing'' coils running tangent to the tokamak plasma on either side of a bundle divertor. The wing coils by themselves pull the edge of the plasma out 1.5 m and spread the thickness of the scrape-off layer from 0.1 to 0.7 m at the midplane. The clear aperture of the bundle divertor throat is 1.0 m high and 1.8 m wide. For maintenance or replacement, the hybrid divertor can be disassembled into three parts, with the bundle divertor part pulling straight out between toroidal field coils and the wing coils then sliding out through the same opening

  8. ALUMINUM BOX BUNDLING PRESS

    Directory of Open Access Journals (Sweden)

    Iosif DUMITRESCU

    2015-05-01

    Full Text Available In municipal solid waste, aluminum is the main nonferrous metal, approximately 80- 85% of the total nonferrous metals. The income per ton gained from aluminum recuperation is 20 times higher than from glass, steel boxes or paper recuperation. The object of this paper is the design of a 300 kN press for aluminum box bundling.

  9. Universal Lagrangian bundles

    NARCIS (Netherlands)

    Sepe, D.

    2013-01-01

    The obstruction to construct a Lagrangian bundle over a fixed integral affine manifold was constructed by Dazord and Delzant (J Differ Geom 26:223–251, 1987) and shown to be given by ‘twisted’ cup products in Sepe (Differ GeomAppl 29(6): 787–800, 2011). This paper uses the topology of universal Lagr

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

    International Nuclear Information System (INIS)

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

  11. Current status of research and development of nuclear fuel elements for PWR in Indonesia

    International Nuclear Information System (INIS)

    Full text: The energy need of Indonesia is increasing due to the population growth and for the economic progress. The government of Indonesia intends to apply an optimum energy mix comprising all viable prospective energy sources. The Government Regulation No. 5 year 2006 indicates the target of energy mix until 2025 and the share of nuclear energy is about 2% of primary energy or 4% of electricity (4000 MWe). The first two units of NPP is expected to be operated before 2020 as stated in Act No. 17 year 2007 on National Long Term Development Planning 2005-2025. The first NPP to be operated in Indonesia is PWR type with capacity of 1000 MWe/unit. One of the strategies to strength and increase national capacity in the program for NPP introduction is domestication industry for nuclear fuel. To reach this purpose, the activities of research and development is focus on nuclear fuel production technology for PWR. Currently research and development activity in Indonesia is to produce prototype of nuclear fuel element for PWR in the form of test fuel pin or mini pin. In this paper we will presenting the pelletization and fabrication technology development. The existing facility was designed for PHWR fuel element of CIRENE type. Development of pelletization technology is carried out by modifying the compacting machine. Parameters of compacting and sintering are determined based on both compressibility and compactibility of the pellet as indicated by density and mechanical strength of the UO2 green pellets. The sintering parameters to be determined are temperature, heating rate, and soaking time. Currently, the fabrication process is under experiment. All of the data resulted from the experiment that will be presented in this meeting. (author)

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

    International Nuclear Information System (INIS)

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

  13. Economical benefits for the use of slightly enriched fuel elements at the Atucha-I nuclear power plant

    International Nuclear Information System (INIS)

    The fuel represents a very important factor in the operative cost of the Atucha I nuclear power plant. This cost is drastically reduced with the use of fuel elements of slightly enriched uranium. The annual saving is analyzed with actual values for fuel elements with an enrichment of 0.85% by weight of U-235. With the reactor core in equilibrium state the annual saving achieved is approximately 7.5-10 u$s. According to the present irradiation plan, the benefit for the transition period is studied. An analysis of the sensitivity to differential increments in factors determining the cost of fuel elements or to changes in manufacturing losses is also performed, calculating its effect on the waste, the storage of irradiated elements and the amount of UO2 required. (Author)

  14. CANDU bundle junction. Misalignment probability and pressure-drop correlation

    International Nuclear Information System (INIS)

    The pressure drop over the bundle junction is an important component of the pressure drop in a CANDU (Canada Deuterium Uranium) fuel channel. This component can represent from ∼ 15% for aligned bundles to ∼ 26% for rotationally misaligned bundles, and is dependent on the degree of misalignment. The geometry of the junction increases the mixing between subchannels, and hence improves the thermal performance of the bundle immediately downstream. It is therefore important to model the junction's performance adequately. This paper summarizes a study sponsored by COG (CANDU Owners Group) and an NSERC (National Science and Engineering Research Council) Industrial Research Grant, undertaken, at CRL (Chalk River Laboratories) to identify and develop a bundle-junction model for potential implementation in the ASSERT (Advanced Solution of Subchannel Equations in Reactor Thermalhydraulics) subchannel code. The work reported in this paper consists of two components of this project: an examination of the statistics of bundle misalignment, demonstrating that there are no preferred positions for the bundles and therefore all misalignment angles are equally possible; and, an empirical model for the single-phase pressure drop across the junction as a function of the misalignment angle. The second section of this paper includes a brief literature review covering the experimental, analytical and numerical studies concerning the single-phase pressure drop across bundle junctions. 32 refs., 9 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-03-15

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

  16. On framed quantum principal bundles

    CERN Document Server

    Durdevic, M

    1995-01-01

    A noncommutative-geometric formalism of framed principal bundles is sketched, in a special case of quantum bundles (over quantum spaces) possessing classical structure groups. Quantum counterparts of torsion operators and Levi-Civita type connections are analyzed. A construction of a natural differential calculus on framed bundles is described. Illustrative examples are presented.

  17. RU fuel development program for an advanced fuel cycle in Korea

    International Nuclear Information System (INIS)

    Korea is a unique country, having both PWR and CANDU reactors. Korea can therefore exploit the natural synergism between the two reactor types to minimize overall waste production, and maximize energy derived from the fuel, by ultimately burning the spent fuel from its PWR reactors in CANDU reactors. As one of the possible fuel cycles, Recovered Uranium (RU) fuel offers a very attractive alternative to the use of Natural Uranium (NU) and slightly enriched uranium (SEU) in CANDU reactors. Potential benefits can be derived from a number of stages in the fuel cycle: no enrichment required, therefore no enrichment tails, direct conversion to UO2, lower sensitivity to 234U and 236U absorption in the CANDU reactor, and expected lower cost relative to NU and SEU. These benefits all fit well with the PWR-CANDU fuel cycle synergy. RU arising from the conventional reprocessing of European and Japanese oxide spent fuel by 2000 is projected to be approaching 25,000 te. The use of RU fuel in a CANDU 6 reactor should result in no serious radiological difficulties and no requirements for special precautions and should not require any new technologies for the fuel fabrication and handling. The use of the CANDU Flexible Fueling (CANFLEX) bundle as the carrier for RU will be fully compatible with the reactor design, current safety and operational requirements, and there will be improved fuel performance compared with the CANDU 37-element NU fuel bundle. Compared with the 37-element NU bundle, the RU fuel has significantly improved fuel cycle economics derived from increased burnups, a large reduction in both fuel requirements and spent fuel, arisings, and the potential lower cost for RU material. There is the potential for annual fuel cost savings in the range of one-third to two-thirds, with enhanced operating margins using RU in the CANFLEX bundle design. These benefits provide the rationale for justifying R and D efforts on the use of RU fuel for advanced fuel cycles in CANDU

  18. Cold-neutron tomography of annular flow and functional spacer performance in a model of a boiling water reactor fuel rod bundle

    International Nuclear Information System (INIS)

    Highlights: → Annular flows w/wo functional spacers are investigated by cold neutron imaging. → Liquid film thickness distribution on fuel pins and on spacer vanes is measured. → The influence of the spacers on the liquid film distributions has been quantified. → The cross-sectional averaged liquid hold-up significantly affected by the spacers. → The sapers affect the fraction of the entrained liquid hold up in the gas core. - Abstract: Dryout of the coolant liquid film at the upper part of the fuel pins of a boiling water reactor (BWR) core constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is both a safety concern and a limiting factor in the thermal power and thus for the economy of BWRs. We have investigated adiabatic, air-water annular flows in a scaled-up model of two neighboring subchannels as found in BWR fuel assemblies using cold-neutron tomography. The imaging of the double suchannel has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institute, Switzerland. Cold-neutron tomography is shown here to be an excellent tool for investigating air-water annular flows and the influence of functional spacers of different geometries on such flows. The high-resolution, high-contrast measurements provide the spatial distributions of the coolant liquid film thickness on the fuel pin surfaces as well as on the surfaces of the spacer vanes. The axial variations of the cross-section averaged liquid hold-up and its fraction in the gas core shows the effect of the spacers on the redistribution of the two phases.

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

    Science.gov (United States)

    Sarta, Josè A.; Castiblanco, Luis A.

    2005-05-01

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

  20. Regulatory inspection practices on fuel elements and core lay-out at NPPs

    International Nuclear Information System (INIS)

    The basic description of the reactor core of a nuclear power plant (NPP) is an important part of the Safety Analysis Report in all countries. Due to increased interest by regulatory authorities in the Member countries, in 1996 WGIP proposed looking at inspection aspects on fuel elements and core lay-out at nuclear power plants. The CNRA subsequently approved proceeding with this report. The report deals primarily with inspection practices and inspection requirements during nuclear power plant (NPP) operation with special emphasis on refuelling procedures. All license related topics, such as fuel and core design (mechanical, neutronic, thermal-hydraulic), as well as inspection philosophy and practices on fuel fabrication are included as appropriate serving as background information and may not be completely described. WGIP members describe their country's inspection programme according to the structure of a questionnaire (appendix 1). The individual contributions are contained in the appendix 2 and are compiled within the main chapters (1 through 3). Report Structure: 1. Licensing and Quality Assurance (QA) requirements for nuclear fuel; 2. Regulatory inspection programme during NPP operation and refuelling outages; 3. Procedures for inspection practices and inspection programme. Appendix: Questionnaire and Country specific contributions. Contributions are presented by Belgium, Finland, France, Germany, Hungary, Italy, Japan, Mexico, The Netherlands, Spain, Sweden, Switzerland, United Kingdom, USA