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Sample records for annular fuel elements

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

    International Nuclear Information System (INIS)

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

  2. 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

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

    International Nuclear Information System (INIS)

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

  4. 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

  5. Assessment of Inner Channel Blockage on the Annular Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. H.; In, W. K.; Oh, D. S.; Chun, T. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    A dual-cooled annular fuel for a pressurized water reactor (PWR) has been introduced for a significant amount of reactor power uprate. The Korea Atomic Energy Research Institute (KAERI) has been performing a research to develop a dual-cooled annular fuel for the power uprate of 20% in an optimized PWR in Korea, OPR1000. An inner channel blockage is principal one of technical issues of the annular fuel rod. The inner channel in an annular fuel is isolated from the neighbor channels unlike the outer channels. The inner channel will be faced with a DNB accident by the partial blockage. In this paper, the largest fractional channel blockage was assessed by subchannel analysis code MATRA-AF and an end plug design to complement inlet blockage of inner channel was estimated by CFD code, CFD-ACE

  6. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  7. Fluidic Analysis in an Annular Centrifugal Contactor for Fuel Reprocessing

    International Nuclear Information System (INIS)

    An annular centrifugal contactor (ACC) is a promising device for fuel reprocessing process, because it offers several advantages—a smaller size, a smaller holdup volume, and a higher separation performance—over conventional contactors such as a mixer-settler and a pulse column. Fluid dynamics and dispersion in an ACC, which has a combined mixer/centrifuge structure, are closely related to its separation performance and capacity, and this information is useful in improving equipment design. In this paper, experimental and computational fluid dynamics (CFD) studies were conducted to analyze fluidic and dispersion behavior in ACCs. Multiphase mixing (water/TBP-dodecane/air) in the annular zone was observed by Particle Imaging Velocimetry, and the change in the fluidic and dispersion behavior was ascertained under several operational conditions. The results of the CFD studies, which considered multiphase turbulent flow in the annular and rotor interior zones, were in a good agreement with the experimental data. (author)

  8. Development of probabilistic design method for annular fuel. Development of BORNFREE-CEPTAR code

    International Nuclear Information System (INIS)

    The increase of linear power and burn-up during the reactor operation is considered as one of measures for the utility of fast reactor in future, and then the application of annular fuels is under consideration. In order to make a design for thus annular fuels, annular fuel design code 'CEPTAR' has been developed in Japan Atomic Energy Agency (JAEA). In addition, probabilistic fuel design code 'BORNFREE' has been also developed for the reasonable fuel design with safety and the quantitative evaluation of design margin. In this study, aiming at the development of probabilistic design method, we developed BORNFREE-CEPTAR code to develop the reasonable design method for annular fuels. As the results of probability evaluation of fuel melting at the transient at the initial power increase, by using the probabilistic annular fuel design code 'BORNFREE-CEPTAR', the melting probability for annular fuels was estimated to be approximately two figures lower than that for solid fuels, and the remarkable decrease of melting probability, which was caused by the fuel restructuring effect, was seen in the estimation results for solid fuels, on the other hand, the results for annular fuels indicated that this effect was comparably small. In addition, the permissive linear power for annular fuels tended to enhance from that for solid fuels with the increase of initial central-hole diameter under the similar fuel melting probability condition. This indicated the possibility of higher linear power operation for high-density annular fuels than low-density solid fuels. (author)

  9. Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Han, Kyu Hyun [Korea Institute of Nuclear Safety, 19, Guseong-dong, Yuseong-gu, Daejeon, 305-338 (Korea, Republic of)

    2008-07-01

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

  10. Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

    International Nuclear Information System (INIS)

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

  11. An Evaluation of the Annular Fuel and Bottle-Shaped Fuel Concepts for Sodium Fast Reactors

    OpenAIRE

    Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

    2010-01-01

    Two innovative fuel concepts, the internally and externally cooled annular fuel and the bottle-shaped fuel, were investigated with the goal of increasing the power density and reduce the pressure drop in the sodium-cooled fast reactor, respectively. The concepts were explored for both high- and low-conversion core configurations, and metal and oxide fuels. The annular fuel concept is best suited for low-conversion metal-fuelled cores, where it can enable a power uprate of ~20%; the magnitude ...

  12. 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.

  13. Annular fuel pin heat transfer and lumped model correction

    International Nuclear Information System (INIS)

    Fuel pin heat transfer studies are important in nuclear reactor accident analysis. Based on the requirement of accuracy and the speed of the computation, a simple lumped heat transfer method or detailed numerical methods are chosen to solve the heat transfer equations. In a nuclear reactor design calculations, accuracy of the solution is very important than the speed. In a nuclear reactor simulator, the speed is important. Lumped model assumes fuel pellet is solid without central hole and the heat transfer coefficient is constant across the fuel pin. In the present study a new modified lumped heat transfer model is developed to consider the annular fuel pin's central hole, and the heat transfer coefficient is made as a function of average fuel pin temperature. Transient analyses are carried out with the above said modifications for a typical LMFBR annular fuel pin. The results of lumped heat transfer model are almost matching with the accurate numerical schemes like Crank-Nicolson method. Comparisons of results with Crank-Nicolson methods are good for small step reactivity addition, ramp reactivity insertion and large step reactivity addition, ramp reactivity insertion with and without reactivity feedbacks. Comparisons of results are good for LOFA also, with and without reactivity feedbacks. With the consideration of reactivity feedbacks, fuel temperature calculated through the present modified lumped model is matching well with Crank-Nicolson methods, and the nominal power also matching well. The modified lumped heat transfer model can be used in nuclear reactor simulation studies and in conservative accident analyses where fastness of the solution is a matter of concern. (author)

  14. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

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

  15. Development of Dual Cooled Annular Fuel Temperature Analysis Program

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong Sik; Shin, C. H.; Bang, J. G.; Kim, D. H.; Kim, S. K.; Lim, I. S.; Koo Yang Hyun [KAERI, Daejeon (Korea, Republic of)

    2010-09-15

    To calculate the temperature distribution of dual cooled annular fuel, the DUOS program has been developed. Various thermal hydraulic models to determine the inner channel and outer channel flow distribution were established based on equal pressure drop condition at the top of fuel rod. The effect of gap width change was considered by employing thermal deformation model of pellet and claddings. Heat conduction model in the pellet was solved by finite difference method to consider burnup and power difference according to pellet radius. Pellet temperature model was validated by comparison with calculated temperature profile, which was determined by analytical solution of heat conduction equation under controlled input condition. Accuracy of thermal hydraulic models of DUOS were validated by core sub-channel analysis code MATRA-AF. Coolant bulk temperature of inner/outer channel and pressure drop prediction results of DUOS program show good agreement with that of MATRA-AF. Further models should be added in DUOS program to describe dual cooled annular fuel in-pile behavior, but basic thermal analysis structure has been established successfully

  16. Development of Dual Cooled Annular Fuel Temperature Analysis Program

    International Nuclear Information System (INIS)

    To calculate the temperature distribution of dual cooled annular fuel, the DUOS program has been developed. Various thermal hydraulic models to determine the inner channel and outer channel flow distribution were established based on equal pressure drop condition at the top of fuel rod. The effect of gap width change was considered by employing thermal deformation model of pellet and claddings. Heat conduction model in the pellet was solved by finite difference method to consider burnup and power difference according to pellet radius. Pellet temperature model was validated by comparison with calculated temperature profile, which was determined by analytical solution of heat conduction equation under controlled input condition. Accuracy of thermal hydraulic models of DUOS were validated by core sub-channel analysis code MATRA-AF. Coolant bulk temperature of inner/outer channel and pressure drop prediction results of DUOS program show good agreement with that of MATRA-AF. Further models should be added in DUOS program to describe dual cooled annular fuel in-pile behavior, but basic thermal analysis structure has been established successfully

  17. Hydraulic lift-off issues for application of high performance annular fuels in pressurized water reactors

    International Nuclear Information System (INIS)

    Highlights: • Pin and assembly lift-off forces are compared between solid and annular fuel. • Annular fuel experiences much stronger uplift forces. • Much stronger hold-down forces are required by annular fuel assembly. • Engineering modifications for hold-down mechanisms are required by annular fuel. - Abstract: In the PWR core, the fuel assembly is firmly seated on the lower core plate during operation. However, if the hydraulic force exerted on the fuel assembly by coolant flow is too large and the fuel assembly is lifted-off from the lower core plate, the excessive vibration will cause fuel failure. Therefore, the hydraulic lift-off issue needs to be addressed when the advanced fuel assembly is developed. It has been shown that the advanced annular fuel design with internal cooling allows power uprating up to 50% while the peak temperature of the fuel can be reduced and the MDNBR can be maintained. However, if the coolant condition in the core is kept unchanged, increasing the core power by 50% requires the core flow rate also increase proportionally, which will give rise to the hydraulic lift-off, an important issue to be addressed. In this paper, taking the 17 × 17 solid fuel design as the reference, the hydraulic lift-off issue is investigated for proposed 12 × 12 and 13 × 13 annular fuel designs. Both the steady-state and start-up operating conditions are evaluated. It is found that the hydraulic lift-off indeed is an issue for annular fuel design which requires careful analysis. By comparison, the lift-off forces and hold-down forces required for the externally and internally cooled annular fuels (13 × 13 and 12 × 12 arrays) are several times larger than that of the referenced solid fuel (17 × 17 array). Therefore, the hold-down mechanism for annular fuel needs to be carefully designed

  18. Measurement of Quasi-periodic Oscillating Flow Motion in Simulated Dual-cooled Annular Fuel Bundle

    International Nuclear Information System (INIS)

    In order to increase a significant amount of reactor power in OPR1000, KAERI (Korea Atomic Energy Research Institute) has been developing a dual-cooled annular fuel. The dual-cooled annular fuel is simultaneously cooled by the water flow through the inner and the outer channels. KAERI proposed the 12x12 dual-cooled annular fuel array which was designed to be structurally compatible with the 16x16 cylindrical solid fuel array by maintaining the same array size and the guide tubes in the same locations, as shown in Fig. 1. In such a case, due to larger outer diameter of dual-cooled annular fuel than conventional solid fuel, a P/D (Pitch-to-Diameter ratio) of dual cooled annular fuel assembly becomes smaller than that of cylindrical solid fuel. A change in P/D of fuel bundle can cause a difference in the flow mixing phenomena between the dual-cooled annular and conventional cylindrical solid fuel assemblies. In this study, the rod bundle flow motion appearing in a small P/D case is investigated preliminarily using PIV (Particle Image Velocimetry) for dual-cooled annular fuel application

  19. Thermal hydraulic performance assessment of dual-cooled annular nuclear fuel for OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang-Hwan, E-mail: shinch@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Chun, Tae-Hyun, E-mail: thchun@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Oh, Dong-Seok, E-mail: dsoh1@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); In, Wang-Kee, E-mail: wkin@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer A thermal hydraulic performance of a 12 Multiplication-Sign 12 annular fuel array is evaluated. Black-Right-Pointing-Pointer The subchannel analysis code for the dual-cooled annular fuel, MATRA-AF is validated. Black-Right-Pointing-Pointer We evaluate the sensitivity for geometry tolerances and operating parameter. Black-Right-Pointing-Pointer We decide the essential design parameters to uprate the power generation by dual-cooled annular fuel. Black-Right-Pointing-Pointer A thermal margin amount accommodating a 20% power-uprate seems viable. - Abstract: An internally and externally cooled annular fuel was proposed for an advance PWR, which can endure substantial power uprating. KAERI is pursuing the development for a reloading of power uprated annular fuel for the operating PWR reactors of OPR-1000. In this paper, the characteristics and verification of the MATRA-AF are described. The thermal hydraulic performance of a 12 Multiplication-Sign 12 annular fuel is calculated for the major design parameters and its performance is compared against the reference 16 Multiplication-Sign 16 cylindrical fuel assembly. In particular, the enhancements of the thermal hydraulic performance of dual-cooled annular fuel are estimated for the 100% normal power reactor core. The purpose of this study is to estimate a normal power for OPR-1000 with dual-cooled annular fuel, and ultimately to assess the feasibility of 120% core power. The parametric study was carried out for the fuel rod dimension, gap conductance, thermal diffusion coefficients, and pressure loss of the spacer grids. As a result of the analysis on the nominal power, annular fuel showed a sufficient margin available on DNB and fuel pellet temperature relative to cylindrical fuel. The margin amount seems accommodating a 20% power-uprate seems viable.

  20. NUCLEAR REACTOR FUEL ELEMENT

    Science.gov (United States)

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

    1961-09-01

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

  1. Nuclear fuel element

    International Nuclear Information System (INIS)

    Purpose: To reduce the probability of stress corrosion cracks in a zirconium alloy fuel can even when tensile stresses are resulted to the fuel can. Constitution: Sintered nuclear fuel pellets composed of uranium dioxide or a solid solution of gadolinium as a burnable poison in uranium dioxide are charged in a tightly sealed zirconium alloy fuel can. The nuclear fuel pellets for the nuclear fuel element are heat-treated in a gas mixture of carbon dioxide and carbon monoxide. Further, a charging gas containing a mixture of carbon dioxide and carbon monoxide is charged within a zirconium alloy fuel can packed with the nuclear fuel pellets and tightly sealed. (Aizawa, K.)

  2. The analysis of the annular fuel performance in steady state condition by using AFPAC code

    International Nuclear Information System (INIS)

    The fuel performance code AFPAC v1.0 is used to analyze annular fuel's behavior under steady state conditions, including neutronics, thermal hydraulic, rod deformation, fission gas release and rod internal pressure. The calculation results show that: 1) Annular fuel has a good steady irradiation performance at 150% power level as current LWRs' with burnup up to 50 GWd/t, and all parameters, such as temperature, rod internal pressure and rod deformation, are meet the rod design criteria for current fuel of PWRs: 2) Compared to the solid fuel under the same irradiation condition. annular fuel has lower temperature, smaller deformation, lower fission gas release and lower pressure at EOL. From the point of view of steady irradiation performance, the safety of reactors can significantly improved by u sing the annular fuel. (authors)

  3. Enhancing VVER annular proliferation resistance fuel with minor actinides

    International Nuclear Information System (INIS)

    reactivity control of the systems into which they are incorporated. In the study, a typical pressurized water reactor (PWR) VVER-1000 annular fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate term goal for future nuclear energy systems

  4. Fabrication of Annular Pellet for HANARO Irradiation Test of Dual Cooled Fuel

    International Nuclear Information System (INIS)

    One of the most important components in a Pressurized Water Reactor affecting its safety and economy is a nuclear fuel. The traditional PWR fuel pellet has a shape of cylindrical tablets of about 8 mm in diameter with a chamfer and dishes. A significant reduction in its failure rate has resulted from the improvements in the fuel and cladding quality. Enhanced fuel assembly design allowed appreciable power density increases. However, it is difficult to achieve a significant increase of a power density under the current fuel pin design. An internally and externally cooled annular fuel has been considered seriously as a promising solution for an extended power uprate of a PWR fuel assembly. A dual cooled annular fuel shows a lot of advantages from the point of a fuel safety and its economy due to its unique configurational merit such as an increased heat transfer area and a thin pellet thickness. There must be a lot of considerations in the various fields to introduce an annular internally and externally cooled fuel to commercial PWR reactors. The dimensional changes of the annular fuel pellets during the early irradiation stage are very important, because they have an influence on the size of the gap between the pellet and the inner/outer claddings. In order to gain an insight to how the annular pellets deform, a HANARO irradiation test is planned for annular pellets with 5 different types. The detailed specification of the annular pellet was shown in Table 1. It is noted that Type C has the same pore structure as a commercial PWR pellet. The purpose of this paper is to report on the manufacturing process of an annular fuel pellet for a HANARO irradiation test

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  6. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  7. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

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

  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. A Parametric Study on the Thermal Hydraulic Design for an Annular Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. H.; Seo, K. W.; Chun, T. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    Recently, MIT proposed an internally and externally cooled annular fuel for an advanced PWR which can endure a substantial power uprating. To apply this annular fuel in the conventional reactors such as OPR-1000, it is desirable to investigate its a structural compatibility for its reloading to operating PWR reactors of OPR-1000 as well as other compatibilities like the fuel to moderator ratio, amount of fissile material and coolant flow area. Conventional fuel assembly has a 16x16 solid rod array with four big guide tubes and one instrumentation tube. A 12x12 annular fuel assembly design which can meet the above compatibilities was proposed, which is structurally compatible with the existing internals of OPR-1000. Actually the advantage of an annular fuel comes from the fuel performance and thermal hydraulics. In the thermal hydraulic analysis, the mixing effect between the neighboring channels has to be carried out in a subchannel analysis. A subchannel analysis code, MATRA has been developed by KAERI. However, MATRA dose not have the capability to model both an internally and externally cooled annular fuel. A subchannel code, MATRA-AF which can be coupled to MATRA and can calculate the coolant flow distribution and heat transfer fraction in the internal and external subchannels has been developed. In this paper, the characteristics and the verification of the MATRA-AF are described. The effects of the thermal hydraulic parameters are estimated through a single fuel assembly.

  10. A Parametric Study on the Thermal Hydraulic Design for an Annular Fuel Assembly

    International Nuclear Information System (INIS)

    Recently, MIT proposed an internally and externally cooled annular fuel for an advanced PWR which can endure a substantial power uprating. To apply this annular fuel in the conventional reactors such as OPR-1000, it is desirable to investigate its a structural compatibility for its reloading to operating PWR reactors of OPR-1000 as well as other compatibilities like the fuel to moderator ratio, amount of fissile material and coolant flow area. Conventional fuel assembly has a 16x16 solid rod array with four big guide tubes and one instrumentation tube. A 12x12 annular fuel assembly design which can meet the above compatibilities was proposed, which is structurally compatible with the existing internals of OPR-1000. Actually the advantage of an annular fuel comes from the fuel performance and thermal hydraulics. In the thermal hydraulic analysis, the mixing effect between the neighboring channels has to be carried out in a subchannel analysis. A subchannel analysis code, MATRA has been developed by KAERI. However, MATRA dose not have the capability to model both an internally and externally cooled annular fuel. A subchannel code, MATRA-AF which can be coupled to MATRA and can calculate the coolant flow distribution and heat transfer fraction in the internal and external subchannels has been developed. In this paper, the characteristics and the verification of the MATRA-AF are described. The effects of the thermal hydraulic parameters are estimated through a single fuel assembly

  11. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    International Nuclear Information System (INIS)

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  12. Safety and Economics of High Power Density PWR with Novel Annular Fuel

    International Nuclear Information System (INIS)

    The internally and externally cooled annular fuel is a new type of fuel for PWRs that enables an increase in core power density by 50% within the same or better safety margins as the traditional solid fuel. Each assembly of traditional side dimensions has 160 annular fuel rods arranged in a 13x13 array. Even at the much higher power density, the fuel exhibits substantially lower temperatures and a MDNBR margin comparable to that of the traditional solid fuel at nominal (100%) power. Safety analyses indicate that the new annular fuel can accommodate 50% power up-rate in a PWR and still maintain adequate safety margins for a variety of transients and accidents including Loss of Flow Accident, Main Steam Line Break, Large Break Loss of Coolant Accident and Rod Ejection Accident. An economic study of 50% up-rate of an existing 1200 MW(e) PWR using the annular fuel shows that: (1) an Internal Rate of Return (IRR) on the order of 20% or more can be expected from such projects, even when accounting for uncertainties in the fuel price, electricity price inflation and cost of equipment; (2) Gradual replacement of the solid core by annular batches prior to up-rating can improve the IRR by 2.3% to 3.5% as it allows to full use of the energy in two already paid for batches of solid fuel rather than discarding them. Mixing of annular and solid fuel assemblies in one core appears feasible due to similar pressure drop characteristics of both assemblies. (authors)

  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 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)

  15. A Study on the Pressure Drop of a Subchannel Analysis Code for an Annular Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. H.; Seo, K. W.; In, W. K.; Chun, T. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    Recently, MIT proposed an internally and externally cooled annular fuel for an advanced PWR which can endure a substantial power uprating. KAERI is pursuing the development for its reloading to operating PWR reactors of OPR-1000. Thermal hydraulic analysis is critical part of annular fuel design because it determines dimensions of the fuel within acceptable MNDBR margins. An annular fuel subchannel analysis code, MATRA-AF which can be coupled to MATRA and can calculate the coolant flow split and heat split in the internal and external subchannels has been developed. In this paper, the effects of the parameters related with a calculation of a single-phase and two-phase pressure drop have been estimated.

  16. Nuclear fuel element cladding

    International Nuclear Information System (INIS)

    Composite cladding for a nuclear fuel element containing fuel pellets is formed with a zirconium metal barrier layer bonded to the inside surface of a zirconium alloy tube. The composite tube is sized by a cold working tube reduction process and is heat treated after final reduction to provide complete recrystallization of the zirconium metal barrier layer and a fine-grained microstructure. The zirconium alloy tube is stress-relieved but is not fully recrystallized. The crystallographic structure of the zirconium metal barrier layer may be improved by compressive deformation such as shot-peening. (author)

  17. Stress Intensity Factor using Finite Element Analysis in Rectangular Orthotropic Composite Annular Disk

    Directory of Open Access Journals (Sweden)

    P. Ravinder Reddy

    1997-01-01

    Full Text Available The quadratic isoparametric elements which embody the inverse squareroot singularity were used to determine the stress intensity factor in an annular disk made of Boron-Epoxy composite material. The displacements and stresses were determined in a rectangular orthotropic composite annular disk using isoparametric finite elements. The singularity in the strain field was provided by means of 8-noded isoparametric elements (4-nodes at the four corners and four mid-side nodes each at l/4th distance from the edge. The results were obtained for various material properties and fibre orientation. The geometry of the annular disk was reported when subjected to a boundary radial and tangential. The r singularity was provided at the boundary of the circular hole and the rest of the annular disk was modelled with ordinary isoparametric elements. The apparent stress intensity factor (K/sub I/= was computed from the stress data near the circular hole, when it was subjected to uniform tension. A curve was drawn for apparent stress intensity factor versus the distance from the crack edge and was extrapolated to r = 0, the actual stress intensity factor was found on the y-axis.

  18. Development of thermal hydraulic analysis code for nuclear reactors with annular fuels and assessment of the KAIST DNB-type theoretical critical heat flux model

    International Nuclear Information System (INIS)

    annular fuel design when operated at 50% higher power density resulted in about 1400 .deg. C lower hot spot fuel temperature with 26.5% higher mass flux in the inner sub-channel. GCRs have been highlighted as a promising option for next generation reactor technology. A thermal hydraulic analysis code for GCRs has been developed with a heat transfer model of a block element which is solved implicitly with the helium energy equation. Validation was carried out through comparison with both experimental and analytical results. A computation module for annular fuel rods has been coupled to the code for comparative analyses of an annular fuel-based block element. At normal operation, the annular fuel shows 80 .deg. C lower peak temperature than the solid fuel for the same power in Japan's High Temperature Engineering Test Reactor (HTTR), even though the pressure drop is higher in the annular fuel. A general CHF prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis at the conceptual design stage for a new Pressurized Water Reactor (PWR). In this study, the KAIST liquid sub-layer dryout CHF prediction model for DNB region has been implemented in a sub-channel analysis code, and investigated for the method's possible use in a rod bundle environment with various non-uniform axial power shapes. The KAIST model showed comparable prediction capability to Lin's method for bottom-, center-, and top-peaked heat flux shapes. The KAIST model, without any correction factors or empirical constants, turned out to be suitable to fulfill the needs for a basis of a general CHF prediction method as compared to Lin's method and Westinghouse-3 (W-3) correlation

  19. The Dual Cooled Annular Test Fuel Analysis Report for HANARO Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong Sik; In, Wang Ki; Chun, Tae Hyun; Kim, Dae Ho; Bang, Je Geon; Song, Kun Woo; Taek, Chae Hee; Seo, Chul Gyo

    2008-09-15

    Thermal safety analysis of double cooled annular fuel was performed for irradiation test of HANARO. Test fuel surface temperature and maximum fuel temperature analysis were calculated. The fuel surface temperature reached about 105 .deg. C, but still very lower than ONB limit temperature of 125 .deg. C. The maximum fuel temperature reached up to 1014 .deg. C but there was great margin with compare to UO{sub 2} melting temperature({approx}2800 .deg. C). The test fuel safety analysis results which examined by ONB and DNBR analysis shows that there is great thermal margins when compared HANARO ONB criteria.

  20. Diametric Tolerance Control of Dual Cooled Annular Fuel Pellet without Inner Surface Grinding

    International Nuclear Information System (INIS)

    A dual cooled fuel consists of internal and external cladding tubes in which annular pellets are stacked and cooling water flows in both internal and external coolant passages. It is recently being reconsidered as a promising option for a power up-rate of a pressurized water reactor fuel assembly because an annular fuel shows a lot of advantages from the point of a fuel safety and its economy due to an increased heat transfer area and a thin pellet thickness. Many technical issues might cause a serious problem to adopt the dual cooled annular fuel to the commercial PWR reactors. One of the most important issues is a heat flux split toward an internal cladding and an external cladding due to the gap conductance asymmetry which results from a preferential expansion of a fuel pellet toward the outside during an irradiation. Gap conductance is directly related to the inner and outer gap thicknesses. Initial gap thicknesses can vary with a pellet's dimensions which are affected by a reactor operation condition. Recently, it is suggested that a fuel rod with a smaller inner gap and a larger outer gap can reduce this gap conductance asymmetry. This approach can be effective only after precise tolerance technology is achieved. Because of an inhomogeneous green density distribution along the compact height, an hour-glassing usually occurred in a sintered cylindrical PWR fuel pellet fabricated by a conventional double-acting press. Thus, a sintered pellet usually undergoes a center-less grinding process in order to secure a pellet's specifications. In the case of an annular pellet fabrication using a conventional double-acting press, the same hour-glass shape would probably occur. The outer diameter tolerance of an annular pellet can be controlled easily similar to that of a conventional cylindrical PWR pellet through a center-less grinding. However, it appears not to be simple in the case of an inner surface grinding. It would be the best way to satisfy the specifications

  1. Development of Technology for Improving the Dual Cooling Annular Fuel Pellet Heat Transfer

    International Nuclear Information System (INIS)

    The purpose of this project is to conduct CHF experiments using nano fluid and to check the application possibility of nano fluid to annular fuel for developing high performance dual cooling annular fuel pellet. To achieve this purpose, We set the direction of research by literature survey and conducted experiments using various experimental apparatus. The main purposes of the experiments contained in the present study are understanding about effect of nano fluid on CHF and investigation of related phenomena. CHF enhancement by nano fluid can increase the the thermal margin of dual cooling annular fuel and thus increase the application possibility of annular fuel to nuclear power plant. The present study consist of two parts. First, we study about the effect of nano fluid on thermal conductivity, wettability, CHF in pool boiling condition. Second, we study about the effect of nano fluid on CHF in flow boiling condition. Part 1 : Thermal conductivity, wettability, CHF experiments using nano fluid in pool boiling condition Part 2 : CHF experiments using nano fluid in flow boiling condition

  2. 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.

  3. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

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

  4. Fuel distribution measurements in a model low NOx double annular combustor using laser induced fluorescence

    OpenAIRE

    Lockett, R. D.; Greenhalgh, D.A.

    2010-01-01

    Planar laser induced fluorescence (PLIF) was employed in a three sector, low NOx double annular combustor in order to characterise the combusting fuel spray. Naphthalene was employed as a fluorescent agent in odourless kerosene in order to determine the behavior of the light fractions in the fuel vapour, and the light to medium fractions in the fuel spray, while 2,5 di-phenyl oxizol (ppo) was employed to determine the behavior of the heavy fractions in the fuel spray. Counter-swirl air blast ...

  5. 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)

  6. 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)

  7. 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

  8. 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.)

  9. Rack for nuclear fuel elements

    International Nuclear Information System (INIS)

    Disclosed is a rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed spent fuel elements. Each fuel element is supported at the lower end thereof by a respective support that rests on the floor of the spent fuel pool for a nuclear power plant. An open rack frame is employed as an upright support for the enclosures containing the spent fuel elements. Legs at the lower corners of the frame rest on the floor of the pool to support the frame. In one exemplary embodiment, the support for the fuel element is in the form of a base on which a fuel element rests and the base is supported by legs. In another exemplary embodiment, each fuel element is supported on the pool floor by a self-adjusting support in the form of a base on which a fuel element rests and the base rests on a ball or swivel joint for self-alignment. The lower four corners of the frame are supported by legs adjustable in height for leveling the frame. Each adjustable frame leg is in the form of a base resting on the pool floor and the base supports a threaded post. The threaded post adjustably engages a threaded column on which rests the lower end of the frame. 16 claims, 14 figures

  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. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

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

  12. 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.)

  13. 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.)

  14. Thermal hydraulic analysis of thorium fuel assemblies loaded with annular seed pins

    International Nuclear Information System (INIS)

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using MATRAA combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and MATRAA showed good agreements for the pressure drops at the internal subchannels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner subchannels of the seed pins, mass fluxes were high due to the grid form losses in the outer subchannels. About 43% of the heat generated from the seed pin flowed into the inner subchannel and the rest into the outer subchannel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 qC. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that interchannel mixing cannot occur in the inner subchannels, temperatures and enthalpies were higher in the inner subchannels

  15. 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.

  16. 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

  17. The Design and Manufacturing Report of Non-Instrumented Rig for Dual-cooled Annular Fuel Irradiation Test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Bang, Je Geon; Lim, Ik Sung; Kim, Sun Ki; Yang, Yong Sik; Song, Kun Woo; Seo, Chul Gyo; Park, Chan Kook

    2008-09-15

    This project is preparing to irradiation test of the developed double cooled annular fuel pellet in HANARO for pursuit advanced performance in High Performance Fuel Technology Development as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented rig designed and manufactured for irradiation test in HANARO OR hole. This non- instrumented rig was confirmed the compatibility of HANARO and the integrity of rig structure, and satisfied the quality assurance requirements. This non- instrumented rig is adopt to the irradiation test for double cooled annular fuel pellet in HANARO.

  18. 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.)

  19. Annular Flow Distribution test

    International Nuclear Information System (INIS)

    This report documents the Babcock and Wilcox (B ampersand W) Annular Flow Distribution testing for the Savannah River Laboratory (SRL). The objective of the Annular Flow Distribution Test Program is to characterize the flow distribution between annular coolant channels for the Mark-22 fuel assembly with the bottom fitting insert (BFI) in place. Flow rate measurements for each annular channel were obtained by establishing ''hydraulic similarity'' between an instrumented fuel assembly with the BFI removed and a ''reference'' fuel assembly with the BFI installed. Empirical correlations of annular flow rates were generated for a range of boundary conditions

  20. Annular force based variable curvature mirror aiming to realize non-moving element optical zooming

    Science.gov (United States)

    Zhao, Hui; Xie, Xiaopeng; Wei, Jingxuan; Ren, Guorui; Pang, Zhihai; Xu, Liang

    2015-10-01

    Recently, a new kind of optical zooming technique in which no moving elements are involved has been paid much attention. The elimination of moving elements makes optical zooming suitable for applications which has exacting requirements in space, power cost and system stability. The mobile phone and the space-borne camera are two typical examples. The key to realize non-moving elements optical zooming lies in the introduction of variable curvature mirror (VCM) whose radius of curvature could be changed dynamically. When VCM is about to be used to implement optical zoom imaging, two characteristics should be ensured. First, VCM has to provide large enough saggitus variation in order to obtain a big magnification ratio. Second, after the radius of curvature has been changed, the corresponding surface figure accuracy should still be maintained superior to a threshold level to make the high quality imaging possible. In this manuscript, based on the elasticity theory, the physical model of the annular force based variable curvature mirror is established and numerically analyzed. The results demonstrate that when the annular force is applied at the half-the-aperture position, the actuation force is reduced and a smaller actuation force is required to generate the saggitus variation and thus the maintenance of surface figure accuracy becomes easier during the variation of radius of curvature. Besides that, a prototype VCM, whose diameter and thickness are 100mm and 3mm respectively, have been fabricated and the maximum saggitus variation that could be obtained approaches more than 30 wavelengths. At the same time, the degradation of surface figure accuracy is weakly correlated to the curvature radius variation. Keywords: optical zooming; variable curvature mirror; surface figure accuracy; saggitus;

  1. 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

  2. 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.

  3. Development of annular-coated-pressurized and sphere-pac LWR fuels

    International Nuclear Information System (INIS)

    Annular-coated (graphite)-pressurized and sphere-pac fuel rod designs, which are expected to exhibit improved PCI-failure resistance, and, thus, more reliable extended burnup performance, are being developed. Data sufficient to provide the technical bases needed to license lead test assemblies of the improved designs for irradiation in commercial LWRs are being obtained. Out-of-reactor experiments, in-reactor instrumented experiments, in-reactor power-ramp tests, and lead-rod demonstration irradiations are providing the needed data to support the technical bases. Results obtained to-date confirm the expected performance improvement compared with a solid-pellet reference design. The degree of improvement with respect to PCI-resistance remains to be quantified during forthcoming power-ramp tests on fuel rod segments irradiated to moderate burnup levels in a commercial LWR

  4. THE TESTING OF COMMERCIALLY AVAILABLE ENGINEERING AND PLANT SCALE ANNULAR CENTRIFUGAL CONTACTORS FOR THE PROCESSING OF SPENT NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Jack D. Law; David Meikrantz; Troy Garn; Nick Mann; Scott Herbst

    2006-10-01

    Annular centrifugal contactors are being evaluated for process scale solvent extraction operations in support of United State Advanced Fuel Cycle Initiative goals. These contactors have the potential for high stage efficiency if properly employed and optimized for the application. Commercially available centrifugal contactors are being tested at the Idaho National Laboratory to support this program. Hydraulic performance and mass transfer efficiency have been measured for portions of an advanced nuclear fuel cycle using 5-cm diameter annular centrifugal contactors. Advanced features, including low mix sleeves and clean-in-place rotors, have also been evaluated in 5-cm and 12.5-cm contactors.

  5. Auto-thermal combustion of lean gaseous fuels utilizing a recuperative annular double-layer catalytic converter

    Energy Technology Data Exchange (ETDEWEB)

    Budzianowski, W.M. [Wroclaw Univ. of Technology, Wroclaw (Poland). Div. of Chemical and Biochemical Processes, Faculty of Chemistry; Miller, R. [Wroclaw Univ. of Technology, Wroclaw (Poland). Inst. of Power Engineering and Fluid Mechanics, Faculty of Mechanical and Power Engineering

    2008-08-15

    This study investigated the auto-thermal combustion of lean gaseous fuels in a recuperative annular double-layer catalytic converter. An analysis of the stationary and transient performance of annular converters was presented. The feasibility of lean gaseous mixture combustion in auto-thermally operated recuperative annular double-layer catalytic converters was investigated. The aim of the study was to build a process model using mass, energy, and momentum differential balances. The model was used to study the static behaviour of a recuperative annular double layer converter; an annular converter operating in transient conditions; and energy accumulation and recuperation interactions. The effects of fuel temperature, external cooling, and fuel concentration were examined. Results of the study showed that a substantial reduction of the inter- and intra-phase resistances to mass and heat transfer was obtained. It was demonstrated that the use of a low value for the substrate's thermal conductivity accelerated ignition and retarded extinction. The recuperative converter was able to transfer short-time inlet disturbances of various parameters due to the energy accumulation and temporal reversed recuperation which counteracted destructive overheating of the catalysts. The stability analyses showed stable and unstable branches of solutions for the different parameters of the recuperative converter. 21 refs., 1 tab., 23 figs.

  6. Fuel element database: developer handbook

    International Nuclear Information System (INIS)

    The fuel elements database which was developed for Atomic Institute of the Austrian Universities is described. The software uses standards like HTML, PHP and SQL. For the standard installation freely available software packages such as MySQL database or the PHP interpreter from Apache Software Foundation and Java Script were used. (nevyjel)

  7. Experimental and numerical investigation of an entrance blockage of an inner channel in dual-cooled annular nuclear fuel

    International Nuclear Information System (INIS)

    A dual-cooled annular nuclear fuel for a Pressurized Water Reactor (PWR) has been introduced for a significant increase in reactor power. The Korea Atomic Energy Research Institute (KAERI) has been researching the development of a dual-cooled annular fuel for a power increase in an optimized PWR in Korea, OPR-1000. The main advantage of a dual-cooled annular fuel is an increased heat transfer area and a reduction in the fuel temperature, which would result in reduced fission gas release and increased fuel melting margin and Departure from Nucleate Boiling (DNB) margin. The annular fuel rod is configured to allow the coolant flow through the inner channel as well as outer channel. Since the inner channel is isolated from the neighbor channels unlike the outer channels, an inner channel blockage is one of the principal technical issues of a dual-cooled annular fuel. Due to a partial blockage, the inner channel may be faced with a DNB accident. A conceptual design used to complement the entrance blockage of an inner channel was suggested by KAERI. The through holes in this design are formed on a cylindrical wall of the lower end plug. When the inner channel is blocked by debris, coolant for the inner channel will be supplied through the side holes. But due to very unusual shape of the lower end plug, it is difficult to estimate the flow resistance of the side flow holes using empirical correlations available in the open literatures. Experimental and Computational Fluid Dynamics (CFD) studies were performed to investigate the bypass flow through the side holes of the lower end plug to complement the entrance blockage of an inner channel. The form loss coefficient in the side holes was also estimated by using the pressure drop along the bypass flow path and DNB Ratio (DNBR) margin was estimated by a subchannel analysis code. (author)

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensor for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280cal g{sup -1} UO{sub 2}. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (author)

  10. 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.)

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

    International Nuclear Information System (INIS)

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

  12. 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.

  13. Nuclear fuel element and container

    International Nuclear Information System (INIS)

    The invention is based on the discovery that a substantial reduction in metal embrittlement or stress corrosion cracking from fuel pellet-cladding interaction can be achieved by the use of a copper layer or liner in proximity to the nuclear fuel, and an intermediate zirconium oxide barrier layer between the copper layer and the zirconium cladding substrate. The intermediate zirconia layer is a good copper diffusion barrier; also, if the zirconium cladding surface is modified prior to oxidation, copper can be deposited by electroless plating. A nuclear fuel element is described which comprises a central core of fuel material and an elongated container using the system outlined above. The method for making the container is again described. It comprises roughening or etching the surface of the zirconium or zirconium alloy container, oxidizing the resulting container, activating the oxidized surface to allow for the metallic coating of such surfaces by electroless deposition and further coating the activated-oxidized surface of the zirconium or zirconium alloy container with copper, iron or nickel or an alloy thereof. (U.K.)

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

    International Nuclear Information System (INIS)

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

  15. Evaluation of loss coefficient for an end plug with side holes in dual cooled annular nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang Hwan; Chun, Tae Hyun; Oh, Dong Seok; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The Korea Atomic Energy Research Institute (KAERI) has been developing a dual cooled annular fuel for a power uprate of 20% in an optimized pressurized water reactor (PWR) in Korea, OPR1000. The dual cooled annular fuel is configured to allow coolant flow through the inner channel as well as the outer channel. Several thermal hydraulic issues exist for the application of dual cooled annular fuel to OPR1000. One is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause a departure from nucleate boiling (DNB) in the inner channel that eventually results in fuel failure. A long lower end plug for the annular fuel was invented to provide flow holes by perforating the side surface of the end plug body. The side holes in the lower end plug are expected to supply a minimum coolant in the inner channel to prevent the DNB occurrence in the event of partial or even complete blockage of the inner channel entrance. But due to the very unusual shape of the lower end plug, it is difficult to estimate the flow resistance of the side flow holes using empirical equations available in the open literature. An experiment and computational fluid dynamics (CFD) analysis were performed to investigate the bypass flow through the side holes of the end plug in the case of complete entrance blockage of the inner channel. The form loss coefficient in the side holes was also estimated using the pressure drop along the bypass flow path.

  16. Fuel elements of thermionic converters

    International Nuclear Information System (INIS)

    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

  17. System for assembling nuclear fuel elements

    International Nuclear Information System (INIS)

    An automatic system is described for assembling nuclear fuel elements, in particular those employing mixed oxide fuels. The system includes a sealing mechanism which allows movement during the assembling of the fuel element along the assembly stations without excessive release of contaminants. (U.K.)

  18. Development of dual cooled annular fuel and its possibility to enhance both economy and safety of light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yanghyun; Kim, Keonsik; Park, Jeongyong; Yang, Yongsik; Kim, Hyungkyu; In, Wangkee; Song, Kunwoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-03-15

    Over the past few decades, extensive studies have been performed to improve the reliability and safety of light water reactor (LWR) fuel. In recent years, power updating of about 10% is being achieved by modifying safety analysis methodology and subsequent increase in safety margin. But departure from nucleate boiling (DNB) and loss of coolant accident (LOCA) are still two of the most important limiting factors which would restrict power updating more than 10%. Duel cooled annular fuel, cooled in both internal and external cooling channel, has advantages of considerably lower heat flux and lower fuel temperature than conventional solid fuel. While lower heat flus gives higher DNB margin for the same power retie, lower temperature reduces the stored energy of fuel. However, there are many technical issues that should be addressed before any new type of fuel can be considered for application to LWR. This paper describes the key technologies that Korea Atomic Energy Research Institute (KAERI) has developed for dual cooled annular fuel and discusses the feasibility of its application to LWR.

  19. Development of dual cooled annular fuel and its possibility to enhance both economy and safety of light water reactor

    International Nuclear Information System (INIS)

    Over the past few decades, extensive studies have been performed to improve the reliability and safety of light water reactor (LWR) fuel. In recent years, power updating of about 10% is being achieved by modifying safety analysis methodology and subsequent increase in safety margin. But departure from nucleate boiling (DNB) and loss of coolant accident (LOCA) are still two of the most important limiting factors which would restrict power updating more than 10%. Duel cooled annular fuel, cooled in both internal and external cooling channel, has advantages of considerably lower heat flux and lower fuel temperature than conventional solid fuel. While lower heat flus gives higher DNB margin for the same power retie, lower temperature reduces the stored energy of fuel. However, there are many technical issues that should be addressed before any new type of fuel can be considered for application to LWR. This paper describes the key technologies that Korea Atomic Energy Research Institute (KAERI) has developed for dual cooled annular fuel and discusses the feasibility of its application to LWR

  20. 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

  1. 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

  2. Micro fuel elements and fuel elements studies with the use of pre-irradiation

    International Nuclear Information System (INIS)

    The ampoule and loop canal designs for irradiation of HTGR fuel elements and methods of investigation of their radiation stability are described. The results are presented on the measurement of fission product yield from fuel elements during irradiation. Irradiation main parameters are in agreement with HTGR operating conditions. The results of metallographic investigations of the micro fuel elements irradiated are given. The processes taking place in fuel elements and microfuel elements during irradiation are discussed

  3. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  5. Handling and inspection of nuclear fuel elements

    International Nuclear Information System (INIS)

    The invention provides improvements in the handling and inspection of nuclear fuel elements. A mobile bridge is mounted astraddle over a water tank, and from said bridge is suspended and immersed insulating plate capable of vertically receiving a fuel element and of taking a horizontal position for inspecting the latter. This can be applied to nuclear power stations

  6. Nuclear reactor fuel elements charging tool

    International Nuclear Information System (INIS)

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

  7. Fuel development program of the nuclear fuel element centre

    International Nuclear Information System (INIS)

    Fuel technology development program pf the nuclear fuel element centre is still devised into two main pillars, namely the research reactors fuel technology and the power reactor fuel technology taking into account the strategic influencing environment such as better access to global market of fuel cycle services, the state of the art and the general trend of the fuel technology in the world. Embarking on the twenty first century the fuel development program has to be directed toward strengthening measure to acquire and self-reliance in the field of fuel technology in support to the national energy program as well as to the utilisation of research reactor. A more strengthened acquisition of fuel cycle technology, in general, and particularly of fuel technology would improve the bargaining power when negotiation the commercial fuel technology transfer in the future

  8. 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)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kjellstroem, B.; Larsson, A.E.

    1967-04-15

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

  11. Soreq Nuclear Reactor Fuel Element Flow Distribution

    International Nuclear Information System (INIS)

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

  12. 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

  13. 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

  14. Combustion of liquid fuel in the counter-swirled jets of a gas turbine plant annular combustion chamber

    Science.gov (United States)

    Tumanovskii, A. G.; Semichastnyi, N. N.; Sokolov, K. Iu.

    1986-03-01

    Tests were carried out on an annular combustion chamber rig with a stabilizer of the type used in the GTN-25 gas turbine plant to determine the feasibility of burning a liquid fuel (diesel fuel, GOST 4749-73) in a combustion chamber of this type. Very high performance was obtained for a number of important characteristics of the microflame combustion process in counterswirled jets where all the air was supplied through the front unit of the chamber. However, the tests did not make it possible to solve some of the problems which arise when operating under full-scale conditions, such as the required high combustion efficiency under variable operating conditions of a gas turbine plant; elimination of soot formation at the walls of the stabilizer and the internal surfaces of the pipes supplying fuel to the atomizers; and a decrease in smoking under conditions of excess air factor.

  15. MRT fuel element inspection at Dounreay

    International Nuclear Information System (INIS)

    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

  16. 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.

  17. 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

  18. 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

  19. 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)

  20. Fundamental aspects of nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO2, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO2, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies

  1. Work on the development of the structure of fuel elements

    International Nuclear Information System (INIS)

    This paper is meant to give a roundup of development work concerning fuel element structure as support and cladding of fuel rods. The fuel element structure is a link between reactor vessel and the power-producing fuel rods, i.e. both the reactor arrangement and fuel rods influence the design of the fuel element structure, whereas the fuel element structure also determine marginal conditions for plant and fuel rods. (orig./RW)

  2. 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

  3. 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.)

  4. Benchmark Specification for HTGR Fuel Element Depletion

    International Nuclear Information System (INIS)

    There are currently several ongoing high-temperature gas-cooled reactor (HTGR) development projects underway throughout the world with the US DOE Next Generation Nuclear Plant (NGNP) representing a significant and growing activity in the United States. HTGR designs utilise graphite-moderated fuel forms and helium gas as a coolant. There are two main forms of HTGR fuels: pebbles are used in the pebble-bed reactor (PBR), while cylindrical rods (or compacts) are used in the modular high temperature gas-cooled reactor (MHTGR). In PBRs, fuel elements are ∼6-cm-diameter spheres; in MHTGRs, the fuel elements are graphite rods that are inserted into graphite hexagonal blocks. In both systems, fuel elements (spheres and rods) are comprised of tri-structural-isotropic (TRISO) fuel particles. The TRISO particles are either dispersed in with the matrix of a graphite pebble for the pebble bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks. In general, fuel grains have a density of a few hundred grains per cm3. The HTGR concept is a significant departure from LWR designs. As such, existing reactor analysis methods and data will be confronted by significant changes in the physics of neutron slowing down, absorption and scattering. Furthermore, the use of localised fuel grains within a larger fuel element result in two levels of heterogeneity that will challenge many existing lattice physics methods. Hence, there is a need for advanced methods for treatment of both levels of heterogeneity effects. In doubly-heterogeneous (DH) systems, heterogeneous fuel particles in a moderator matrix form the fuel region of the fuel element (pebble or rod) and thus constitute the first level of heterogeneity. Fuel elements themselves are also heterogeneous with fuel and moderator or reflector regions, forming the second level of heterogeneity. The fuel elements may also form regular or irregular lattices. Continuous energy (CE) methods are able to

  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

    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

  7. Safety assessment for Dragon fuel element production

    International Nuclear Information System (INIS)

    This report shall be the Safety Assessment covering the manufacture of the First Charge of Fuel and Fuel Elements for the Dragon Reactor Experiment. It is issued in two parts, of which Part I is descriptive and Part II gives the Hazards Analysis, the Operating Limitations, the Standing Orders and the Emergency Drill. (author)

  8. 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.

  9. Annular pancreas

    Science.gov (United States)

    ... page: //medlineplus.gov/ency/article/001142.htm Annular pancreas To use the sharing features on this page, please enable JavaScript. An annular pancreas is a ring of pancreatic tissue that encircles ...

  10. 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.)

  11. Developments in fabrication of annular MOX fuel pellet for Indian fast reactor

    International Nuclear Information System (INIS)

    Mechanical rotary presses along with adoption of core rod feature were inducted for fabrication of intricate annular Mixed Oxide (MOX) pellets for Prototype Fast Breeder Reactor (PFBR). In the existing tooling, bottom plungers contain core rod whereas top plungers contain a central hole for the entry of core rod during compaction. Frequent manual clean up of top plungers after few operations were required due to settling of powder in the annular hole of top plungers during compaction. Delay in cleaning can also result in breakage of tooling apart from increase in the dose to extremities of personnel. New design of tooling has been introduced to clean up the top plungers online during the operation of rotary press. It leads to increase in the productivity, reduces the spillage of valuable nuclear material and also reduces man-rem to operators significantly. The present paper describes the modification in tooling design and compaction sequence established for online cleaning of top plungers. (author)

  12. Stability analysis of a liquid fuel annular combustion chamber. M.S. Thesis

    Science.gov (United States)

    Mcdonald, G. H.

    1979-01-01

    The problems of combustion instability in an annular combustion chamber are investigated. A modified Galerkin method was used to produce a set of modal amplitude equations from the general nonlinear partial differential acoustic wave equation. From these modal amplitude equations, the two variable perturbation method was used to develop a set of approximate equations of a given order of magnitude. These equations were modeled to show the effects of velocity sensitive combustion instabilities by evaluating the effects of certain parameters in the given set of equations. By evaluating these effects, parameters which cause instabilities to occur in the combustion chamber can be ascertained. It is assumed that in the annular combustion chamber, the liquid propellants are injected uniformly across the injector face, the combustion processes are distributed throughout the combustion chamber, and that no time delay occurs in the combustion processes.

  13. 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)

  14. Comparisons of physical experiment and discrete element simulations of sheared granular materials in an annular shear cell

    Science.gov (United States)

    Ji, S.; Hanes, D.M.; Shen, H.H.

    2009-01-01

    In this study, we report a direct comparison between a physical test and a computer simulation of rapidly sheared granular materials. An annular shear cell experiment was conducted. All parameters were kept the same between the physical and the computational systems to the extent possible. Artificially softened particles were used in the simulation to reduce the computational time to a manageable level. Sensitivity study on the particle stiffness ensured such artificial modification was acceptable. In the experiment, a range of normal stress was applied to a given amount of particles sheared in an annular trough with a range of controlled shear speed. Two types of particles, glass and Delrin, were used in the experiment. Qualitatively, the required torque to shear the materials under different rotational speed compared well with those in the physical experiments for both the glass and the Delrin particles. However, the quantitative discrepancies between the measured and simulated shear stresses were nearly a factor of two. Boundary conditions, particle size distribution, particle damping and friction, including a sliding and rolling, contact force model, were examined to determine their effects on the computational results. It was found that of the above, the rolling friction between particles had the most significant effect on the macro stress level. This study shows that discrete element simulation is a viable method for engineering design for granular material systems. Particle level information is needed to properly conduct these simulations. However, not all particle level information is equally important in the study regime. Rolling friction, which is not commonly considered in many discrete element models, appears to play an important role. ?? 2009 Elsevier Ltd.

  15. Numerical analyses and experiment investigations of an annular micro gas turbine power system using fuels with low heating values

    Institute of Scientific and Technical Information of China (English)

    YANG; ChunHsiang; LEE; ChengChia; HSIAO; JenHao; CHEN; ChiunHsun

    2009-01-01

    This study investigates the effects of using fuels with low heating values on the performance of an annular micro gas turbine(MGT)experimentally and numerically.The MGT used in this study is MW-54, whose original fuel is liquid(Jet A1).Its fuel supply system is re-designed to use biogas fuel with low heating value(LHV).The purpose is to reduce the size of a biogas distributed power supply system and to enhance its popularization.This study assesses the practicability of using fuels with LHVs by using various mixing ratios of methane(CH4)and carbon dioxide(CO2).Prior to experiments,the corresponding simulations,aided by the commercial code CFD-ACE+,were carried out to investigate the cooling effect in a perforated combustion chamber and combustion behavior in an annular MGT when LHV gas was used.The main purposes are to confirm that there are no hot spots occurring in the liners and the exhaust temperatures of combustor are lower than 700°C when MGT is operated under different conditions.In experiments,fuel pressure and mass flow rate,turbine rotational speed,generator power output,and temperature distribution were measured to analyze MGT performance.Experimental results indicate that the presented MGT system operates successfully under each tested condition when the minimum heating value of the simulated fuel is approximately 50%of pure methane.The power output is around 170 W at 85000 r/min as 90%CH4 with 10%CO2 is used and 70 W at 60000 r/min as 70%CH4 with 30%CO2 is used.When a critical limit of 60%CH4 is used,the power output is extremely low. Furthermore,the best theoretical Brayton cycle efficiency for such MGT is calculated as 23%according to the experimental data while LHV fuel is used.Finally,the numerical results and experiment results reveal that MGT performance can be improved further and the possible solutions for performance im- provement are suggested for the future studies.

  16. Numerical analyses and experiment investigations of an annular micro gas turbine power system using fuels with low heating values

    Institute of Scientific and Technical Information of China (English)

    YANG ChunHsiang; LEE ChengChia; HSIAO JenHao; CHEN ChiunHsun

    2009-01-01

    This study investigates the effects of using fuels with low heating values on the performance of an annular micro gas turbine(MGT)experimentally and numerically.The MGT used in this study is MW-54,whose original fuel is liquid(Jet al).Its fuel supply system is re-designed to use biogas fuel with low heating value(LHV).The purpose is to reduce the size of a biogas distributed power supply system and to enhance its popularization.This study assesses the practicability of using fuels with LHVs by using various mixing ratios of methane(CH_4)and carbon dioxide(CO_2).Prior to experiments,the corresponding simulations,aided by the commercial code CFD-ACE+,were carried out to investigate the cooling effect in a perforated combustion chamber and combustion behavior in an annular MGT when LHV gas was used.The main purposes are to confirm that there are no hot spots occurring in the liners and the exhaust temperatures of combustor are lower than 700℃ when MGT is operated under different conditions,in experiments,fuel pressure and mass flow rate,turbine rotational speed,generator power output,and temperature distribution were measured to analyze MGT performance.Experimental results indicate that the presented MGT system operates successfully under each tested condition when the minimum heating value of the simulated fuel is approximately 50%of pure methane.The power output is around 170 W at 85000 r/min as 90%CH_4 with 10%CO_2 is used and 70 W at 60000 r/min as 70%CH_4 with 30%CO_2 is used.When a critical limit of 60%CH_4 is used,the power output is extremely low.Furthermore,the best theoretical Brayton cycle efficiency for such MGT is calculated as 23%according to the experimental data while LHV fuel is used.Finally,the numerical results and experiment results reveal that MGT performance can be improved further and the possible solutions for performance improvement are suggested for the future studies.

  17. 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

  18. 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

  19. Pressure loss coefficient and flow rate of side hole in a lower end plug for dual-cooled annular nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang-Hwan, E-mail: shinch@kaeri.re.kr; Park, Ju-Yong, E-mail: juyong@kaeri.re.kr; In, Wang-Kee, E-mail: wkin@kaeri.re.kr

    2013-12-15

    Highlights: • A lower end plug with side flow holes is suggested to provide alternative flow paths of the inner channel. • The inlet loss coefficient of the lower end plug is estimated from the experiment. • The flow rate through the side holes is estimated in a complete entrance blockage of inner channel. • The consequence in the reactor core condition is evaluated with a subchannel analysis code. - Abstract: Dual-cooled annular nuclear fuel for a pressurized water reactor (PWR) has been introduced for a significant increase in reactor power. KAERI has been developing a dual-cooled annular fuel for a power uprate of 20% in an optimized PWR in Korea, the OPR1000. This annular fuel can help decrease the fuel temperature substantially relative to conventional cylindrical fuel at a power uprate. Annular fuel has dual flow channels around itself; however, the inner flow channel has a weakness in that it is isolated unlike the outer flow channel, which is open to other neighbouring outer channels for a coolant exchange in the reactor core. If the entrance of the inner channel is, as a hypothetical event, completely blocked by debris, the inner channel will then experience a rapid increase in coolant temperature such that a departure from nucleate boiling (DNB) may occur. Therefore, a remedy to avoid such a postulated accident is indispensable for the safety of annular fuel. A lower end plug with side flow holes was suggested to provide alternative flow paths in addition to the central entrance of the inner channel. In this paper, the inlet loss coefficient of the lower end plug and the flow rate through the side holes were estimated from the experimental results even in a complete entrance blockage of the inner channel. An optimization for the side hole was also performed, and the results are applied to a subchannel analysis to evaluate the consequence in the reactor core condition.

  20. Annular dilatation and loss of sino-tubular junction in aneurysmatic aorta: Implications on leaflet quality at the time of surgery. A finite element study

    NARCIS (Netherlands)

    Weltert, L.; De Tullio, M.D.; Afferante, L.; Salica, A.; Scaffa, R.; Maselli, D.; Verzicco, R.; De Paulis, R.

    2013-01-01

    OBJECTIVES In the belief that stress is the main determinant of leaflet quality deterioration, we sought to evaluate the effect of annular and/or sino-tubular junction dilatation on leaflet stress. A finite element computer-assisted stress analysis was used to model four different anatomic condition

  1. Thermal analysis of nuclear fuel elements

    International Nuclear Information System (INIS)

    Full text: This work deals with the effect of non-uniform heat generation, non-uniform heat transfer conditions and variable thermophysical properties on the temperature and heat flux distribution in a rod type nuclear fuel element. The behaviour of maximum temperature in the fuel element under these conditions would be examined. Depending on complexity of different special cases, closed form analytical, approximate analytical (such as Poisson's integral, Fourier series and ∫kdT methods) and numerical methods have been employed. It is found that uniform heat generation only within the fuel pellet with constant thermophysical properties yields conservative estimation of fuel center-line temperature. But the temperature distribution predicted under other (more realistic) condition are duly useful for different thermodynamic and structural analyses

  2. 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

  3. 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

  4. HTGR fuel element structural design considerations

    International Nuclear Information System (INIS)

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabalistic stress analysis techniques coupled with probabalistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistent with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the USA is discussed in the context of stress analysis uncertainty and structural criteria development

  5. Fuel elements and safety engineering goals

    International Nuclear Information System (INIS)

    There are good prospects for silicon carbide anti-corrosion coatings on fuel elements to be realised, which opens up the chance to reduce the safety engineering requirements to the suitable design and safe performance of the ceramic fuel element. Another possibility offered is combined-cycle operation with high efficiencies, and thus good economic prospects, as with this design concept combining gas and steam turbines, air ingress due to turbine malfunction is an incident that can be managed by the system. This development will allow economically efficient operation also of nuclear power reactors with relatively small output, and hence contribute to reducing CO2 emissions. (orig./DG)

  6. Fuel element handling equipment for nuclear reactor

    International Nuclear Information System (INIS)

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

  7. 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

  8. Spherical coated particle fuel for fuel elements of HTGR

    International Nuclear Information System (INIS)

    The main results of the investigations on the development of spherical particles fuel for fuel elements of HTGR are described. Typical characteristics of UO2 spherical particles (size, shape, density, microstructure etc.) and PyC and SiC protective layers (thickness, density, fission product release etc.) are presented. Sol-gel technique and slip casting are used for spheroidization; deposition of protective layers is carried out in the fluidized bed apparatus

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

    International Nuclear Information System (INIS)

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

  10. Fuel elements for pulsed TRIGA research reactors

    International Nuclear Information System (INIS)

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

  11. Advanced sipping facilities for fuel elements

    International Nuclear Information System (INIS)

    The sipping facilities for BWR type plants and PWR type plants of the Russian type WWER-440 are equipped with a bell instead of caps, which is used above the opened reactor, moved by the fuel handling machine, and covers up to eight fuel elements in the core during inspection. In all sipping facilities, the complete inspection sequence is controlled by a desk switchboard near the fuel element storage pool or the reactor well. Siemens' sipping facilities are used in all Siemens-built nuclear power plants and in many others by different manufacturers. Part of them has been in operation already for more than 20 years with a high degree of reliability. Inspection safety is more than 99.5%. (orig./DG)

  12. Performance comparisons of Navy jet mix and MIL-F-5624A (JP-3) fuels in tubular and annular combustors / Richard J. McCafferty

    Science.gov (United States)

    Mccafferty, Richard J

    1954-01-01

    The performances of Navy Jet Mix and MIL-F-5624A (JP-3) fuels were compared in J33, J47, and NACA experimental annular combustion chambers. Combustion efficiencies, altitude operational limits, and carbon-forming tendencies wer determined and discussed. The results indicate that Jet Mix fuel can be utilized satisfactorily in a number of current turbojet engines over the normal operating range. Small differences in combustion efficiences and altitude operational limits existed between the two fuels but these differences depended on the particular combustor and simulated altitude and rotor speed condition. Excessive carbon deposition is not predicted for Jet Mix fuel, although this property may be marginal.

  13. 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

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

    International Nuclear Information System (INIS)

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

  15. Research and Test Reactor Fuel Elements (RTRFE)

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur is fabricating Uranium - Plutonium Mixed Oxide Fuel (MOX) for different types of reactors. Recently an experimental fuel subassembly of 37 pins has been fabricated for irradiation in Fast Breeder Test Reactor at Kalpakkam near Chennai. MOX fuel pins containing 45% PUO2 have also been also made for the hybrid core of FBTR. The experimental sub-assembly for irradiation testing in FBTR consists of 37 short length PFBR MOX fuel elements. The composition of the fuel was (0.71 U - 0.29 Pu) O2 with U233 O2 content of 53.5% of total UO 2. Uranium enriched with U233 was used to stimulate the heat flux of PFBR in FBTR neutron spectrum. MOX fuel pellets were made by powder metallurgy process consisting of pre-compaction, granulation, final compaction and sintering at high temperature. Initially U3233 O8 / U233 O3 powder was subjected to heat treatment. MOX powder were mixed, milled, pre-compacted and granulated. The final compaction was done using a multistation rotary press with suitable tooling for making annular MOX pellets. The technology for making annular pellets was developed for this purpose. The pellets were sintered at reducing atmosphere at 1650 deg. C for 4 hours to obtain acceptable quality pellets. Over sized pellets were centrelessly ground without using a liquid coolant. The acceptable pellets were degassed before encapsulation. MOX fuel stack, UO2 insulation pellets, plenum spring and spring support were loaded in bottom endplug welded clad tube. The end plug welding was carried out by TIG welding technique. The welded elements after inspection were wire wrapped. During the fabrication of pins for experimental subassembly, technology was developed and conditions were optimized for making annular pellets, TIG welding of D9 tubes with SS 316 end plugs and wire wrapping. Quality control procedures and process control procedures at different stages of fabrication were developed. The

  17. 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)

  18. Automatic inspection for remotely manufactured fuel elements

    International Nuclear Information System (INIS)

    Two classification techniques, standard control charts and artificial neural networks, are studied as a means for automating the visual inspection of the welding of end plugs onto the top of remotely manufactured reprocessed nuclear fuel element jackets. Classificatory data are obtained through measurements performed on pre- and post-weld images captured with a remote camera and processed by an off-the-shelf vision system. The two classification methods are applied in the classification of 167 dummy stainless steel (HT9) fuel jackets yielding comparable results

  19. Failure analysis for WWER-fuel elements

    International Nuclear Information System (INIS)

    If the fuel defect rate proves significantly high, failure analysis has to be performed in order to trace down the defect causes, to implement corrective actions, and to take measures of failure prevention. Such analyses are work-consuming and very skill-demanding technical tasks, which require examination methods and devices excellently developed and a rich stock of experience in evaluation of features of damage. For that this work specifies the procedure of failure analyses in detail. Moreover prerequisites and experimental equipment for the investigation of WWER-type fuel elements are described. (author)

  20. 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)

  1. 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)

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

    International Nuclear Information System (INIS)

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

  3. 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

  4. Fuel temperature characteristics of the 37-element and CANFLEX fuel bundle

    International Nuclear Information System (INIS)

    This report describes the fuel temperature characteristics of CANFLEX fuel bundles and 37-element fuel bundles for a different burnup of fuel. The program was consisted for seeking the fuel temperature of fuel bundles of CANFLEX fuel bundles and 37-element fuel bundles by using the method in NUCIRC. Fuel temperature has an increasing pattern with the burnup of fuel for CANFLEX fuel bundles and 37-element fuel bundles. For all the case of burnup, the fuel temperature of CANFLEX fuel bundles has a lower value than that of 37-element fuel bundles. Especially, for the high power channel, the CANFLEX fuel bundles show a lower fuel temperature as much as about 75 degree, and the core averaged fuel temperature has a lower fuel temperature of about 50 degree than that of 37-element fuel bundles. The lower fuel temperature of CANFLEX fuel bundles is expected to enhance the safety by reducing the fuel temperature coefficient. Finally, for each burnup of CANFLEX fuel bundles and 37-element fuel bundles, the equation was present for predicting the fuel temperature of a bundle in terms of a coolant temperature and bundle power

  5. 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

  6. 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)

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

    International Nuclear Information System (INIS)

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

  8. 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

  9. 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)

  10. Nuclear reactor and associated fuel element

    International Nuclear Information System (INIS)

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

  11. Impact tests with fuel element cans

    International Nuclear Information System (INIS)

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

  12. 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)

  13. 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.

  14. 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.)

  15. 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.

  16. Storage rack for long fuel elements

    International Nuclear Information System (INIS)

    The storage rack for PWR's usually has a lower grid plate, which has holes at the positions intended for fuel elements and stiffeners in the form of straight fins on the underside, which run flush in the direction of the midpoint of the holes. According to the invention, there are pieces of pipe on the underside of the plate concentric to all holes, which are connected by straight bars. This produces a stiffening just at the critical places. The invention can best be implemented in the form of a casting. (orig./HP)

  17. Facility modernization Annular Core Research Reactor

    International Nuclear Information System (INIS)

    The Annular Core Research Reactor (ACRR) has undergone numerous modifications since its conception in response to program needs. The original reactor fuel, which was special U-ZrH TRIGA fuel designed primarily for pulsing, has been replaced with a higher pulsing capacity BeO fuel. Other advanced operating modes which use this increased capability, in addition to the pulse and steady state, have been incorporated to tailor power histories and fluences to the experiments. Various experimental facilities have been developed that range from a radiography facility to a 50 cm diameter External Fuel Ring Cavity (FREC) using 180 of the original ZrH fuel elements. Currently a digital reactor console is being produced with GA, which will give enhanced monitoring capabilities of the reactor parameters while leaving the safety-related shutdown functions with analog technology. (author)

  18. 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

  19. 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.

  20. Neutron and thermo - hydraulic model of a reactivity transient in a nuclear power plant fuel element

    International Nuclear Information System (INIS)

    A reactivity transient without reactor scram was modeled and calculated using analytical expressions for the space distributions of the temperature fields, combined with discrete numerical calculations for the time dependences of thermal power and temperatures. The transient analysis covered the time dependencies of reactivity, global thermal power, fuel heat flux and temperatures in fuel, cladding and cooling water. The model was implemented in Microsoft Office Excel, dividing the Excel file in several separated worksheets for input data, initial steady-state calculations, calculation of parameters non-depending on eigenvalues, eigenvalues determination, calculation of parameters depending on eigenvalues, transient calculation and graphical representation of intermediate and final results. The results show how the thermal power reaches a new equilibrium state due to the negative reactivity feedback derived from the fuel temperature increment. Nevertheless, the reactor mean power increases 40% during the first second and, in the hottest channel, the maximum fuel temperature goes to a significantly high value, slightly above 2100 deg C, after 8 seconds of transient. Consequently, the results confirm that certain degree of fuel damage could be expected in case of a reactor scram failure. Once the basic model has being established the scope of accidents for future analyses can be extended, modifying the nuclear power behavior (reactivity) during transient and the boundary conditions for coolant temperature. A more complex model is underway for an annular fuel element. (author)

  1. 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.

  2. UK development of stage-2 CAGR fuel elements

    International Nuclear Information System (INIS)

    Britain has developed Stage-2 Commercial AGR fuel elements suitable for all AGR stations employing 71/2-inch bore fuel, and large-scale use will start in the initial charges of Heysham-II and Torness reactors

  3. Container for the storage of new fuel elements

    International Nuclear Information System (INIS)

    The fuel elements are placed vertically at defined positions of a store by a fixed support before introduction into the reactor. Each fuel element is surrounded for at least the length of its can by a box made of absorber material. This box is surrounded by a sleeve, which is fixed to the support so that it is easy to undo. The new store is particularly intended for highly enriched fuel elements. (orig./HP)

  4. Development and operating experience with new LWR fuel elements

    International Nuclear Information System (INIS)

    The Advanced Nuclear Fuels Corporation (ANF) supplies fuel elements and services for pressurized and boiling water reactors in Europe, the USA and the Far East. During the 19 years of its existence the ANF produced more than 16.300 fuel elements in the two manufacturing plants of Richland, USA and Lingen, FRG for 43 pressurized and boiling water reactors. In this context a series of innovations as regards the design of fuel cans, Zircaloy for spacers and Gd absorber in the fuel rod for the improvement of the operating behaviour of the elements was realized. (orig./DG)

  5. Stress analysis of coated particle fuel using finite element method

    International Nuclear Information System (INIS)

    The fuel element of high temperature gas-cooled reactor is composed of coated particle fuel which is dispersed in graphite matrix. In normal operation, the stress due to irradiation and a variety of complex physical and chemical reactions will cause failure of the coated particle fuel. Therefore, the stress analysis of coated particle fuel is important for the safety of fuel element and reactor. The stress was analyzed by the finite element method based on the inner pressure failure mechanism considering asphericity of the particles. (authors)

  6. Studies of direct final disposal of fuel elements

    International Nuclear Information System (INIS)

    The research and development programme for 'Direct Final Disposal' comprises works compiled for direct disposal of high-temperature fuel elements which, as regards the direct disposal of IWR fuel elements, are either carried out independently by the DWK (conditioning and development of tanks), or coordinated by the project group for Other Waste Disposal Techniques (PAE) of the KfK on behalf of the Federal Ministry of Research and Technology (repository). Part A of the research and development programme includes work on the direct disposal of high-temperature fuel elements. Part B comprises work on the direct disposal of LWR fuel elements. (orig./DG)

  7. 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

  8. Nuclear fuel element and method of manufacturing it

    International Nuclear Information System (INIS)

    Nuclear fuel pellets incorporating fission products capturing carbonaceous materials are disposed at upper and lower ends of a nuclear fuel element. Further, nuclear fuel pellets incorporating fission product capturing Zr-Cu series materials are disposed at the intermediate portion of the nuclear fuel element respectively. With such a constitution, fission products formed during burning of the nuclear fuel pellets are absorbed and kept by the fission product capturing materials incorporated in the nuclear fuel pellets, thereby enabling to reduce the amount of the fission products released. In addition, stress corrosion cracks caused by pellet/cladding tube interactions and dynamic interactions can be prevented. (T.M.)

  9. Three-dimensional numerical simulation of a continuously rotating detonation in the annular combustion chamber with a wide gap and separate delivery of fuel and oxidizer

    Science.gov (United States)

    Frolov, S. M.; Dubrovskii, A. V.; Ivanov, V. S.

    2016-07-01

    The possibility of integrating the Continuous Detonation Chamber (CDC) in a gas turbine engine (GTE) is demonstrated by means of three-dimensional (3D) numerical simulations, i. e., the feasibility of the operation process in the annular combustion chamber with a wide gap and with separate feeding of fuel (hydrogen) and oxidizer (air) is proved computationally. The CDC with an upstream isolator damping pressure disturbances propagating towards the compressor is shown to exhibit a gain in the total pressure of 15% as compared with the same combustion chamber operating in the deflagration mode.

  10. 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.

  11. 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

  12. Coherence of reactor design and fuel element design

    International Nuclear Information System (INIS)

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

  13. Fabrication of Confinement Facility of Failed Fuel Elements

    International Nuclear Information System (INIS)

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

  14. Granuloma annulare.

    Science.gov (United States)

    Gupta, Diptesh; Hess, Brian; Bachegowda, Lohith

    2010-01-01

    We present a case of a 77-year-old, diabetic male with a 20-year history of a migratory erythematous, asymptomatic, generalized, nonscaly, and nonitchy rash that started over the dorsum of his left hand. On examination, there were multiple annular erythematous plaques, distributed symmetrically and diffusely over his torso and arms, with central clearing and no scales. A punch biopsy of the skin helped us to arrive at the diagnosis of a generalized granuloma annulare (GA). GA is a benign, self-limiting skin condition of unknown etiology that is often asymptomatic. The cause of this condition is unknown, but it has been associated with diabetes mellitus, infections such as HIV, and malignancies such as lymphoma. These lesions typically start as a ring of flesh-colored papules that slowly progress with central clearing. Lack of symptoms, scaling, or associated vesicles helps to differentiate GA from other skin conditions such as tinea corporis, pityriasis rosea, psoriasis, or erythema annulare centrifugum. Treatment is often not needed as the majority of these lesions are self-resolving within 2 years. Treatment may be pursued for cosmetic reasons. Available options include high-dose steroid creams, PUVA, cryotherapy, or drugs such as niacinamide, infliximab, Dapsone, and topical calcineurin inhibitors. PMID:20209383

  15. Hollow fuel tablets for improvement of characteristics of rod fuel elements

    International Nuclear Information System (INIS)

    It is suggested to substitute compact fuel tablets for hollow ones. At that fuel temperature can be significantly reduced for equal thermal loadings. A lower fuel temperature when changing capacity results in decreasing thermal fuel expansion (reduction of mechanical stresses) as well as in decreasing the fission product release. Therefore, there is a possibility to improve the rod fuel element behaviour when changing linear power. Considerable reduction of fuel temperature in the hollow tablets with respect to the compact ones and a lesser energy content of a fuel element caused by its result in an additional advantage with respect to fuel behaviour during emergency leakage of coolant

  16. Determination of oxygen to uranium plus plutonium atom ratio in high density annular mixed oxide fuel pellets for fast reactor

    International Nuclear Information System (INIS)

    This paper highlights the encountered difficulties and applied modifications in the analytical steps for the determination of [O/(U+Pu)] in high density annular (NatU0.335233U0.37 Pu0.295)O2 pellets, manufactured for irradiation in FBTR and discusses the results. (author)

  17. Out-of-pile experiments on severe fuel damage behaviour of LWR fuel elements (CORA programme)

    International Nuclear Information System (INIS)

    The out-of-pile experiments of the CORA programme performed within the Project Group LWR Safety at the Kernforschungszentrum Karlsruhe are intended to provide information on the damage mechanisms of LWR fuel elements under severe fuel damage conditions, i.e. in the temperature region from 1200 deg. C to above 2000 deg. C. In these experiments the decay heat is simulated by electrical heating of a central tungsten rod within annular pellets, which are placed inside the Zircaloy cladding. The test bundle in the CORA facility is arranged from 16 heated (1000 mm length) and 9 unheated rods (solid pellets) surrounded by a Zircaloy shroud. The shroud itself is insulated by ZrO2 fibre insulation to obtain a uniform radial temperature distribution. In the test programme, 15 experiments are planned, 4 experiments have been performed. In the paper, the results of tests with A12O3 pellets and UO2 pellets with Inconel spacer only (no absorber material) are reported. In the tests with A12O3 pellets, simulating burnable poison rods, early melt formation at about 1350 deg C was observed. The liquefaction increases distinctly at 1500 deg. C. In the refrozen melt two metallic types - α-Zr(O) and (Zr,A1) alloy - and one porous ceramic (ZrO2, A12O3) eutectic can be distinguished. Large blockages form at the lower end of the bundle. In the tests with UO2 pellets, the melting starts at the elevation of the Inconel spacer. By eutectic melt formation in contact with the Zircaloy the liquefaction begins below the melting point of the Inconel. Further interaction of this melt with the UO2 results in partial dissolution of the pellets. Refreezing of the melt led to blockage formation at the lower end of the bundle, but at higher elevations compared to the tests with alumina pellets. At some locations fragmentation of fuel pellets to fine powder took place during cooldown. (author). 9 refs, 6 figs

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

    International Nuclear Information System (INIS)

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

  19. 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

  20. 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.)

  1. LOAD-CHECK. Disposal planning for LWR fuel elements

    International Nuclear Information System (INIS)

    With the changes of the German atomic law from November 8, 2011 the operation licensing of LWR plants expire latest 2022, for eight NPPs the operation licenses are already expired. In order to optimize the fuel element management in the still operated but also in the decommissioned nuclear power plants the computer code module LOAD-CHECK was developed. LOAD-CHECK allows the foresight container planning for an optimized schedule and the container amount for loading campaigns esp. in case of the disposal of special fuel elements (MOX fuel elements or high-burnup fuel elements). The program can also be used a s tool for development of transport licensing and storage licensing according of CASTOR registered V casks. In the contribution the LOAD-CHECK program for the PWR and BWR fuel element disposal management in CASTOR registered B casks is presented.

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

    International Nuclear Information System (INIS)

    In the fuel stack test section (T1) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T2). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs

  3. CARA, new concept of advanced fuel element for HWR

    International Nuclear Information System (INIS)

    All Argentinean NPPs (2 in operation, 1 under construction), use heavy water as coolant and moderator. With very different reactor concepts (pressure Vessel and CANDU type designs), the fuel elements are completely different in its concepts too. Argentina produces both types of fuel elements at a manufacturing fuel element company, called CONUAR. The very different fuel element's designs produce a very complex economical behavior in this company, due to the low production scale. The competitiveness of the Argentinean electric system (Argentina has a market driven electric system) put another push towards to increase the economical competitiveness of the nuclear fuel cycle. At present, Argentina has a very active Slightly Enriched Uranium (SEU) Program for the pressure vessel HWR type, but without strong changes in the fuel concept itself. Then, the Atomic Energy Commission in Argentina (CNEA) has developed a new concept of fuel element, named CARA, trying to achieve very ambitious goals, and substantially improved the competitiveness of the nuclear option. The ambitious targets for CARA fuel element are compatibility (a single fuel element for all Argentinean's HWR) using a single diameter fuel rod, improve the security margins, increase the burnup and do not exceed the CANDU fabrication costs. In this paper, the CARA concept will be presented, in order to explained how to achieve all together these goals. The design attracted the interest of the nuclear power operator utility (NASA), and the fuel manufacturing company (CONUAR). Then a new Project is right now under planning with the cooperation of three parts (CNEA - NASA - CONUAR) in order to complete the whole development program in the shortest time, finishing in the commercial production of CARA fuel bundle. At the end of the this paper, future CARA development program will be described. (author)

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

    International Nuclear Information System (INIS)

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

  5. 生物质固体成型燃料环模成型技术研究进展%The research progress in biomass annular mould forming for fuel technology

    Institute of Scientific and Technical Information of China (English)

    欧阳双平; 侯书林; 赵立欣; 田宜水; 孟海波

    2011-01-01

    综合分析了国内外生物质固体成型燃料环模成型技术、成型设备及产业发展现状.比较了生物质环模颗粒成型机和生物质环模压块成型机的性能和产品,指出了生物质固体成型燃料环模技术及设备存在着基础理论薄弱、原料适应差、易损件寿命短等问题;提出了我国生物质固体成型燃料环模成型技术的发展方向.%The research status of annular mould forming for fuel technology, forming equipment and the industrial development has been comprehensively analyzed, the performance and the product of annular mould biomass pellet machine and annular mould biomass briquette machine have been compared, the problems of biomass annular mould forming for fuel technology and equipment, such as weak theory foundation, poor feedstock suitability, short life span of wearing part, were pointed out, the development orientation of biomass annular mould forming for fuel technology in China was proposed.

  6. Electrochemical method to disintegrate spherical fuel elements of HTGR

    International Nuclear Information System (INIS)

    Spherical fuel elements of high temperature gas-cooled reactor are employed to demonstrate electrochemical method with NaNO3 as electrolyte after an overall study of simulative fuel elements. The X-ray diffraction and the total carbon content of graphite fragments were determined, and the results were in agreements with graphite fragments from simulative elements. The characterization and leaching experiments of coated fuel particles and the determination uranium of the recovery solutions were detected, the results of which demonstrated the integrity of coated fuel particles and no contamination to the graphite fragments. The present work indicates that the improved electrochemical method is a promising option to disintegrate graphite matrix from high temperature gas-cooled reactor spent fuel elements in the head-end process of reprocessing. (author)

  7. 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.)

  8. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

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

  9. Operational requirements of spherical HTR fuel elements and their performance

    International Nuclear Information System (INIS)

    The German development of spherical fuel elements with coated fuel particles led to a product design which fulfils the operational requirements for all HTR applications with mean gas exit temperatures from 700 deg C (electricity and steam generation) up to 950 deg C (supply of nuclear process heat). In spite of this relatively wide span for a parameter with strong impact on fuel element behaviour, almost identical fuel specifications can be used for the different reactor purposes. For pebble bed reactors with relatively low gas exit temperatures of 700 deg C, the ample design margins of the fuel elements offer the possibility to enlarge the scope of their in-service duties and, simultaneously, to improve fuel cycle economics. This is demonstrated for the HTR-500, an electricity and steam generating 500 MWeleq plant presently proposed as follow-up project to the THTR-300. Due to the low operating temperatures of the HTR-500 core, the fuel can be concentrated in about 70% of the pebbles of the core thus saving fuel cycle costs. Under all design accident conditions fuel temperatures are maintained below 1250 deg C. This allows a significant reduction in the engineered activity barriers outside the primary circuit, in particular for the loss of coolant accident. Furthermore, access to major primary circuit components and the reuse of the fuel elements after any design accident are possible. (author)

  10. 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.)

  11. 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)

  12. RITM device for fuel element testing under power ramping

    International Nuclear Information System (INIS)

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

  13. 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

  14. Management of Rossendorf research reactor spent fuel elements

    International Nuclear Information System (INIS)

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

  15. Licensing procedure for the Hanau fuel element fabrication plant

    International Nuclear Information System (INIS)

    Licensing procedure for the Hanau fuel element fabrication plant. The fuel element plant at Hanau fabricates at present fuel elements on the basis of licences according to para. 9 Atomic Energy Law. In 1975, however, it was decided to carry out a subsequent licensing procedure according to para. 7 Atomic Energy Law. This led to protracted proceedings before the Administrative Court and, in addition, to criminal proceedings against the managing director and officials. Most of the proceedings were settled in favor of the operator. The present state of partial licenses is described. (DG)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  18. 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)

  19. Measurement of fission gas release from irradiated nuclear fuel elements

    International Nuclear Information System (INIS)

    A fission gas measurement system for the analysis of released gases from MOX and PHWR fuels has been designed, fabricated and commissioned in the hot cells of Post Irradiation Examination Division of Bhabha Atomic Research Centre, Mumbai. The system was used for the measurement of fission gases released from natural UO2 fuels and ThO2 fuels from PHWRs. The burnups of these fuels ranged from 2 GWD/TeU to 15 GWD/TeU. Some of the results from PHWR fuel elements from Kakrapar Atomic Power Station are presented in the paper, to highlight the utility of the system. (author)

  20. Mechanical design and operating behaviour of advanced LWR fuel elements

    International Nuclear Information System (INIS)

    The development of fuel elements for pressurized and and boiling water reactors during the last years was marked by a reduction of the fuel cycle costs with security and reliability in operation remaining constant. The heightening of fuel discharge burnup and the improvement of neutron economy contributed essentially to that. The latter had been achieved by a reduction of the parasitic absorption within the fuel element and the leakage of neutrons of the reactor cores. These improvements could be obtained under complete observance of the safety-relevant requirements. Due to the change to fuel elements with a higher number of rods and correspondingly lower rod power it was even possible to raise the security margins partly. A survey of the state of experiences of Siemens/KWU is given. (orig./DG)

  1. CANDU fuel elements behaviour in the load following tests

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-08-15

    Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the fuel element withstood 200 power cycles from 27 to 54 kW/m average linear power as well as additional ramps due to reactor trips and restarts during test period. New LF tests are planed to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in TRIGA Research Reactor and their relation to CANDU fuel performance in LF conditions. (orig.)

  2. 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)

  3. 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)

  4. 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.

  5. Dynamic characterization of the CAREM fuel element prototype

    International Nuclear Information System (INIS)

    As a previous step to make a complete test plan to evaluate the hydrodynamic behavior of the present configuration of the CAREM type fuel element, a dynamic characterization analysis is required, without the dynamic response induced by the flowing fluid. This paper presents the tests made, the methods and instrumentation used, and the results obtained in order to obtain a complete dynamic characterization of the CAREM type fuel element. (author)

  6. 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

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

    International Nuclear Information System (INIS)

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

  8. Irradiation tasks within development of fuel elements in Sweden

    International Nuclear Information System (INIS)

    This report contains description of the hot laboratory RMA for irradiation in the R-2 reactor in Studsvik. Activities of the AB Atomenegiyu concerning irradiation and testing of fuel rods and fuel elements are described, as well as methods for testing of irradiated samples in hot cells. Concerning the importance of the problem, determination of burnup level and neutron flux were examined particularly

  9. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

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

  10. Transferencia de calor incrementada en espacios anulares con elementos helicoidales insertados//Review of augmentation techniques for heat transfer coefficient in annular spaces using helical elements

    Directory of Open Access Journals (Sweden)

    Josué Imbert‐González

    2014-08-01

    Full Text Available La transferencia de calor incrementada por métodos pasivos se emplea en diversosintercambiadores de calor de alta efectividad. El objetivo del trabajo presentado fue la evaluación del estado de las investigaciones en el campo de la transferencia de calor mejorada en espacios anulares, a partir del empleo de elementos turbulizadores helicoidales como técnicas pasivas. La revisión se centró en el empleo de láminas helicoidales y espirales, la obtención de ecuaciones de correlación del coeficiente de transferencia de calor incrementado, el coeficiente de fricción y la evaluación que se realiza de este proceso por parte de diferentes autores. El análisis crítico permitió realizar valoraciones integradas y recomendar sobre los aspectos que podrían ser analizados en el futuro en esta temática.Palabras claves: transferencia de calor incrementada, láminas helicoidales, espirales, espacios anulares, métodos pasivos._______________________________________________________________________________AbstractThe transfer enhancement by passive methods is used in several heat exchanger of high effectiveness. The objective of the presented work was the evaluation of the state of the investigations in heat transfer enhancement in annular spaces, from the employment of elements helical. The revision was centered in the employment of twisted tape and wire coil in spiral, the equations of correlation obtained of the coefficient of transfer of increased heat, the coefficient of friction and the evaluation that was carried out of this process on the part of different authors. From the critical analysis of the published results, the authors recommend on the topics that can be analyzed in the future in this area.Key words: heat transfer enhancement, twisted tape, helical springs, annular spaces, passive methods.

  11. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Sipping test on a failed MTR fuel element

    International Nuclear Information System (INIS)

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

  15. Sipping tests on a failed irradiated MTR fuel element

    International Nuclear Information System (INIS)

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

  16. 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

  17. Spent HIFAR fuel elements behaviour under extended dry storage

    International Nuclear Information System (INIS)

    Previously unpublished observations of the behaviour of HIFAR spent fuel under extended dry storage conditions are reported. The two fuel elements EC802 (Mark III type) were irradiated in 1966, first examined in hot cells in 1967 and again examined in hot cells in 1983 following 16 years of stage, 11 years of which were in the ANSTO engineered dry storage facility. The elements showed negligible deterioration over this extended dry storage period, lending considerable confidence to the viability of dry storage technologies for the long term storage of spent aluminium clad research reactor fuels. 1 tab., 1 fig., 17 ills

  18. Process for assembling a nuclear fuel element

    International Nuclear Information System (INIS)

    Before insertion into the spacers, the fuel rocks are coated with a self-hardening layer of water-soluble polyvinyl and/or polyether polymer to prevent scratches on the cladding tubes. After insertion, the protective conting is removed by means of water. (orig.)

  19. Handling system for nuclear reactor fuel and reflector elements

    International Nuclear Information System (INIS)

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

  20. Transuranium element recovering method for spent nuclear fuel

    International Nuclear Information System (INIS)

    Spent fuels are dissolved in nitric acid, the obtained dissolution liquid is oxidized by electrolysis, and nitric acid of transuranium elements are precipitated together with nitric acid of uranium elements from the dissolution solution and recovered. Namely, the transuranium elements are oxidized to an atomic value level at which nitric acid can be precipitated by an oxidizing catalyst, and cooled to precipitate nitric acid of transuranium elements together with nitric acid of transuranium elements, accordingly, it is not necessary to use a solvent which has been used so far upon recovering transuranium elements. Since no solvent waste is generated, a recovery method taking the circumstance into consideration can be provided. Further, nitric acid of uranium elements and nitric acid of transuranium elements precipitated and recovered together are dissolved in nitric acid again, cooled and only uranium elements are precipitated selectively, and recovered by filtration. The amount of wastes can be reduced to thereby enabling to mitigate control for processing. (N.H.)

  1. End plug welding of nuclear fuel elements-AFFF experience

    International Nuclear Information System (INIS)

    Advanced Fuel Fabrication Facility is engaged in the fabrication of mixed oxide (U,Pu)O2 fuel elements of various types of nuclear reactors. Fabrication of fuel elements involves pellet fabrication, stack making, stack loading and end plug welding. The requirement of helium bonding gas inside the fuel elements necessitates the top end plug welding to be carried out with helium as the shielding gas. The severity of the service conditions inside a nuclear reactor imposes strict quality control criteria, which demands for almost defect free welds. The top end plug welding being the last process step in fuel element fabrication, any rejection at this stage would lead to loss of effort prior to this step. Moreover, the job becomes all the more difficult with mixed oxide (MOX) as the entire fabrication work has to be carried out in glove box trains. In the case of weld rejection, accepted pellets are salvaged by cutting the clad tube. This is a difficult task and recovery of pellets is low (requiring scrap recovery operation) and also leads to active metallic waste generation. This paper discusses the experience gained at AFFF, in the past 12 years in the area of end plug welding for different types of MOX fuel elements

  2. Quality control in the fuel elements production process

    International Nuclear Information System (INIS)

    Recently great attention has been paid at the international level to the analysis of production processes and quality control of fuel and fuel elements with the aim to speed up activity of proposing and accepting standards and measurement methods. IAEA also devoted great interest to these problems appealing to more active participation of all users and producers fuel elements in a general effort to secure successful work of nuclear plants. For adequate and timely participation in future in the establishment and analysis of general requirements and documentation for the control of purchased or self produced fuel elements in out country it is necessary to be well informed and to follow this activity at the international level. (author)

  3. RA-3 core with uranium silicide fuel elements

    International Nuclear Information System (INIS)

    Following on with studies on uranium silicide fuel elements, this paper reports some comparisons between the use of standard ECN [U3O8] fuel elements and type P-06 [from U3Si2] fuel elements in the RA-3 core.The first results showed that the calculated overall mean burn up is in agreement with that reported for the facility, which gives more confidence to the successive ones. Comparing the mentioned cores, the silicide one presents several advantages such as: -) a mean burn up increase of 18 %; -) an extraction burn up increase of 20 %; -) 37.4 % increase in full power days, for mean burn up. All this is meritorious for this fuel. Moreover, grouped and homogenized libraries were prepared for CITVAP code that will be used for planning experiments and other bidimensional studies. Preliminary calculations were also performed. (author)

  4. Numerical simulations of heat transfer in an annular fuel channel with three-dimensional spacer ribs set up periodically under a fully developed turbulent flow

    International Nuclear Information System (INIS)

    Thermal-hydraulic characteristics of an annular fuel channel with spacer ribs for high temperature gas-cooled reactors were analyzed numerically by three-dimensional heat transfer computations under a fully developed turbulent flow. The two-equations κ-ε turbulence model was applied to the present turbulent analysis. In particular, the κ-ε turbulence model constants and the turbulent Prandtl number were improved from the previous standard values proposed by Jones and Launder in order to obtain heat transfer predictions with higher accuracy. Consequently, heat transfer coefficients and friction factors in the spacer-ribbed fuel channel were predicted with sufficient accuracy in the range of Reynolds number exceeding 3000. It was clarified quantitatively from the present study that main mechanism for the heat transfer augmentation in the spacer-ribbed fuel channel was combined effects of the turbulence promoter effect by the spacer ribs and the velocity acceleration effect by a reduction in the channel cross-section. (author)

  5. Numerical simulations of heat transfer in an annular fuel channel with three-dimensional spacer ribs set up periodically under a fully developed turbulent flow

    Energy Technology Data Exchange (ETDEWEB)

    Takase, Kazuyuki; Akino, Norio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1996-06-01

    Thermal-hydraulic characteristics of an annular fuel channel with spacer ribs for high temperature gas-cooled reactors were analyzed numerically by three-dimensional heat transfer computations under a fully developed turbulent flow. The two-equations {kappa}-{epsilon} turbulence model was applied to the present turbulent analysis. In particular, the {kappa}-{epsilon} turbulence model constants and the turbulent Prandtl number were improved from the previous standard values proposed by Jones and Launder in order to obtain heat transfer predictions with higher accuracy. Consequently, heat transfer coefficients and friction factors in the spacer-ribbed fuel channel were predicted with sufficient accuracy in the range of Reynolds number exceeding 3000. It was clarified quantitatively from the present study that main mechanism for the heat transfer augmentation in the spacer-ribbed fuel channel was combined effects of the turbulence promoter effect by the spacer ribs and the velocity acceleration effect by a reduction in the channel cross-section. (author)

  6. 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

  7. 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)

  8. Temperature distribution calculations in TRIGA fuel element after the pulse

    International Nuclear Information System (INIS)

    The computer program TEMPUL for calculating radial temperature distribution in a fuel element after the pulse operation is shortly described. It is based on one-dimensional diffusion equation for heat transfer in cylindrical geometry and implicit boundary condition at the element-coolant interface, defined by empirical boiling curve, which relates the heat flux from the rod and the difference between the fuel element surface temperature and water boiling point. As an example the results of such analysis of maximal allowed pulse at TRIGA Mark II reactor in Ljubljana are presented. (author)

  9. 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.

  10. 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

  11. 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.

  12. Algorithm and computer code for calculating the swelling of the fuel elements with a ceramic fuel

    International Nuclear Information System (INIS)

    Algorithm and the OVERAT program intended for calculating the strain deformed state of a cylindrical axially symmetric fuel element with ceramic fuel and thin-walled shell are described. Calculations are performed with account for creep deformation, fuel swelling, coolant and gas pressures in the axial cavity. At each moment of time deformations and strains in the shell as well as the spatial (by rod radius) dependence of fuel swelling are calculated. Fuel swelling is determined on the basis of a theoretical model, in which gas swelling is related to formation and development only of intergrain porosity. The reactor operation at a constant power at invariable in time temperature and energy release distributions in the fuel element core rod are considered. For description of the processes taking place in a fuel element a hard system of usual differential first order equations which is solved by the Gear method has been used. The OVERAT program is written in FORTRAN and at BESM-6 computer debuged. The results of test calculations of strain-deformed state and fuel element swelling with an UO2 hollow rod in a molybdenum shell are presented. It is pointed out that the described program in a complex with other programs can be used for investigating serviceability of various type reactors fuel elements

  13. 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)

  14. Commercial Aspect of Research Reactor Fuel Element Production

    International Nuclear Information System (INIS)

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

  15. Fire and blast safety manual for fuel element manufacture

    International Nuclear Information System (INIS)

    The manual aims to enable people involved in the planning, operation, supervision, licensing or appraisal of fuel element factories to make a quick and accurate assessment of blast safety. In Part A, technical plant principles are shown, and a summary lists the flammable materials and ignition sources to be found in fuel element factories, together with theoretical details of what happens during a fire or a blast. Part B comprises a list of possible fires and explosions in fuel element factories and ways of preventing them. Typical fire and explosion scenarios are analysed more closely on the basis of experiments. Part B also contains a list and an assessment of actual fires and explosions which have occurred in fuel element factories. Part C contains safety measures to protect against fire and explosion, in-built fire safety, fire safety in plant design, explosion protection and measures to protect people from radiation and other hazards when fighting fires. A distinction is drawn between UO2, MOX and HTR fuel elements. (orig./DG)

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

    International Nuclear Information System (INIS)

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

  17. 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.)

  18. 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.)

  19. Shock absorber for a fuel element storage rack

    International Nuclear Information System (INIS)

    The invention describes a shock absorber device for a nuclear fuel element deposited in a sheath provided with a bottom portion comprising centrally a hole of a diameter slightly larger than that of the lower portion of the fuel element, within a fuel storage rack, characterised in that it comprises a non-deformable annulus connected to a collar bearing on a transverse member of the storage rack, by means of a plurality of elastically and/or plastically deformable elements, and in that the non-deformable annulus, coaxial with the sheath, is provided with a central aperture having a diameter substantially equal to that of the hole in the bottom portion of the sheath and serves as a support for the bottom portion of the sheath

  20. The AVR as a test bed for fuel elements

    International Nuclear Information System (INIS)

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

  1. Analysis of the ATR fuel element swaging process

    International Nuclear Information System (INIS)

    This report documents a detailed evaluation of the swaging process used to connect fuel plates to side plates in Advanced Test Reactor (ATR) fuel elements. The swaging is a mechanical process that begins with fitting a fuel plate into grooves in the side plates. Once a fuel plate is positioned, a lip on each of two side plate grooves is pressed into the fuel plate using swaging wheels to form the joints. Each connection must have a specified strength (measured in terms, of a pullout force capacity) to assure that these joints do not fail during reactor operation. The purpose of this study is to analyze the swaging process and associated procedural controls, and to provide recommendations to assure that the manufacturing process produces swaged connections that meet the minimum strength requirement. The current fuel element manufacturer, Babcock and Wilcox (B ampersand W) of Lynchburg, Virginia, follows established procedures that include quality inspections and process controls in swaging these connections. The procedures have been approved by Lockheed Martin Idaho Technologies and are designed to assure repeatability of the process and structural integrity of each joint. Prior to July 1994, ATR fuel elements were placed in the Hydraulic Test Facility (HTF) at the Idaho National Engineering Laboratory (AGNAIL), Test Reactor Area (TRA) for application of Boehmite (an aluminum oxide) film and for checking structural integrity before placement of the elements into the ATR. The results presented in this report demonstrate that the pullout strength of the swaged connections is assured by the current manufacturing process (with several recommended enhancements) without the need for- testing each element in the HTF

  2. 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.

  3. 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)

  4. Irradiation of Fuel Elements in the Belgian BR3 Reactor

    International Nuclear Information System (INIS)

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

  5. 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.

  6. Review of the metallurgical studies carried out at the CEA on the in-service behaviour of the fuel elements of the gas graphite type

    International Nuclear Information System (INIS)

    After a brief review of the three types of fuel elements used in the EdF reactors and a description of working conditions, the principal stages of five metallurgical investigations concerning these elements are presented. In each of these studies, success was possible only by combining research into basic mechanisms with an effort to find quickly the most suitable solution for the operator. The following are described in turn: (a) the reduction of growth, creep, and swelling of uranium to acceptable levels by controlling the mechanisms of these three elementary modes of distortion and by checking any interaction; (b) how, in the case of annular element of the Bugey 1 reactor, a metallurgical bond between fuel and cans has considerably limited the possible consequences of a burst can; (c) the main lines of investigations into the release of fission gases usually occluded in the uranium, but which under certain conditions can produce an interruption in thermal continuity between can and fuel; (d) the problems raised by the diffusion of plutonium into Mg-Zr alloy cans and the solutions which have been found; (e) the mechanism of rupture of the lugs positioning the fuel elements in their graphite sleeves and the slight modification in the design of the fuel element which prevents fracture. The conclusion emphasizes the advantage obtained by the operator from the multiple links established between himself, the fuel design team, the manufacturers, and the test services. (author)

  7. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

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

    International Nuclear Information System (INIS)

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

  10. Effects of pin bowing in the CAGR fuel element

    International Nuclear Information System (INIS)

    A theoretical and experimental investigation of the effects of bowing on pin temperatures in CAGR fuel elements is described. A subchannel code, SCANDAL, has been developed to calculate the effects of bow in arbitrary rod clusters with single phase coolant. The fundamental assumptions of the code and the extra components needed to handle pin bowing are presented. In order to validate SCANDAL a heat transfer experiment has been performed, in which selected pins in a 36 pin CAGR fuel element have been mechanically bowed and detailed temperature effects measured. Results from this experiment are presented and compared with SCANDAL predictions. (author)

  11. Testing the surface contamination resuspension of a fuel element

    International Nuclear Information System (INIS)

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

  12. 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

  13. Dart model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

    International Nuclear Information System (INIS)

    The Dispersion Analysis Research Tool (DART) contains models for fission-gas induced fuel swelling, interaction of fuel with the matrix aluminum, resultant reaction-product swelling, and calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U3SiAl-Al and U3Si2-Al for various dispersion fuel element designs with the data. DART results are compared with data for fuel swelling Of U3SiAl-Al in plate, tube, and rod configurations as a function of fission density. Plate and tube calculations were performed at a constant fuel temperature of 373 K and 518 K, respectively. An irradiation temperature of 518 K results in a calculated aluminide layer thickness for the Russian tube that is in the center of the measured range (16 μm). Rod calculations were performed with a temperature gradient across the rod characterized by surface and central temperatures of 373 K and 423 K, respectively. The effective yield stress of irradiated Al matrix material and the aluminide was determined by comparing the results of DART calculations with postirradiation immersion volume measurement of U3SiAl plates. The values for the effective yield stress were used in all subsequent simulations. The lower calculated fuel swelling in the rod-type element is due to an assumed biaxial stress state. Fuel swelling in plates results in plate thickness increase only. Likewise, in tubes, only the wall thickness increases. Irradiation experiments have shown that plate-type dispersion fuel elements can develop blisters or pillows at high U-235 burnup when fuel compounds exhibiting breakaway swelling are used at moderate to high fuel volume fractions. DART-calculated interaction layer thickness and fuel swelling follows the trends of the observations. 3 refs., 2 figs

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

    International Nuclear Information System (INIS)

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

  15. 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)

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

    Science.gov (United States)

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

    2013-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

  17. Review of fuel element development for nuclear rocket engines

    International Nuclear Information System (INIS)

    The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program involving uranium-loaded graphite fuels, hydrogen propellant, and a target exhaust temperature of approximately 25000C. A very extensive uranium-loaded graphite fuel element technology evolved from the program. Selection and composition of raw materials for the extrusion mix had to be coupled with heat treatment studies to give optimum element properties. The highly enriched uranium in the element was incorporated as UO2, pyrocarbon-coated UC2, or solid solution UC . ZrC particles. An extensive development program resulted in successful NbC or ZrC coatings on elements to withstand hydrogen corrosion at elevated temperatures. Hot gas, thermal shock, thermal stress, and NDT evaluation procedures were developed to monitor progress in preparation of elements with optimum properties. Final evaluation was made in reactor tests at NRDS. Aerojet-General, Westinghouse Astronuclear Laboratory, and the Oak Ridge Y-12 Plant of Union Carbide Nuclear Company entered the program in the early 1960's, and their activities paralleled those of LASL in fuel element development. (U.S.)

  18. 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)

  19. 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)

  20. 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

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

    Science.gov (United States)

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

    1962-01-23

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

  2. Fuel element transport container with a removable cover

    International Nuclear Information System (INIS)

    The cover of the fuel element transport container is removably fixed with screws on a flange as mechanical loads have to be expected during the transfer to the disposal plant. A ring-shaped or star-shaped clamping device grips over the cover. It has a clamp claw to lock the cover and permits unscrewing without unlocking the cover. (DG)

  3. 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

  4. 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.

  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. 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)

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

    Science.gov (United States)

    Moran, Robert P.; Emrich, William

    2011-01-01

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

  8. 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

  9. 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.)

  10. Marangoni convection in fuel elements with liquid metal sublayer

    International Nuclear Information System (INIS)

    Analysis of heat- and mass-transfer in liquid metal sublayer of fuel element in the presence of gas bubbles is conducted. Analysis of the effects related with developing Marangoni convection is done. Assessed values are present for liquid metal flow velocities, temperature nonuniformity on inner side of fuel element cladding and in fuel pellets depending on gap size, physical properties of liquid metal in the gap, on heat generation rate and on average temperature in liquid-metal sublayer. It is shown that Marangoni convection can lead to fast corrosion on inner surface of the cladding. It is pointed out that at high values of convection rate the mechanism of material erosion also can be initiated

  11. Extraction process of fission products from spent nuclear fuel elements

    International Nuclear Information System (INIS)

    Process for extracting fission products contained in irradiated nuclear fuel elements consisting in bringing these elements into contact with water after having treated them mechanically to remove their cladding and/or cut them up, then separate these treated elements from the aqueous solution and recuperating at least one of the fission products concerned from this by concentrating it by distillation so as to obtain a concentrate containing these fission products and then processing this concentrate in order to ensure a long term storage of these fission products

  12. 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)

  13. 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.

  14. 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)

  15. 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)

  16. Leaching of actinide elements from simulated fuel debris into seawater

    International Nuclear Information System (INIS)

    For the prediction of the leaching behavior of actinide elements contained in the fuel debris that has arisen from the severe accident in Fukushima Daiichi Nuclear Power Station (NPS), a simulated fuel debris consisting of UO2 - ZrO2 solid solution doped with 137Cs, 237Np, 236Pu and 241Am tracers was synthesized, and agitated leaching tests were conducted for the simulated fuel debris in seawater. The synthesized simulated fuel debris was immersed and shaken in natural seawater collected at a coast 11 km away from Fukushima Daiichi NPS. The brief leaching test conditions were T = 25°C and solid-liquid ratio = 4 g/l, and the test duration was up to 31 days. The ratio of tracers leached into seawater from the simulated fuel debris by the agitated leaching test for 4 days was evaluated to be 0.09% for U, 0.01% for Np, 0.01% for Pu, 0.01% for Am and 35.39% for Cs by the α or γ spectrometry of the soluble fraction. The leaching of actinides from the real fuel debris in reactor units 1 - 3 in Fukushima Daiichi NPS is expected to be suppressed in comparison with that from normal light water reactor spent fuel. (author)

  17. Fuel-element inspection stand in the cooling pond of an atomic power plant

    International Nuclear Information System (INIS)

    A fuel-element inspection stand has been built in the cooling pool of the second power unit at the Ignalina Atomic Power Plant for the purpose of monitoring fuel elements unloaded from the reactor and for performing research involving the acquisition and analysis of statistically significant information concerning the reliability and efficiency of fuel elements and fuel bundles. The uses and specifications of the fuel-element inspection stand are given in this paper. 1 ref., 4 figs

  18. On-line elemental analysis of fossil fuel process streams by inductively coupled plasma spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Chisholm, W.P.

    1995-06-01

    METC is continuing development of a real-time, multi-element plasma based spectrometer system for application to high temperature and high pressure fossil fuel process streams. Two versions are under consideration for development. One is an Inductively Coupled Plasma system that has been described previously, and the other is a high power microwave system. The ICP torch operates on a mixture of argon and helium with a conventional annular swirl flow plasma gas, no auxiliary gas, and a conventional sample stream injection through the base of the plasma plume. A new, demountable torch design comprising three ceramic sections allows bolts passing the length of the torch to compress a double O-ring seal. This improves the reliability of the torch. The microwave system will use the same data acquisition and reduction components as the ICP system; only the plasma source itself is different. It will operate with a 750-Watt, 2.45 gigahertz microwave generator. The plasma discharge will be contained within a narrow quartz tube one quarter wavelength from a shorted waveguide termination. The plasma source will be observed via fiber optics and a battery of computer controlled monochromators. To extract more information from the raw spectral data, a neural net computer program is being developed. This program will calculate analyte concentrations from data that includes analyte and interferant spectral emission intensity. Matrix effects and spectral overlaps can be treated more effectively by this method than by conventional spectral analysis.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sund, R.E.; Strong, D.E.; Engholm, B.A.

    1977-02-01

    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 cm/sup 3/ 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.

  20. 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)

  1. Fuel Element Mechanical Design for CAREM-25 Reactor

    International Nuclear Information System (INIS)

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

  2. 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

  3. The behaviour of spherical HTR fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Hypothetical accidents may lead to significantly higher temperatures in HTR fuel than during normal operation. In order to obtain meaningful statements on fission product behaviour and release, irradiated spherical fuel elements containing a large number of coated particles (20,000-40,000) with burnups between 6 and 16% FIMA were heated at temperatures between 1400 and 2500 deg. C. HTI-pyrocarbon coating retains the gaseous fission products (e.g. Kr) very well up to about 2400 deg. C if the burnup does not exceed the specified value for THTR (11.5%). Cs diffuses through the pyrocarbon significantly faster than Kr and the diffusion is enhanced at higher fuel burnups because of irradiation induced kernel microstructure changes. Below about 1800 deg. C the Cs release rate is controlled by diffusion in the fuel kernel; above this temperature the diffusion in the pyrocarbon coating is the controlling parameter. An additional SiC coating interlayer (TRISO) ensures Cs retention up to 1600 deg. C. However, the release obtained in the examined fuel elements was only by a factor of three lower than through the HTI pyrocarbon. Solid fission products added to UO2-TRISO particles to simulate high burnup behave in various ways and migrate to attack the SiC coating. Pd migrates fastest and changes the SiC microstructure making it permeable

  4. Radial heat conduction in a power reactor fuel element

    International Nuclear Information System (INIS)

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

  5. Compaction of spent fuel elements from light water reactors

    International Nuclear Information System (INIS)

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

  6. 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

  7. 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)

  8. Three Dimensional Finite Element Modelling of a CANDU Fuel Pin Using the ANSYS Finite Element Package

    International Nuclear Information System (INIS)

    The ANSYS finite element modelling package has been used to construct a three-dimensional, thermomechanical model of a CANDU fuel pin. The model includes individual UO2 pellets with end dishes and chamfers, and a Zircaloy-4 fuel cladding with end caps. Twenty node brick elements are used with both mechanical and thermal degrees of freedom, allowing for a full coupling between the thermal and mechanical solutions under both steady state and transient conditions. Each fuel pellet is modelled as a separate entity that interacts both thermally and mechanically with the cladding and other pellets via contact elements. The heat transfer between the pellets and cladding is dependent on both the interface pressure and temperature, and all material properties of both the pellets and the sheath are temperature dependant. Spatially and temporally varying boundary conditions for heat generation and convective cooling can be readily applied to the model. The model naturally exhibits phenomena such as pellet hour glassing and ridging of the cladding at the Pellet to pellet interfaces, allowing for the prediction of localized sheath stresses. The model also allows for the prediction of fuel pin bowing due to asymmetric thermal loads and fuel pin sagging due to overheating of the cladding, which may occur under accident conditions. (author)

  9. 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

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

    International Nuclear Information System (INIS)

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

  11. Decommissioning of the HOBEG fuel element fabrication plant in Hanau

    International Nuclear Information System (INIS)

    The HOBEG fuel element fabrication plant was operated to manufacture graphite fuel elements for the thorium/high-temperature reactor in Hamm/Westf., Germany. The site comprises a 6000-m2 fenced area, an office/laboratory unit, and the production unit. In 1989, Nukem applied for a license to shut down the HOBEG fabrication plant in compliance with German atomic law (ATG article 7.3) and the radiation protection code with the goal of using the site and buildings for any other nonradioactive purpose. Approval for decommissioning was received in April 1995. Meanwhile, the existing equipment is being dismantled on the basis of single planning permissions and release for further use, for remelting, or for intermediate storage

  12. Advanced nuclear fuel cycle. Optimization by recycling instructive elements

    International Nuclear Information System (INIS)

    Rare-metals and rare-earths produced by fission reaction of uranium 235 in nuclear reactors and consequently contained in spent fuels are considered as potential resources for strategic material in many fields of recent industry. The report consists of several contributed papers concerning with possible utility of such fission products as ruthenium, rhodium, palladium, technetium, and neodymium, and with their recovery and separation from spent fuels as well as possible utilization of actinides and long-lived radioactive elements as radiation sources. To conclude, the present report proposes a new national strategy study to reorient the present scheme of reprocessing of spent fuels and radioactive waste disposal from a new perspective. (S. Ohno)

  13. 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

  14. Electoral structure of building foundations in nuclear fuel element plant

    International Nuclear Information System (INIS)

    Plant structures of nuclear fuel elements have a substantial burden. This requires analysis of the selection of the proper foundation for building support for a variety of different soil conditions found in two locations, first at a location near the nuclear power plant in Jepara and the second location BATAN Serpong area. Expected to know the location of soil conditions, we can determined the type of foundation that will be used based on the criteria requirements of the building. (author)

  15. Hydraulic Design Criteria for Spacer Grids of Nuclear Fuel Element

    International Nuclear Information System (INIS)

    In this paper a hydraulic model for calculating the pressure drop on the CARA spacer grids is extended.This model is validated and feedback from experimental hydraulic test performed in a low pressure loop.The importance of the spacer grid geometric parameter (that is, its thickness and length, the number and kind of their fix spacer), developing hydraulic design criteria for spacer grid on fuel element

  16. Improving the useful life of a 37-element fuel bundle

    International Nuclear Information System (INIS)

    Preliminary results indicate that CANDU burnup using 37-element fuel bundle with a slight enrichment can improve the useful life in the core. A slight enrichment in this study is increasing U-235 from 0.72 to 0.9 mass percent. A parametric study on criticality using Atomic Energy of Canada Limited’s WIMSAECL 3.1 and the Monte Carlo code, MCNP 5, developed by Los Alamos National Laboratory, is presented in this paper. (author)

  17. 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)

  18. Dry storage of spent fuel elements: interim facility

    International Nuclear Information System (INIS)

    Apart from the existing facilities to storage nuclear fuel elements at Argentina's nuclear power stations, a new interim storage facility has been planned and projected by the Argentinean Atomic Energy Commission (CNEA) that will be constructed by private group. This article presents the developments and describes the activities undertaken until the national policy approach to the final decision for the most suitable alternative to be adopted. (B.C.A.). 09 refs, 01 fig, 09 tabs

  19. Design verification testing for fuel element type CAREM

    International Nuclear Information System (INIS)

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

  20. 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)

  1. 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)

  2. Fuel element bundle shears with dust extraction when cutting

    International Nuclear Information System (INIS)

    To prevent deposits of dust when cutting in this very inaccessible area of the fuel element bundle shears, a grating is fitted, which is connected via extraction devices (a collecting funnel and extraction duct) to the downward shaft carrying flushing air for the pipe pieces cut off. The measures taken make it possible to remove dust during cutting by the joint action of flushing air and gravity. (orig./HP)

  3. Transport wagon for a fuel element transport container

    International Nuclear Information System (INIS)

    The transport containers are moved in the disposal plant with transport wagons on rails. The wagon consists of shielding walls, that surround the container for spent fuel elements of LWR at certain distances. The side walls can be moved as sliding doors. One of the end walls in connected with the driver cabine that contains the control equipment for the wagon. Through lead windows the inside space of the wagon can be observed from the cabine. (DG)

  4. Fabrication and Resintering of Annular UO2 Pellet

    International Nuclear Information System (INIS)

    Nuclear fuel is one of the most important components in a PWR affecting its safety and economy. The traditional PWR fuel pellet has a shape of cylindrical tablets of about 800 μm in diameter with a chamfer and dishes. A significant reduction in its failure rate has resulted from the improvements in fuel and cladding quality. Enhanced fuel assembly design allowed appreciable power density increases. However, it is difficult to achieve a significant increase of a power density under the current fuel pin design. Recently, Massachusetts Institute of Technology (MIT) has proposed an annular UO2 fuel with an internal cooling of each fuel rod. Annular fuel pellets with a voided central region have been used in VVER reactors without an internal cooling. Annular fuels with both internal and external cooling have been proposed for high temperature gas cooled reactors. However, commercial PWR reactors have not used such annular internally and externally cooled fuel rods, yet. There must be a lot of considerations in the various fields to introduce an annular internally and externally cooled fuel to commercial PWR reactors. The dimension tolerance and the thermal stability of a pellet are very important from the viewpoint of fabrication technology, because they have an influence on the size of the gap between the pellet and the inner/outer claddings. In this study, annular UO2 pellets with various densities were fabricated and then a resintering test was conducted. The changes of dimension and density of the sintered pellets were characterized

  5. 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)

  6. Selection of Isotopes and Elements for Fuel Cycle Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet

    2009-04-01

    Fuel cycle system analysis simulations examine how the selection among fuel cycle options for reactors, fuel, separation, and waste management impact uranium ore utilization, waste masses and volumes, radiotoxicity, heat to geologic repositories, isotope-dependent proliferation resistance measures, and so forth. Previously, such simulations have tended to track only a few actinide and fission product isotopes, those that have been identified as important to a few criteria from the standpoint of recycled material or waste, taken as a whole. After accounting for such isotopes, the residual mass is often characterized as “fission product other” or “actinide other”. However, detailed assessment of separation and waste management options now require identification of key isotopes and residual mass for Group 1A/2A elements (Rb, Cs, Sr, Ba), inert gases (Kr, Xe), halogens (Br, I), lanthanides, transition metals, transuranic (TRU), uranium, actinide decay products. The paper explains the rationale for a list of 81 isotopes and chemical elements to better support separation and waste management assessment in dynamic system analysis models such as Verifiable Fuel Cycle Simulation (VISION)

  7. On-site interim stores for spent fuel elements

    International Nuclear Information System (INIS)

    Since June 14 this year, the subject of a nuclear power consensus has been mentioned in the headlines less frequently than in past years. On that day, the government and operators of power plants agreed in Berlin on residual amounts of electricity to be produced and on management of the spent fuel elements of the nineteen German nuclear power plants. One sub-item under the heading of waste management, which continues to arouse debates not only at nuclear power plant sites despite the consensus reached, and which may become vitally important to the operation of plants, will be covered in more detail below: the construction of so-called decentralized interim stores. When present contracts with French and British firms on nuclear fuel reprocessing have been fulfilled and reprocessing has been phased out, these interim stores are to minimize the number of transports within Germany, a notorious source of general unrest, and are supposed to accommodate the spent fuel elements until a suitable repository will have been built where they can then be stored permanently. The whole development of a management concept for spent nuclear fuel in the Federal Republic of Germany, and the requirements to be met by decentralized interim stores, are explained in the article. The resultant standardized concept of dry interim cask storage is outlined in the light of its legal and technical criteria. Finally, the site-dependent variants of this concept are presented, and the status and the special features of the ongoing licensing procedures are explained. (orig.)

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

    International Nuclear Information System (INIS)

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

  9. The fabrication of nuclear fuel elements in Mexico

    International Nuclear Information System (INIS)

    The situation of the nucleoelectrical generation in Mexico by 1976 is described: two nuclear reactors under construction but no defined program on the type and start-up dates for the next power plants. However the existence of a general plan on nuclear power plants is mentioned, which, according to the last estimates reaches to 10,000 MW installed by 1990. The national intension, definitely expressed in the Law, is to supply domestic nuclear fuel to the power reactors operating in the country, starting with the first reload for the two BWR's at the first national station in Laguna Verde, which will be required at the end of 1981 and of 1982, respectively. Before such circumstances and the relatively short amounts of fuel elements that should be produced for those two unique reactors, Mexico already has to adopt a strategy to follow in respect to fuel elements fabrication. The two main options are analyzed: 1. To delay the local fabrication until a National Nuclear Program may be defined, meanwhile purchasing abroad the necessary reloads and initial cores; and 2. To start as soon as possible the local fuel elements fabrication in order to supply fuel for the first reload of the first unit of Laguna Verde, confronting the economical risks of such posture with the advantages of an immediate action. Both options are analyzed in detail comparing them specially under the economic point of view, standing out immediately the big effect of some factors which are economically imponderable, as experience and independance that would be gained with the second option. Emphasis is made on the advantages and risks of any case. According to the first option and once a National Program is defined, the work would be heavy but of simple strategy. On the contrary, the second option requires the adoption of a more complicated strategy, as either the project of the factory as its initial operation should be made under transient conditions, in view of the expected future expansion still

  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. Improvements in the fabrication of HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Braehler, Georg, E-mail: georg.braehler@nukemtechnologies.de [NUKEM Technologies GmbH, Industriestrasse 13, 63755 Alzenau (Germany); Hartung, Markus [NUKEM Technologies GmbH, Industriestrasse 13, 63755 Alzenau (Germany); Fachinger, Johannes; Grosse, Karl-Heinz [FNAG Furnaces Nuclear Applications Grenoble S.A.S., Wilhelm-Rohn Strasse 35, 63450 Hanau (Germany); Seemann, Richard [ALD Vacuum Technologies GmbH, Wilhelm-Rohn Strasse 35, 63450 Hanau (Germany)

    2012-10-15

    The application of High Temperature Reactor (HTR) Technology in the course of the continuously increasing world wide demand on energy is taken more and more under serious consideration in the power supply strategy of various countries. Especially for the emerging nations the HTR Technology has become of special interest because of its inherent safety feature and due to the alternative possibilities of applications, e.g. in the production of liquid hydrocarbons or the alternative application in H{sub 2} generation. The HTR fuel in its various forms (spheres or prismatic fuel blocks) is based on small fuel kernels of about 500 {mu}m in diameter. Each of these uranium oxide or carbide kernels are coated with several layers of pyrocarbon (PyC) as well as an additional silicon carbide (SiC) layer. While the inner pyrocarbon layer is porous and capable to absorb gaseous fission products, the dense outer PyC layer forms the barrier against fission product release. The SiC layer improves the mechanical strengths of this barrier and considerably increases the retention capacity for solid fission products that tent to diffuse at these temperatures. Especially the high quality German LEU TRISO spherical fuel based on the NUKEM design, has demonstrated the best fission product release rate, particular at high temperatures. The {approx}10% enriched uranium triple-coated particles are embedded in a moulded graphite sphere. A fuel sphere consists of approximately 9 g of uranium (some 15,000 particles) and has a diameter of 60 mm. As the unique safety features, especially the inherent safety of the HTR is based on the fuel design, this paper shall reflect the complexity but also developments and economical aspects of the fabrication processes for HTR fuel elements.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Determination of heterogeneous medium parameters by single fuel element method

    International Nuclear Information System (INIS)

    The neutron pulse propagation technique was employed to study an heterogeneous system consisting of a single fuel element placed at the symmetry axis of a large cylindrical D2O tank. The response of system for the pulse propagation technique is related to the inverse complex relaxation length of the neutron waves also known as the system dispersion law ρ (ω). Experimental values of ρ (ω) were compared with the ones derived from Fermi age - Diffusion theory. The main purpose of the experiment was to obtain the Feinberg-Galanin thermal constant (γ), which is the logaritmic derivative of the neutron flux at the fuel-moderator interface and a such a main input data for heterogeneous reactor theory calculations. The γ thermal constant was determined as the number giving the best agreement between the theoretical and experimental values of ρ (ω). The simultaneous determination of two among four parameters η,ρ,τ and Ls is possible through the intersection of dispersion laws of the pure moderator system and the fuel moderator system. The parameters τ and η were termined by this method. It was shown that the thermal constant γ and the product η ρ can be computed from the real and imaginary parts of the fuel-moderator dispersion law. The results for this evaluation scheme showns a not stable behavior of γ as a function of frequency, a result not foreseen by the theoretical model. (Author)

  19. The properties of spherical fuel elements and its behavior in the modular HTR

    International Nuclear Information System (INIS)

    The reference fuel element for all future HTR applications in the Federal Republic of Germany as developed by NUKEM/HOBEG in the framework of the 'High temperature Fuel-Cycle Project' had to be scrutinised for its compatibility with all the other design principles of the modular HTR, or possibly for restrictions forced upon reactor layout. This reference fuel element can be characterized by the following features: moulded spherical fuel element of 60 mm in diameter with fuel free shell of 5 mm thickness, based on carbon matrix; low enriched uranium (U/Pu fuel cycle); UO2 fuel kernels; TRISO coating (pyrocarbon and additional SiC layers)

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

    International Nuclear Information System (INIS)

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

  1. The modeling experience of fuel element units operation under MSC.MARC and MENTAT 2008R1

    International Nuclear Information System (INIS)

    MSC Software is leading developer of CAE-software in the world, so behaviour of fuel elements modeling with MSC.MARC use is of great practical importance. Behaviour of fuel elements usually is modeled in the elastic-viscous-plastic statement with account on fuel swelling during irradiation. For container type fuel elements contact interaction between fuel pellets and cladding or other parts of fuel element in top and bottom plugs must be in account. Results of simulated behaviour of various type fuel elements - container type fuel elements for PWR and RBMK reactors, dispersion type fuel elements for research reactors are presented. (authors)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  4. Actual Status of CAREM-25 Fuel Element Development

    International Nuclear Information System (INIS)

    In the frame of the CAREM Project, under Cnea s Reactor and Nuclear Plants Program, the Nuclear Fuel Thematic Area is one among others on which the project is organized. In this area, the primary objective to reach is to actualize the mechanical fuel element and reactivity control designs, taking in account the recents conceptual and engineering modifications introduced in the reactor, and ending with a consolidated conceptual and basic development.In order to reach these objectives, it is presented the way on which the area was organized, the participating working groups, the task required, the personnel involucrated, the grade of global development reached in the areas of engineering, developments, fabrication and essays of design verification, and the found difficulties, the tasks under ejecution, just finished and necessaries to fulfill completely the objectives. Finally, it is possible to say that due to the work realized, the conceptual design of both components is finished and the basic design is under development

  5. Surface coating Zr or Zr alloy nuclear fuel elements

    International Nuclear Information System (INIS)

    A method is disclosed for preventing stress corrosion cracking or metal embrittlement of a zirconium or zirconium alloy container that is to be coated on the inside surface with a layer of a metal such as copper, a copper alloy, nickel, or iron and used for holding nuclear fuel material as a nuclear fuel element. The zirconium material is etched in an etchant solution, desmutted mechanically or ultrasonically, oxidized to form an oxide coating on the zirconium, cleaned in an aqueous alkaline cleaning solution, activated for electroless deposition of a metal layer and contacted with an electroless metal plating solution. This method provides a boundary layer of zirconium oxide between the zirconium container and the metal layer. (author)

  6. Analysis of the operational reliability of VVER-1000 fuel elements and bundles in a three-year fuel cycle

    International Nuclear Information System (INIS)

    At the Novo-Voronezh Nuclear Power Plant, the fifth VVER-1000 unit, which was operated at nominal power from February 1980, completed nine fuel cycles in July 1990. The first unit of the Kalinin Nuclear Power Plant has operated from April 1984; in October 1990 the sixth fuel loading was completed. To data these power units are operating in steady-state in three-year fuel cycles (from June 1986 and from September 1989, respectively). By the end of 1988, operational experience had been accumulated on 1407 fuel element bundles on the third to the sixth fuel loading at Kalinin and the fifth to the ninth at Novo-Voronezh, which are in the transient and steady-state regimes of a three-year cycle. Of the 561 fuel element bundles monitored for gamma radiation, 14 were designated as leaking, which was 2.5% of the total bundles or 0.008% of the total number of fuel elements. Thus, a high degree of reliability was attained with enriched fuel elements. Here the authors analyze the reliability of fuel element bundles in taking the VVER-1000s to a three-year fuel cycle, and also generalize and systematize information on the fundamental characteristics of a group of fuel element bundles in going to to steady-state conditions of the three-year fuel cycle

  7. 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)

  8. 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)

  9. 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.)

  10. 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)

  11. A sipping system for irradiated fuel elements diagnostic

    International Nuclear Information System (INIS)

    This paper presents a wet-sipping system for quick diagnosis of irradiated fuel elements. Gamma detection is equated with respect to counting rate and system global efficiency. The theoretical results are verified by an experimental simulation employing NaI (Tl) and HPGe detectors in a I-131 activated water tank. The tank volume is parameterized for dimensioning to reach a minimum sensitivity of 600 Bq/l for gamma rays of 364 keV. It is found that the HPGe is the detector type which best suits the resolution requirement for a sipping system. (author)

  12. Graphitic matrix materials for spherical HTR fuel elements

    International Nuclear Information System (INIS)

    The report comprises the graphical documentation of irradiation results on graphitic matrix materials for spherical HTR fuel elements. The plotted results are based on data analyses of the series of exposures in the High Flux Reactor Petten (HFR). The documentation includes information about the changes of - the dimensions - the dynamic modulus of elasticity - the coefficient of thermal expansion of the materials after irradiation with fast neutrons. The irradiation experiments and the data analyses are part of the matrix development and irradiation programme, whose objective, realization and results obtained are summarized. (orig./IHOE)

  13. Fuel elements assembling for the DON project exponential experience

    International Nuclear Information System (INIS)

    It is described the fuel unit used in the DON exponential experience, the manufacturing installments and tools as well as the stages in the fabrication.These 74 elements contain each 19 cartridges loaded with synterized urania, uranium carbide and indium, gold, and manganese probes. They were arranged in calandria-like tubes and the process-tube. This last one containing a cooling liquid simulating the reactor organic. Besides being used in the DON reactor exponential experience they were used in critic essays by the substitution method in the French reactor AQUILON II. (Author) 6 refs

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

    International Nuclear Information System (INIS)

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

  15. Experiments of replacement of a single fuel element. Interpretation method

    International Nuclear Information System (INIS)

    An original method of measurement of effective cross sections of fissile materials has been developed by the CEA. According to it, the central fuel element of an experimental critical reactor is replaced by a sample containing the material to be studied. This report proposes a method of comprehensive interpretation of these experiments of replacement of a single element. A first part presents the method principle (problem definition, study of the propagation of neutron density disturbance in a critical multiplier medium) and the notion of equivalent sample. The second part reports the study of the disturbed area, and the third part the Uranium-235 and Boron calibration of the reactor (approximation order in disturbance theory, interpretation of calibration measurements by the heterogeneous method)

  16. The beginning of the LEU fuel elements manufacturing in the Chilean Commission of Nuclear Energy

    International Nuclear Information System (INIS)

    The U3 Si2 LEU fuel fabrication program at CCHEN has started with the assembly of four leaders fuel elements for the RECH-1 reactor. This activity has involved a stage of fuel plates qualification, to evaluate fabrication procedures and quality controls and quality assurance. The qualification extent was 50% of the fuel plates, equivalent to the number of plates required for the assembly of two fuel elements. (author)

  17. 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

  18. 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.

  19. 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

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

    International Nuclear Information System (INIS)

    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

  1. 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

  2. Annular core for modular high temperature gas-cooled reactor (MHTGR)

    International Nuclear Information System (INIS)

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40 % greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93 m tall. Fuel elements contain TRISO-coated microspheres of 19.8 % enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above. (author)

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

    International Nuclear Information System (INIS)

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

  4. 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.)

  5. 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)

  6. 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

  7. 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

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

    International Nuclear Information System (INIS)

    Calculations of fuel swelling of U3SiAl-Al and U3Si2 were performed for various dispersion fuel element designs. Breakaway swelling criteria in the form of critical fuel volume fractions were derived with data obtained from U3SiAl-Al plate irradiations. The results of the analysis show that rod-type elements remain well below the pillowing threshold. However, tubular fuel elements, which behave essentially like plates, will likely develop pillows or blisters at around 90% 235U burnup. The U3Si2-Al compounds demonstrate stable swelling behavior throughout the entire burnup range for all fuel element designs

  9. 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

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

    International Nuclear Information System (INIS)

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

  11. A mechanistic code for intact and defective nuclear fuel element performance

    Science.gov (United States)

    Shaheen, Khaled

    During reactor operation, nuclear fuel elements experience an environment featuring high radiation, temperature, and pressure. Predicting in-reactor performance of nuclear fuel elements constitutes a complex multi-physics problem, one that requires numerical codes to be solved. Fuel element performance codes have been developed for different reactor and fuel designs. Most of these codes simulate fuel elements using one-or quasi-two-dimensional geometries, and some codes are only applicable to steady state but not transient behaviour and vice versa. Moreover, while many conceptual and empirical separate-effects models exist for defective fuel behaviour, wherein the sheath is breached allowing coolant ingress and fission gas escape, there have been few attempts to predict defective fuel behaviour in the context of a mechanistic fuel performance code. Therefore, a mechanistic fuel performance code, called FORCE (Fuel Operational peRformance Computations in an Element) is proposed for the time-dependent behaviour of intact and defective CANDU nuclear fuel elements. The code, which is implemented in the COMSOL Multiphysics commercial software package, simulates the fuel, sheath, and fuel-to-sheath gap in a radial-axial geometry. For intact fuel performance, the code couples models for heat transport, fission gas production and diffusion, and structural deformation of the fuel and sheath. The code is extended to defective fuel performance by integrating an adapted version of a previously developed fuel oxidation model, and a model for the release of radioactive fission product gases from the fuel to the coolant. The FORCE code has been verified against the ELESTRES-IST and ELESIM industrial code for its predictions of intact fuel performance. For defective fuel behaviour, the code has been validated against coulometric titration data for oxygen-to-metal ratio in defective fuel elements from commercial reactors, while also being compared to a conceptual oxidation model

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

    International Nuclear Information System (INIS)

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

  13. Annular pancreas (image)

    Science.gov (United States)

    Annular pancreas is an abnormal ring or collar of pancreatic tissue that encircles the duodenum (the part of the ... intestine that connects to stomach). This portion of pancreas can constrict the duodenum and block or impair ...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Simulation of steam ingress accidents with irradiated fuel elements

    International Nuclear Information System (INIS)

    Accident sequences are considered for the gas-cooled High Temperature Reactor (HTR), in which water may enter into the primary circuit and reactor core as a consequence of pipe rupture in the steam generator. Irradiation experiments with intermittent water injections have demonstrated that moisture in the sweep gas lead to an increase of the release of fission gases and iodine from defective/failed particles. A special apparatus KORA was constructed in the Hot Cells of the Research Centre Juelich to study the effects of moisture-related fission product release as a function of temperature and water vapour partial pressure with different fuel samples. Initial experiments with irradiated UO2 and UCO fuels at 800 deg. C showed an increased of 85Kr release with water vapour additions. In contrast, intact particles are not affected even by extremely long water vapour injections. UO2 kernels obtained by cracking particles from spherical fuel elements correspond to irradiation-induced failures; they show the following release fractions at 800 deg. C after repeated injections of water vapour: with a medium burn-up of 5% FIMA; with a high burn-up of 9% FIMA; release of 0.4 to 2.6% of the 85Kr inventory; release of 17% of the 85Kr inventory. In the case of defective UO2 TRISO particles, which would dominate the release in an HTR-MODUL, some of the free fuel may have been carburized in the fabrication process during the final heat treatment at 1950 deg. C, which could lead to changed release behaviour. Further studies will have to show whether the release as a consequence of the influence of water vapour is similar to that from UO2 kernels or possible higher. There was a complete moisture-induced release from high-burnup UCO kernels or designed-to-fail particles with a burnup of 20% FIMA. Together with the knowledge that unirradiated UO2 kernels show practically no changes due to moisture, the moisture-induced fission gas release - and similar the iodine release - from fuel

  18. 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)

  19. 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

  20. Analysis of Ya-21u thermionic fuel elements

    International Nuclear Information System (INIS)

    The Ya-21u unit of the Soviet-made TOPAZ-II power system has recently been tested at the Thermionic Evaluation Facility in Albuquerque, New Mexico. A change in the unit performance was measured during these tests. In an attempt to identify the causes of this change performance, data were examined and used to estimate surface properties of electrodes of thermionic fuel elements (TFEs) of the power system. The effective emissivity was estimated at ∼0.03 to 0.035 higher than for as-fabricated TFE and cesiated work functions of the electrodes, which were higher than for as-fabricated TFEs. These changes in the effective emissivity and cesiated work functions, caused by gaseous impurities and air incursion in the TFEs interelectrode gap, lowered both the emitter temperature and the output load voltage thus contributing to the measured decrease in output power

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

    International Nuclear Information System (INIS)

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

  2. 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

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

    International Nuclear Information System (INIS)

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

  4. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

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

  6. 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

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

    International Nuclear Information System (INIS)

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

  8. 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,.

  9. 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

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

    International Nuclear Information System (INIS)

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

  11. C A R A fuel element for Atucha nuclear power plants and development plan

    International Nuclear Information System (INIS)

    This paper presents the current state and the development plan of the C A R A fuel element.Main activities were carried out towards to welding of the end plates of the C A R A fuel element by a new process, and the assembling and hanging of the C A R A fuel element in its Atucha configuration, by using an external basket

  12. 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)

  13. Investigation on laser welding characteristics for appendage of bearing pads of nuclear fuel element

    International Nuclear Information System (INIS)

    In CANDU nuclear fuel manufacturing the brazing technology has been adopted conventionally to attach the bearing pads of nuclear fuel elements. However, in order to meet good performance of nuclear fuel and improved working efficiency, we started developing the laser welding technology for attachments of the bearing pads. Since the YAG laser can be suitable for small parts and transmit the beam through the optical fiber, the process is corresponding to mass-production with working shops. Making the most of this feature, we have developed the laser welding for appendage of the bearing pads of nuclear fuel elements, and has studied on the laser welding characterisitcs of appendages for nuclear fuel element

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

    International Nuclear Information System (INIS)

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

  15. 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.

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

    International Nuclear Information System (INIS)

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

  17. FUEL ASSAY REACTOR

    Science.gov (United States)

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

    1962-12-25

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

  18. 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)

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

    International Nuclear Information System (INIS)

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

  20. Fuel Element Designs for Achieving High Burnups in 220 MW(e) Indian PHWRs

    International Nuclear Information System (INIS)

    Presently 19-element natural uranium fuel bundles are used in 220 MW(e) Indian PHWRs. The core average design discharge burnup for these bundles is 7000 MW·d/Te U and maximum burnup for assembly goes upto of 15 000 MWD/Te U. Use of fuel materials like MOX, Thorium, slightly enriched uranium etc in place of natural uranium in 19-element fuel bundles, in 220 MW(e) PHWRs is being investigated to achieve higher burnups. The maximum burnup investigated with these bundles is 30 000 MW·d/Te U. In PHWR fuel elements no plenum space is available and the cladding is of collapsible type. Studies have been carried out for different fuel element target burnups with different alternative concepts. Modification in pellet shape and pellet parameters are considered. These studies for the PHWR fuel elements/assemblies have been elaborated in this paper. (author)

  1. Quantification of factors affecting thermally-induced bow in a CANDU fuel element simulator

    International Nuclear Information System (INIS)

    Thermally induced bow, caused by a circumferential temperature distribution around a fuel element, was investigated in this study using a fuel element simulator. The objective was to identify the factors affecting CANDU fuel element bow induced by dryout as a result of some predicted reactor transients in which the maximum fuel temperature reaches 600 deg C. The results showed that circumferential temperature distribution, pellet-to-sheath mechanical interaction and creep were the major factors affecting bow. Transient bow increased with increasing diametral sheath temperature difference and with mechanical interaction between the pellet and the sheath. Permanent bow of the fuel element was observed in some tests which was the result of creep. Mechanical interaction between the sheath and pellet produced the stresses necessary for creep deformation. A simplified ABAQUS model was developed to explain the experimental findings and could be used to predict the bow behaviour of fuel elements during reactor transients, where the dry patches are of different sizes. (author)

  2. Radiation resistance of pyrocarbon-boned fuel and absorbing elements for HTGR

    International Nuclear Information System (INIS)

    In choosing the reactor type, problems of nuclear and radiation safety are outstanding. The analysis of the design and experiments show that HTGR type reactors helium cooled satisfy all the safety requirements. It has been planned in the Soviet Union to construct two HTGR plants, VGR-50 and VG-400. Later it was decided to construct an experimental plant with a low power high temperature reactor (VGM). Spherical uranium-graphite fuel elements with coated fuel particles are supposed to be used in HTGR core. A unique technology for producing spherical pyrocarbon-bound fuel and absorbing elements of monolithic type has been developed. Extended tests were done to to investigate fuel elements behaviour: radiation resistance of coated fuel particles with different types of fuel; influence of the coated fuel particles design on gaseous fission products release; influence of non-sphericity on coated fuel particle performance; dependence of gaseous fission products release from fuel elements on the thickness of fuel-free cans; confining role of pyrocarbon as a factor capable of diminishing the rate of fission products release; radiation resistance of spherical fuel elements during burnup; radiation resistance of spherical absorbing elements to fast neutron fluence and boron burnup

  3. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. 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

  7. Behavior of mixed-oxide fuel elements during an overpower transient

    International Nuclear Information System (INIS)

    A slow-ramp (0.1%/s), extended overpower (∼90%) transient test was conducted in EBR-II on 19 mixed-oxide fuel elements with conservative, moderate, and aggressive designs. Claddings for the elements were Type 316, D9, or PNC-316 stainless steel. Before the transient, the elements were preirradiated under steady-state or steady-state plus duty-cycle (periodic 15% overpower transient) conditions to burnups of 2.5-9.7 at%. Cladding integrity during the transient test was maintained by all fuel elements except one, which had experienced substantial overtemperature in the earlier stedy-state irradiation. Extensive centerline fuel melting occurred in all test elements. Significantly, this melting did not cause any elements to breach, although it did have a strong effect on the other aspects of fuel element behavior. (orig.)

  8. 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)

  9. Plate-shaped high power nuclear fuel element containing low enrichment uranium and its preparation

    International Nuclear Information System (INIS)

    The present invention provides a plate-shaped high power nuclear fuel element containing low enrichment uranium (5 to 20 percent by weight uranium235 in the uranium component) as the fissionable material, the fuel element essentially comprising a plate of UAl4 provided with a sheath (clad) of aluminum or an aluminum alloy and impurities inherent to the manufacturing process. (DG)

  10. Evaluation of the fuel-element assembly non-hermeticity at its early stage

    International Nuclear Information System (INIS)

    The given paper deals with control of the fuel-element assembly shell state at the early stage of failure development. Technique for the fuel-element assembly shell state evaluation are described. A method for assembly failure detection, used at WWR of the Institute for Nuclear Research is described also

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

    International Nuclear Information System (INIS)

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

  12. Aerodynamic study of the fluid flow in the channel of a reactor filled with internally and externally cooled fuel elements

    International Nuclear Information System (INIS)

    A study is made of the problem of the flow-rate and pressure distributions along the length of two volumes, internal and external, bounded by a series of non-continuous annular elements placed along the channel axis. It is observed that the phenomenon can easily be represented by equations. The theoretical expressions observed are particularly simple when the distances between the elements are above a certain minimum value. The experimental work has made it possible to show that the theoretical formulation derived is valid with a very great accuracy. The experimental study has also been carried out in the case of a very small spacing between the elements. It has been possible to show in this case that the hypothesis made for deriving the theoretical expressions was perfectly justified. In the last part finally, we consider the practical problem of evaluating the pressure-drops between the ends of a series of annular elements. (author)

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

    International Nuclear Information System (INIS)

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

  14. Corrosion electrochemistry of fuel element materials in pond storage conditions

    International Nuclear Information System (INIS)

    The integrity of spent fuel element clad is imperative in avoiding undesirable release of soluble fission products into cooling ponds following discharge from reactor into aqueous storage. Corrosion is a possible route to clad penetration and pond chemistry is controlled to maintain passivity. In order to allow optimisation of operating chemistry and hence minimise corrosion, the aqueous corrosion behaviours of commercial magnesium alloy Magnox Al80 and natural uranium have been studied with the use of modern electrochemical techniques. Historical corrosion investigations have generally employed conventional methods such as weight loss and hydrogen evolution which are often insensitive to certain mechanisms and require significant time-scales. Here the use of electrochemical noise and impedance spectroscopy alongside DC polarisation has allowed the study of general and localized corrosion, notably clad pitting, with greatly enhanced sensitivity and reduced time-scales. Significant practical benefits are also afforded, for example, the radiological wastes arising are reduced by shorter measurement times. Uranium exhibited a simple corrosion behaviour whilst Magnox corrosion proved to be a more complex and dynamic system. The formation of a passive layer and potential to subsequent breakdown are dictated by a number of factors, including pH and anion concentration as well as the presence of a crevice or galvanic couple. This has demanded the development and application of novel techniques and real time observations of pitting initiation and propagation have been made. (authors)

  15. Implementation of Tempul Code on Calculation of Radial Temperature Distribution in TRIGA Fuel Element after Pulse

    International Nuclear Information System (INIS)

    TEMPUL is one dimensional computer code for calculating radial fuel temperature distribution in a fuel immediately after the pulse. Implementation of TEMPUL code was performed to calculate of radial temperature distribution on TRIGA fuel element. The gap between fuel element and cladding is treated to be in contact (without gap), gap is filled with air and gap is filled with helium gas, respectively. Equilateral triangular arrangement coolant channel is assumed. The calculated results on calculation of radial temperature distribution in TRIGA fuel element immediately after the pulse occur relatively high ascending tendency in zirconium rod (radius 0.3175 cm) and fuel element-cladding interface (radius 1.82245 cm) at the first second after pulse with no gap and gap filled with helium gas treatment. Rising of cladding and interface between cladding and coolant average temperature reach up to 500 oC drastically occur in the first second after the pulse. (author)

  16. Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

    Science.gov (United States)

    Stewart, Mark E.; Schnitzler, Bruce G.

    2015-01-01

    This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

  17. Metallurgy Sub-directorate, Departments of Technology, Department of Fuel Elements and Structures, Division of Study of Fuel Elements - 13. semi annual report on the 1 April 1969

    International Nuclear Information System (INIS)

    This document gathers several reports of studies performed on fuel elements in the different types of nuclear power stations: graphite-gas, heavy water-gas, light water, fast neutrons. For the graphite-gas reactors, the addressed topics are: Study of thermal cycling of fuel elements in a reactor and exploitation of the CEA 4 apparatus; Fatigue strength of devices for the centring of fuel elements in graphite-gas reactors; Overview of studies on Bugey 1 element, one year before serial production. For heavy water reactors, the studies addressed the EL4 Z2 clusters: description of the first 100 clusters in production, planned improvements and cost assessment. For light water reactors, the study concerned the irradiation preparation in BR3 of experimental fuel rods (UO2 in a Zircaloy 4 sheath). For fast neutrons reactors, the report proposes an overview of fast neutron reactors (blowhole study, study of the Phenix hexagonal sheath, the Phenix assembly mock-up, the Rapsodie reactor). A last part contains common studies related to UO2: assessment of various free volumes in UO2 fuel rods, evolution of pressure within a UO2/Zr-Cu fuel rod at the beginning of irradiation, and evolution of internal pressure in a fuel rod

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

    International Nuclear Information System (INIS)

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

  19. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. 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...

  3. Effect of annular secondary conductor in a linear electromagnetic stirrer

    Indian Academy of Sciences (India)

    R Madhavan; V Ramanarayanan

    2008-10-01

    This paper presents the variation of average axial force density in the annular secondary conductor of a linear electromagnetic stirrer. Different geometries of secondaries are considered for numerical and experimental validation namely, 1. hollow annular ring, 2. annular ring with a solid cylinder and 3. solid cylinder. Experimental and numerical simulations are performed for a 2-pole in house built 15 kW linear electromagnetic stirrer (EMS). It is observed for a supply current of 200 A at 30 Hz the force densities in the hollow annular ring is 67% higher than the equivalent solid cylinder. The same values are 33% for annular ring with a solid cylinder. Force density variation with supply frequency and current are also reported. Numerical simulations using finite element model are validated with experimental results.

  4. Generalized granuloma annulare

    Directory of Open Access Journals (Sweden)

    Khatri M

    1995-01-01

    Full Text Available A 35-years-old female patient had generalized pruritic papular lesions, distributed like dermatitis herpetiformis for last 4 years. Histopathologic changes were typical of granuloma annulare with negative results of direct immunofluorescence. The patient did not have association of diabetes mellitus or any other systemic disease. She failed to respond to dapsone therapy and 13-cis-retinoic acid.

  5. Oscillating annular liquid membranes

    International Nuclear Information System (INIS)

    The response of annular liquid membranes (e.g. used as protection systems in laser fusion reactors) to sinusoidal mass flow rate fluctuations at the nozzle exit is analyzed as a function of the amplitude and frequency of the axial velocity fluctuations at the nozzle exit and thermodynamic compression of the gas enclosed by the membrane. The pressure of the gases enclosed by the annular membrane and the axial distance at which the annular membrane merges on the symmetry axis are periodic functions of time which have the same period as that of the mass flow rate fluctuations at the nozzle exit. They are also nearly sinusoidal functions of time for small amplitudes of the mass flow rate fluctuations at the nozzle exit, and exhibit delay and lag times with respect to the sinusoidal axial velocity fluctuations at the nozzle exit. The delay and the lag times are functions of the amplitude and frequency of the mass flow rate fluctuations at the nozzle exit and the polytropic exponent. The amplitudes of both the pressure of the gases enclosed by the annular liquid membrane and the convergence length increase and decrease, resp., as the amplitude and frequency of the mass flow rate fluctuations at the nozzle exit, resp., are increased. They also increase as the polytropic exponent is increased. (orig.)

  6. Annular Planar Monopole Antennas

    OpenAIRE

    Chen, Z. N.; Ammann, Max; Chia, W.Y. W.; See, T.S. P.

    2002-01-01

    A type of annular planar monopole antenna is presented. The impedance and radiation characteristics of the monopole with different holes and feed gaps are experimentally examined. The measured results demonstrate that the proposed antenna is capable of providing significantly broad impedance bandwidth with acceptable radiation performance.

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

    International Nuclear Information System (INIS)

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

  8. Detonation Initiation by Annular Jets and Shock Waves

    OpenAIRE

    Shepherd, Joseph E.

    2005-01-01

    The objective of this research is to experimentally determine the feasibility of initiating detonation in fuel-air mixtures using only the energy in hot, compressed air. The existing 6-inch shock tube at Caltech was used to create hot, high-pressure air behind a reflected shock wave. The hot air created an imploding annular shock wave when it jetted through an annular orifice into a 76-mm-diameter, 1-m-long tube attached to the end of the shock tube. A special test section with an annular ...

  9. 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

  10. 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

  11. Drying results of K-Basin fuel element 5744U (Run 4)

    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 basins 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. 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 fourth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 5744U. This element (referred to as Element 5744U) was stored underwater in the K-West Basin from 1983 until 1996. Element 5744U 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

  12. Drying results of K-Basin fuel element 1164M (run 6)

    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 sixth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 1164 M. This element (referred to as Element 1164M) was stored underwater in the K-West Basin from 1983 until 1996. Element 1164M 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

  13. 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...

  14. Extraction process of fission products contained in irradiated nuclear fuel elements

    International Nuclear Information System (INIS)

    In the process described, the fission products contained in irradiated nuclear fuel elements are extracted before the fuel is dissolved by wet process. After the element have been mechanically removed from their cladding and/or sliced up, they are processed in water to cause the fission products to be dissolved in an aqueous solution, after which the processed elements are separated from the aqueous solution obtained and at least one of the fission products is retrieved from this aqueous solution

  15. 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)

  16. Development of TIG welding technique for endcap welding of PHWR MOX fuel elements

    International Nuclear Information System (INIS)

    Full text: Fabrication of PHWR fuel elements involves manufacture of fuel pellets, loading them in the zircaloy-4 clad tube and endcap welding of filled zircaloy-4 clad tube by resistance welding technique. This welding technique gives higher production rate but the welds are not amenable to non destructive techniques like radiography. The quality of the weld is assured by the destructive metallographic technique which is statistical in nature. DAE has recently decided to go ahead with plutonium recycling in PHWR in order to increase the burnup to around 10,000 MWD/Te to reduce the fuel cycle cost and reduce the requirement of uranium. The bundle design remains the same as it is being used in 235 MWe PHWR's in India. In the proposed nineteen element, fuel bundle the internal seven elements will contain MOX fuel pellets and external twelve elements will contain standard natural uranium dioxide pellets. MOX fuel elements will be fabricated at Advanced Fuel Fabrication Facility, BARC, Tarapur. It is proposed to make the fuel element by TIG welding technique which has the advantage of using radiography for the evaluation of the end plug weld. Further there will be no machining required over the weld bead which is a must for a resistance weld. However, the use of TIG technique requires change in endcap design and use of these endcaps leads to marginal (<1 %) decrease in stack length to maintain the same fuel element length as used in natural uranium dioxide PHWR, fuel bundle. This paper describes the development work carried out at AFFF on TIG welding of endplugs for PHWR fuel elements with the new plug design, optimisation of welding parameters and the results of the welding trials

  17. Analysis of burnt nuclear fuel elements by gamma-spectrometry

    International Nuclear Information System (INIS)

    Gamma-spectrometry allows a non-destructive determination of the fission and activation product content of spent nuclear fuel. The concentration of some of these products depends significantly on the so-called fuel parameters which describe the irradiation history and the fuel composition. The use of these dependences for deriving ''unknown fuel parameters'' from measured fission product activities is investigated in this work. Relevant application fields are burnup determination, fuel testing and inspections within the nuclear materials safeguards programme. The present thesis investigates how these dependences can be used to derive unknown fuel parameters. The possibilities and basic limitations of deriving information from a measured gamma spectrum are considered on principle. The main conclusion is that only ratios of fission product activities allow the development of an interpretation method which is generally applicable to all types of fuel from different reactors. The dependence of activity ratios on cooling time, irradiation time, integrated and final neutron flux, fuel composition, as well as fission and breeding rates are then investigated and presented graphically in a way suitable for applicaton. These relationships can be used for the analysis of spent fuel, and the detailed procedures, which depend on the applicaton field, are worked out in this work. In order to test the interpretation methods, samples of nuclear fuel have been irradiated and the gamma spectra analysed. The methods developed in this work can be applied successfully to the analysis of burnt fuel in the frame of fuel testing programmes and to safeguards verification purposes. If however, apart from a gamma spectrum, no information on the investigated fuel is available, the above-mentioned parameters can be derived with low accuracy only. (author)

  18. 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)

  19. Operating experience with the fuel element repair shop at Muelheim-Kaerlich nuclear power plant

    International Nuclear Information System (INIS)

    A dismountable fuel element repair shop has been installed at Muelheim-Kaerlich Nuclear Power Plant. This permits irradiated fuel elements to be repaired under water using the available repair equipment. The withdrawal of the defective fuel rod is effected at about 3 metres below the surface of the water using a fixed withdrawal tool by lowering the fuel element with an integrated holsting unit. This type of repair is possible using relatively short tools and with maximum visibility of the end of the fuel element and is independent of plant-related lifting equipment. The repair shop is designed in accordance with the appropriate nuclear technical plant regulations and has been accepted by the German Technical Supervisory Board. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-11-01

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

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

    International Nuclear Information System (INIS)

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

  2. 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.)

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

    International Nuclear Information System (INIS)

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

  4. Studies on the fission gas release behaviour of BWR and experimental MOX fuel elements

    International Nuclear Information System (INIS)

    Fission gas release data were generated on 13 fuel elements from the two boiling water reactors (BWRs) at the Tarapur Atomic Power Station (TAPS). The burn-up of these fuel elements varied from 3 000 to 24 000 MWd/t. The fuel elements were taken from fuel assemblies that were irradiated at different core locations in single and multiple irradiation cycles. A new fission gas measuring set-up was designed and fabricated to analyse fuel elements with low void volumes and low fission gas releases. Fifteen experimental mixed oxide (MOX) fuel pins were fabricated and irradiated in the pressurised water loop (PWL) of the CIRUS reactor to burn-ups ranging from 2 000 MWd/t to 16 000 MWd/t. The fission gas release from MOX fuels was predicted with the computer code PROFESS using the fuel fabrication and irradiation data. The results from the fission gas release measurements from some of the irradiated MOX fuel elements are compared with those predicted using the code. (author)

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

    International Nuclear Information System (INIS)

    Three clusters, each containing 6 zircaloy-2 clad short length fuel elements of either MOX or UO2 fuel pellets were fabricated for irradiation in pressurized water loop of CIRUS. The major objectives of the programme were: (a) to optimize the various fabrication parameters for developing a flow sheet for MOX fuel element fabrication; (b) to study the performance of the MOX fuel elements at a peak heat flux of 110 W/cm2; and (c) to study the effect of various fuel pellet design changes on the behaviour of the fuel element under irradiation. Two clusters, one each of UO2 and MOX, have been successfully irradiated to the required burn-up level and are now awaiting post irradiation examinations. The third MOX cluster is still undergoing irradiation. Fabrication of these fuel elements involved considerable amount of developing work related to the fabrication of the MOX fuel pellets and the element welding technique and is reported in detail in this report. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-12-15

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

  8. Nuclear fuel element and method for its fabrication

    International Nuclear Information System (INIS)

    Within a special gas-permeable container particles of the ternary alloy Zr, Ni and Ti are contained in the fuel assembly for the BWR or PWR. Position and shape of the ternary alloy allow to remove water, water vapor and reactive gases from the fuel assembly utilizing the getter properties of this alloy. Moreover, the alloy is arranged at the coldest position of the fuel assembly, any inverse reaction thus being prevented. (DG)

  9. Nondestructive examination of 54 fuel and reflector elements from Fort St. Vrain core segment 2

    International Nuclear Information System (INIS)

    Fifty-four fuel and reflector elements irradiated in core segment 2 of the Fort St. Vrain high-temperature gas-cooled reactor (HTGR) were nondestructively examined. The time- and volume-averaged graphite irradiation temperatures for the elements ranged from approx. 3500 to 7500C. The element-averaged fast neutron fluences ranged from approx. 0.2 to 1.6 x 1025 n/m2 (E > 29 fJ)/sub HTGR/. The elements, except for two fuel elements in which single localizeed cracks developed during irradiation, were in excellent condition. No evidence was observed of significant graphite oxidation or mechanical interaction beween elements. The cracks in the two elements did not affect their performance or handling. These elements were, otherwise, in excellent condition. Nearly all elements shrank in both the axial and radial directions, but the dimensional changes were relatively small

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

    International Nuclear Information System (INIS)

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

  11. 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)

  12. 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.)

  13. Preparation of spherical fuel elements for HTR-PM in INET

    International Nuclear Information System (INIS)

    Highlights: • Modifications and optimizations in the manufacture of spherical fuel elements (SFE) for HTR-PM are presented. • A newly developed overcoater exhibits good stability and high efficiency in the preparation of overcoated particles. • The optimized carbonization process reduces the process time from 70 h in the period of HTR-10 to 20 h. • Properties of the prepared SFE and matrix graphite balls meet the design specifications for HTR-PM. • In particular the mean free uranium fraction of 5 consecutive batches is only 8.7 × 10−6. -- Abstract: The spherical fuel elements were successfully manufactured in the period of HTR-10. In order to satisfy the mass production of fuel elements for HTR-PM, several measures have been taken in modifying and optimizing the manufacture process of fuel elements. The newly developed overcoater system and its corresponding parameters exhibited good stability and high efficiency in the preparation of overcoated particles. The optimized carbonization process could reduce the carbonization time from more than 70 h to 20 h and improve the manufacturing efficiency. Properties of the manufactured spherical fuel elements and matrix graphite balls met the design specifications for HTR-PM. The mean free uranium fraction of 5 consecutive batches was 8.7 × 10−6. The optimized fuel elements manufacturing process could meet the requirements of design specifications of spherical fuel elements for HTR-PM

  14. 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)

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

    International Nuclear Information System (INIS)

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

  16. Development of laser welding system for attachment of bearing pads of nuclear fuel element

    International Nuclear Information System (INIS)

    In CANDU nuclear fuel manufacturing the brazing technology has been adopted conventionally to attach the bearing pads of nuclear fuel elements. However, in order to meet good performance of nuclear fuel and improved working efficiency, we started developing the laser welding technology for attachments of the bearing pads. Since the YAG laser can be suitable for small parts and transmit the beam through the optical fiber, the process is corresponding to mass-production with working shops. Making the most of this feature, we have developed the laser welding system for attachment of the bearing pads on nuclear fuel elements, and has carried out basic welding experiments

  17. Experience in producing LEU fuel elements for the RSG-GAS

    International Nuclear Information System (INIS)

    To achieve a self-reliance in the operation of the 30 MW Multipurpose Research Reactor at Serpong (the RSG-GAS), a fuel element production facility has been constructed nearby. The main task of the facility is to produce MTR type fuel and control elements containing U3O8-Al dispersion LEU fuel for the RSG-GAS. The hot commissioning activity has started in early 1988 after completion of the cold commissioning using depleted uranium in 1987, marking the beginning of the real production activity. This paper briefly describes the main features of the fuel production facility, the production experience gained so far, and its current production activity. (orig.)

  18. Low enriched aluminide and silicide fuel element technology at B and W (USA)

    International Nuclear Information System (INIS)

    Babcock and Wilcox is fabricating full size fuel elements with low enriched uranium silicide and uranium aluminide. BandW also provides high enrichred U3O8 and UA Lsub(x) for United States Research Reactors, and Test Research and Training Reactors (TRTR). BandW and Argonne National Laboratry (ANL) are actively involved in the Reduced Enrichment Research and Test Reactor (RERTR) Program and have undertaken a joint effort in which BandW is fabricating two Oak Ridge Reactor (ORR ) elements with uranium silicide fuel. During plate development, fuel plates were fabricated with compacts containing U3SiAl and U3Si2 fuel. (author)

  19. Metallographic examination of irradiated nuclear fuel elements at Radiometallurgy Hot Cell Facility

    International Nuclear Information System (INIS)

    Radiometallurgy Hot Cells at the Bhabha Atomic Research Centre, Bombay, are fully equipped to carry out detailed metallographic examination of irradiated fuel elements. Procedures have been standardi.sed for the various steps needed in the preparation of samples suitable for metallographic observation. Existing facilities afford estimation of various parameters, like grain size and other structural changes in fuel and cladding materials, corrosion aspects and the various types of hydride formation in zircaloy clad, pellet clad interaction between fuel and clad which will help in assessing the actual behaviour of fuel elements during operation. (author)

  20. Connection between end plates and rods in a BWR fuel element

    International Nuclear Information System (INIS)

    The problem of the connection between the end plates and the rods of a BWR fuel element is analytically formulated. The behaviour of the springs coupling the rods with the upper plate is analyzed with particular detail since the deformation of these springs affects the forces at the interface of the fuel element structure components. A tool is given to design the springs according to some considerations regarding the mechanical strength of the interacting components as well as the influence of the possible geometrical unevennes of the system that can arise during the fuel element lifetime. (Cali', G.P.)

  1. 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)

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

    International Nuclear Information System (INIS)

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

  3. 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...

  4. 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.)

  5. A finite element model for static strength analysis of CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Ionescu, D.V. [Institute for Nuclear Research, Pitesti (Romania)

    2006-08-15

    A static strength analysis finite-element model has been developed using the ANSYS computer code in order to simulate the axial compression in CANDU type fuel bundle subject to hydraulic drag loads, deflection of fuel elements and stresses and displacements in the end plates. The validation of the finite-element model has been done by comparison with the out-reactor strength test results. Comparison of model predictions with the experimental results showed very good agreement. The comparative assessment reveals that SEU43 and SEU43L fuel bundles are able to withstand high flow rate without showing a significant geometric instability. (orig.)

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

    International Nuclear Information System (INIS)

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

  7. 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)

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

    International Nuclear Information System (INIS)

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

  9. 77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Science.gov (United States)

    2012-03-22

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

  10. 78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Science.gov (United States)

    2013-06-03

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

  11. 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

  12. Estimated Reactivity Increase of Spent Fuel Pool Criticality during Transfer of Fuel Element

    International Nuclear Information System (INIS)

    The spent fuel pool (SFP) planned for Reaktor TRIGA PUSPATI (RTP) are designed to be far below criticality. Acts of withdrawing or inserting more fuel into the SFP shall add a certain amount of reactivity into the system. This paper investigates the amount of reactivity increase or decrease due to insertion of fuel, withdrawal of fuel, insertion of a rack full of fuel and also the withdrawal of this fuel filled rack. This knowledge is essential for the safe handling of the fuel on route to the SFP. (author)

  13. 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)

  14. 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

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

    International Nuclear Information System (INIS)

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

  16. Fuel element in rod form for nuclear reactors

    International Nuclear Information System (INIS)

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

  17. Development and operation of an alarm and monitoring systems for fuel element transports from nuclear facilities

    International Nuclear Information System (INIS)

    A fast measuring system using a NaJ(Tl) scintillator is described which can distinguish between fuel element transports, transports of high- and medium-activity nuclear waste, and transports of other radioactive emitters. (orig.)

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

    International Nuclear Information System (INIS)

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

  19. Plant Design Nuclear Fuel Element Production Capacity Optimization to Support Nuclear Power Plant in Indonesia

    International Nuclear Information System (INIS)

    The optimization production capacity for designing nuclear fuel element fabrication plant in Indonesia to support the nuclear power plant has been done. From calculation and by assuming that nuclear power plant to be built in Indonesia as much as 12 NPP and having capacity each 1000 MW, the optimum capacity for nuclear fuel element fabrication plant is 710 ton UO2/year. The optimum capacity production selected, has considered some aspects such as fraction batch (cycle, n = 3), length of cycle (18 months), discharge burn-up value (Bd) 35,000 up 50,000 MWD/ton U, enriched uranium to be used in the NPP (3.22 % to 4.51 %), future market development for fuel element, and the trend of capacity production selected by advances country to built nuclear fuel element fabrication plant type of PWR. (author)

  20. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore; Ionescu, Dragos; Pauna, Eduard [Institute for Nuclear Research (INR), Pitesti (Romania)

    2011-07-01

    Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the ME01 fuel element withstood 200 power cycles from 27 to 54 kW/m average linear power as well as additional ramps due to reactor trips and restarts during the test period. Both LF tests were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets. This paper presents the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (orig.)

  1. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    International Nuclear Information System (INIS)

    Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the ME01 fuel element withstood 200 power cycles from 27 to 54 kW/m average linear power as well as additional ramps due to reactor trips and restarts during the test period. Both LF tests were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets. This paper presents the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (orig.)

  2. Drying results of K-Basin fuel element 1990 (Run 1)

    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 basins 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 fuels 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 first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element 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 testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0

  3. 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)

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

    International Nuclear Information System (INIS)

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

  5. FEAT4.1: modeling of sheath oxidation and heat flow in CANDU fuel elements

    International Nuclear Information System (INIS)

    This paper describes recent developments in the AECL-developed computer program, FEAT (Finite Element Analysis for Temperature), which is used to assess the thermal integrity of CANDU ® fuel elements. The FEAT code is used to calculate temperatures in the fuel pellet and in the Zircaloy sheath of a CANDU fuel element under normal operating conditions (NOC), as well as the temperature peaking due to end flux peaking during a transient such as a postulated loss of coolant accident (LOCA). For normal operation of high burnup fuel, the Zircaloy oxidation effect on fuel temperatures needs to be considered due to the long residence time in the reactor. The oxide layer on the coolant side of the fuel sheath has a lower thermal conductivity than that of Zircaloy. Therefore, the heat flow from the fuel element to coolant will be reduced resulting in increased fuel pellet and sheath temperatures. To ensure that the FEAT code is suitable for application in analysis of advanced fuels such as the ACR®-1000 fuel, a number of model developments and code improvements were conducted based on the existing version FEAT 4.0, including the modeling of sheath oxidation and its effect on heat flow in the fuel element, time-dependence of end flux peaking during the postulated LOCA (Loss Of Coolant Accident) conditions, pre-processing and postprocessing of analysis data. This paper describes the theories for the models, as well as other improvements, and verification and validation of the new FEAT version (i.e., FEAT 4.1). (author)

  6. On the state of the dispute on the Hanau fuel element factory

    International Nuclear Information System (INIS)

    Siemens are maintaining their position against the political will of the present Land Hesse Government on the Hanau fuel element factory as the place for the manufacture of uranium and MOX fuel elements. The Paragraph 7 AfG authorisation should be provided quickly. A lawsuit is being started against irregular shutdown procedures. One expects support from the Federal German Ministry of the Environment. (DG)

  7. 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)

  8. Nuclear criticality safety assessment of ORR, NBS, and HFBR fuel element shipping package

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, J.T.

    1979-01-01

    A fuel element shipping package employing a borated-phenolic foam as a thermal insulating material is designed to transport as many as seven fuel elements for use in the Oak Ridge Research Reactor, the Brookhaven Fast Beam Reactor, or the National Bureau of Standards Reactor. This report presents the criticality safety evaluation and demonstrates that the requirements for a Fissile Class I package are satisfied by the design.

  9. Nuclear criticality safety assessment of ORR, NBS, and HFBR fuel element shipping package

    International Nuclear Information System (INIS)

    A fuel element shipping package employing a borated-phenolic foam as a thermal insulating material is designed to transport as many as seven fuel elements for use in the Oak Ridge Research Reactor, the Brookhaven Fast Beam Reactor, or the National Bureau of Standards Reactor. This report presents the criticality safety evaluation and demonstrates that the requirements for a Fissile Class I package are satisfied by the design

  10. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. A comparison of measured and predicted can temperatures in simulated CAGR fuel elements

    International Nuclear Information System (INIS)

    The thermal performance of AGR fuel cans is assessed. Can temperatures measured in experiments on simulated AGR fuel elements are compared with predictions made by the HOTSPOT code. The program has been improved by the use of input data based on toughness parameters derived from single pin tests. Agreement between measurement and prediction has been improved. (U.K.)

  14. Use of ELOCA.Mk5 to calculate transient fission product release from CANDU fuel elements

    International Nuclear Information System (INIS)

    A change in fuel element power output, or a change in heat transfer conditions, will result in an immediate change in the temperature distribution in a fuel element. The temperature distribution change will be accompanied by concomitant changes in fuel stress distribution that lead, in turn, to a release of fission products to the fuel-to-sheath gap. It is important to know the inventory of fission products in the fuel-to-sheath gap, because this inventory is a major component of the source term for many postulated reactor accidents. ELOCA.Mk5 is a FORTRAN-77 computer code that has been developed to estimate transient releases to the fuel-to-sheath gap in CANDU reactors. ELOCA.Mk5 is an integration of the FREEDOM fission product release model into the ELOCA fuel element thermo-mechanical code. The integration of FREEDOM into ELOCA allows ELOCA.Mk5 to model the feedback mechanisms between the fission product release and the thermo-mechanical response of the fuel element. This paper describes the physical model, gives details of the ELOCA.Mkt code, and describes the validation of the model. We demonstrate that the model gives good agreement with experimental results for both steady state and transient conditions

  15. Diffractive analysis of annular resonators.

    Science.gov (United States)

    Morin, M; Bélanger, P A

    1992-04-20

    The modal properties of annular resonators are investigated by using an approximate version of the Kirchhoff-Fresnel integral. It is shown that the radial diffraction of a thin annular beam with a large inside radius is similar to that of a cylindrical field distribution. This permits the formal demonstration of the equivalence that exists between large Fresnel number annular resonators and infinite strip resonators. The model explains the properties of annular resonators that have been observed either experimentally or numerically by others, such as the lack of azimuthal discrimination. PMID:20720842

  16. 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

  17. 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))

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

    International Nuclear Information System (INIS)

    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 137Cs 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 134Cs, /sup 110m/Ag, 60Co, and 154Eu 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 3H, 14C, and 90Sr 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 134Cs, relative to 137Cs, 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

  19. Technique for calculating temperature conditions for fuel elements with bilateral heat removal taking into account axial heat fluxes

    International Nuclear Information System (INIS)

    The technique for calculating a fuel element with bilateral heat generation taking into account axial heat fluxes is described. Some results of numerical investigations of fuel element temperature behavior under different coolant flow schemes and when varying operating parameters, fuel element thickness and its thermal conductivity coefficient are presented. It is demonstrated that the effect of axial heat fluxes on heat removal zone redistribution and fuel element temperature conditions respectively enchances with increasing the fuel element thickness and on thermal conductivity coefficient. 5 refs.; 11 figs.; 1 tab

  20. An Experiment on Preparation of Matrix Graphite Powder for HTR fuel element

    International Nuclear Information System (INIS)

    Nuclear fuel for HTGR (High Temperature Gas Cooled Reactor) to produce nuclear hydrogen is called TRISO coated fuel, which consists of 500-μm spherical UO2 particles coated with Pyrolytic Carbon(PyC) and SiC in four layers. The coated TRISO particles are mixed with matrix graphite powder and pressed into a spherical shape of about 60 mm in diameter or a cylindrical compact and heat-treated at about 1900 .deg. C. These fuel elements have a variety of sizes and forms depending on the types of nuclear reactors. Basic steps of manufacturing the fuel element include preparation of graphite matrix powder, overcoating fuel particles, mixing fuel particles with matrix, carbonating green compact and the final high temperature heat processing of the carbonated fuel compact. In order to develop fuel compact fabrication technology, it is important to develop a basic technology for matrix graphite powder (MGP) preparation which has strong influence on the material properties of fuel element. In this work, an experiment was attempted by mixing natural and artificial graphite powders, kneading with binder in methanol medium, drying and milling to prepare a simulated matrix graphite powder with proper characteristics for further steps, i.e., further mixing with coated particles, compaction and heat treatment

  1. 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

  2. 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.

  3. KNK-II fuel elements and their operational behavior

    International Nuclear Information System (INIS)

    Design, fabrication, authorization for operation and the operation itself signify for the project partners involved the step from an research and development program to the procurment of a larger number of fuel subassemblies for a nuclear power plant. Although priority is given to operations experience, those assemblies have to be part of the long-term fuel development program. The first fast core of the KNK-II has reached a residence time of 273 efpd corresponding to a peak burnup of 65.000 MWd/tsub(HM) in 5 assemblies. By such, the originally planned target of 255 efpd has already been exceeded; the authorization for an operation up to 355 efpd has been granted. The extended periods of operation at 100% load as well as the reached burnup prove that the two fuel pin defects (in two assemblies one pin each) do not have any systematic origin. (orig.)

  4. 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)

  5. Dimensional changes in operating UO2 fuel elements: effects 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).

  6. 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)

  7. An experimental investigation of the thermal performance of the CAGR stage II fuel element design

    International Nuclear Information System (INIS)

    The demand for high-power refuelling and extended irradiation in the Civil Advanced Gas-Cooled Reactor (CARG) has necessitated the development of a new fuel element design (Stage II). This element has several geometrical features which differ from the present CAGR reference design (Stage I) and which affect the gas flow and pin temperature distribution within the fuel cluster. This paper describes an experimental investigation to quantify the effect of the design changes on the thermal performance, and thus modify the HOTSPOT computer code to make predictions for the Stage II fuel geometry. (author)

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

    International Nuclear Information System (INIS)

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

  9. 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

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

    International Nuclear Information System (INIS)

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

  11. Remarks on the transportation of spent fuel elements

    International Nuclear Information System (INIS)

    In this chapter topics discussed are the need for contracts, a transport company and risk insurance. Also, a section on transportation covers cranes, subpressure, contamination, cask limitations, physical protection and shipping. Reprocessing discusses minimum reprocessing batch and spent fuel. Finally, economical considerations concerning transportation and reprocessing are given

  12. 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

  13. 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

  14. 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.

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

    International Nuclear Information System (INIS)

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

  16. Critical experiments simulating accidental water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    The paper focuses on experimental analysis of nuclear criticality safety at accidental water immersion of fuel elements of the Russian TOPAZ-2 space nuclear power system reactor. The structure of water-moderated heterogeneous critical assemblies at the NARCISS facility is described in detail, including sizes, compositions, densities of materials of the main assembly components for various core configurations. Critical parameters of the assemblies measured for varying number of fuel elements, height of fuel material in fuel elements and their arrangement in the water moderator with a uniform or variable spacing are presented. It has been found from the experiments that at accidental water immersion of fuel elements involved, the minimum critical mass equal to approximately 20 kg of uranium dioxide is achieved at 31-37 fuel elements. The paper gives an example of a physical model of the water-moderated heterogeneous critical assembly with a detailed characterization of its main components that can be used for calculations using different neutronic codes, including Monte Carlo ones. (author)

  17. 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.

  18. 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)

  19. 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)

  20. 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)