Nuclear fuel element

International Nuclear Information System (INIS)

Thompson, J.R.; Rowland, T.C.

1976-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

Effect of lattice deformation on temperature fields and heat transfer in the fuel elements of characteristic zones for a model of fast reactor fuel assembly

International Nuclear Information System (INIS)

Zhukov, A.V.; Matyukhin, N.M.; Sviridenko, E.Ya.

1980-01-01

Given are the experimental results for temperature fields in the model assembly in nonribbed simulators of the BN-600-type reactor fuel elements in the course of deformation of the lattice caused by shifting of the central and peripheral (lateral, angular) fuel elements by the value of the gap between the fuel elements (the limiting case when the fuel elements touch each other along the whole length). An assembly consisting of 37 electroheated pipes arranged in a triangular lattice with a relative step of S/d=1.185 is used as a model. The experiments were carried out on the sodium stand at constant energy release along the length of the fuel element simulators and at the Pe number changing in the 14-700 range. The data obtained show considerable increase of nonuniformities of the fuel element temperatures for characteristic zones of the fuel cassette assembly models of the fast reactor at deviations of the lattice geometric sizes from the nominal ones. For the central nonribbed element the temperature nonuniformity increases approximately 7.5 times and for the lateral element approximately 6 times when the elements touch each other along the whole length. The shift the central nonribbed element by the value of the gap between the fu.el elements leads to the decrease of heat transfer in comparison with heat transfer at the nominal geometry approximately 3-7 times in the 10-450 range for the Pe numbers. It is shown that the coolant temperature distribution along the assembly radius has a complex character (with a peak between the centre and the perifery) caused by redistribution of coolant consumptions due to fuel element lattice deformation

Thermally-induced bowing of CANDU fuel elements

International Nuclear Information System (INIS)

Suk, H.C.; Sim, K.S.; Park, J.H.; Park, G.S.

1995-01-01

Considering only the thermally-induced bending moments which are generated both within the sheath and between the fuel and sheath by an asymmetric temperature distribution with respect to the axis of an element, a generalized and explicit analytical formula for the thermally-induced bending is developed in this paper, based on the cases of 1) the bending of an empty tube treated by neglecting of the fuel/sheath mechanical interaction and 2) the fuel/sheath interaction due to the pellet and sheath temperature variations. In each of the cases, the temperature asymmetries in sheath are modelled to be caused by the combined effects of (i) non-uniform coolant temperature due to imperfect coolant mixing, (ii) variable sheath/coolant heat transfer coefficient, (iii) asymmetric heat generation due to neutron flux gradients across an element and so as to inclusively cover the uniform temperature distributions within the fuel and sheath with respect to the axial centerline. Investigating the relative importance of the various parameters affecting fuel element bowing, the element bowing is found to be greatly affected with the variations of element length, sheath diameter, pellet/sheath mechanical interaction and neutron flux depression factors, pellet thermal expansion coefficient, pellet/sheath heat transfer coefficient in comparison with those of other parameters such as sheath thickness, film heat transfer coefficient, sheath thermal expansion coefficient, and sheath and pellet thermal conductivities. Also, the element bowing of the standard 37-element bundle and CANFLEX 43-element bundle for the use in CANDU-6 reactors was analyzed with the formula, which could help to demonstrate the integrity of the fuel. All the required input data for the analyses were generated in terms of the reactor operation conditions on the reactor physics, thermal hydraulics and fuel performance by using various CANDU computer codes. The analysis results indicate that the CANFLEX 43-element

Critical heat fluxes in tubular fuel elements of nuclear power reactors

International Nuclear Information System (INIS)

Subbotin, V.I.; Alekseev, G.V.; Peskov, O.L.

1974-01-01

The results of the experiments carried out show that with appropriate choice of tube, type and dimensions of intensifier the attainment of critical conditions at certain parameters is not accompanied by sharp or considerable increases in temperature of the heat removing surface. Increase in power to above critical under these conditions does not lead to considerable variation in temperature either. Thus, it appears possible to change from heat removal by steam-water mixture to convective heat removal by wet steam without manifestation of intolerable temperature conditions of the heating surface (Fig. 6). A change to convective heat removal by wet steam is possible at different levels of heat fluxes which depend during constant conditions at the inlet on tube length and the degree of the disturbing influence on the flow. This is especially important since in principle the possibility arises for developing a power reactor with tubular fuel elements, in which a once-through cycle with steam superheat involving no intermediate separation can be realised

Nuclear fuel elements

International Nuclear Information System (INIS)

Kawada, Toshiyuki; Hirayama, Satoshi; Yoneya, Katsutoshi.

1980-01-01

Purpose: To enable load-depending operation as well as moderation for the restriction of operation conditions in the present nuclear reactors, by specifying the essential ingredients and the total weight of the additives to UO 2 fuel substances. Constitution: Two or more additives selected from Al 2 O 3 , B 2 O, CaO, MgO, SiO 2 , Na 2 O and P 2 O 5 are added by the total weight of 2 - 5% to fuel substances consisting of UO 2 or a mixture of UO 2 and PuO 2 . When the mixture is sintered, the strength of the fuel elements is decreased and the fuel-cladding interactions due to the difference in the heat expansion coefficients between the ceramic fuel elements and the metal claddings are decreased to a substantially harmless degree. (Horiuchi, T.)

Heat conduction in a plate-type fuel element with time-dependent boundary conditions

International Nuclear Information System (INIS)

Faya, A.J.G.; Maiorino, J.R.

1981-01-01

A method for the solution of boundary-value problems with variable boundary conditions is applied to solve a heat conduction problem in a plate-type fuel element with time dependent film coefficient. The numerical results show the feasibility of the method in the solution of this class of problems. (Author) [pt

Loads on pebble bed fuel elements

Energy Technology Data Exchange (ETDEWEB)

Teuchert, E.; Maly, V.

1974-03-15

A comparison is made of key parameters for multi-recycle pebbles and single-pass once-through (OTTO) pebbles. The parameters analyzed include heat transfer characteristics with burn-up, temperature profiles, power per element as a function of axial position in the core, and burn-up. For the OTTO-scheme, the comparisons addressed the use of the conventional fuel element and the advanced "shell ball" designed to reduce the peak fuel temperature in the center of the fuel element. All studies addressed the uranium-thorium fuel cycle.

Thermal insulation of fuel elements

International Nuclear Information System (INIS)

Dubrovcak, P.; Pec, V.; Pitonak, J.

1978-01-01

The claim of the invention concerns thermal insulation of fuel elements heated for measurement of uranium fuel physical properties. For this, layers of aluminium film and of glass fibre are wound onto the inner tube of the element cladding. The space between the inner and the outer tubes is evacuated and the tubes are spaced using spacer wires. (M.S.)

Experimental research of fuel element reliability

International Nuclear Information System (INIS)

Cech, B.; Novak, J.; Chamrad, B.

1980-01-01

The rate and extent of the damage of the can integrity for fission products is the basic criterion of reliability. The extent of damage is measurable by the fission product leakage into the reactor coolant circuit. An analysis is made of the causes of the fuel element can damage and a model is proposed for testing fuel element reliability. Special experiments should be carried out to assess partial processes, such as heat transfer and fuel element surface temperature, fission gas liberation and pressure changes inside the element, corrosion weakening of the can wall, can deformation as a result of mechanical interactions. The irradiation probe for reliability testing of fuel elements is described. (M.S.)

Model studying the processes arising during fuel element overheating

International Nuclear Information System (INIS)

Usynin, G.B.; Anoshkin, Yu.I.; Vlasichev, G.N.; Galitskikh, Yu.N.; Semenychev, M.A.

1986-01-01

A calculational technique for studying heating and melting of fuel elements in the BN type reactors during an accident with heat release failure and a simulator with central rod heater intended for out-of-pile experiments is developed. The time rangeof the characteristic melting steps for the most thermally stressed fuel element at the reactor nominal power is calculated. The experimental study of the fuel element melting using a simulator with a tungsten heater has proved that the technique for the simultor and fuel can melting, respectively, is correct. The developed technique is used for determining the geometrical values and operational conditions for experiments with simulators, when quantitative and qualitative characteristics of the process under study are rather close to those natural for fuel elements

Effect of the Heat Treatment on the Graphite Matrix of Fuel Element for HTGR

International Nuclear Information System (INIS)

Lee, Chungyong; Lee, Seungjae; Suh, Jungmin; Jo, Youngho; Lee, Youngwoo; Cho, Moonsung

2013-01-01

In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength for the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and Phenol as a binder were chosen and mixed with each other, formed and heated for the compressive strength test. The objective of this research is to optimize the kinds and composition of the mixed graphite and the forming process by evaluating the compressive strength before/after heat treatment (carbonization of binder). In this study, the effect of heat treatment on graphite matrix was studied in terms of the density and the compressive strength. The size (diameter and length) of pellet is increased by heat treatment. Due to additional weight reduction and swelling (length and diameter) of samples the density of graphite pellet is decreased from about 2.0 to about 1.7g/cm 3 . From the mechanical test results, the compressive strength of graphite pellets was related to the various conditions such as the contents of binder, the kinds of graphite and the heat treatment. Both the green pellet and the heat treated pellet, the compressive strength of G+S+P pellets is relatively higher than that of R+S+P pellets. To optimize fuel element matrix, the effect of Phenol and other binders, graphite composition and the heat treatment on the mechanical properties will be deeply investigated for further study

Heat transfer analysis in internally-cooled fuel elements by means of a conformal mapping approach

International Nuclear Information System (INIS)

Sarmiento, G.S.; Laura, P.A.A.

1981-01-01

The present paper deals with an approximate solution of the steady-state heat conduction problem in internally cooled fuel elements of fast breeder reactors. Explicit expressions for the dimensionless temperature distribution in terms of the governing physical and geometrical parameters are determined by means of a coupled conformal mapping-variational approach. The results obtained are found to be in very good agreement with those calculated by means of a finite element code. (orig.)

Requirements for materials of dispersion fuel elements

International Nuclear Information System (INIS)

Samojlov, A.G.; Kashtanov, A.I.; Volkov, V.S.

1982-01-01

Requirements for materials of dispersion fuel elements are considered. The necessity of structural and fissile materials compatibility at maximum permissible operation temperatures and temperatures arising in a fuel element during manufacture is pointed out. The fuel element structural material must be ductile, possess high mechanical strength minimum neutron absorption cross section, sufficient heat conductivity, good corrosion resistance in a coolant and radiation resistance. The fissile material must have high fissile isotope concentration, radiation resistance, high thermal conductivity, certain porosity high melting temperature must not change the composition under irradiation

Applying Thermodynamics to Fossil Fuels: Heats of Combustion from Elemental Compositions.

Science.gov (United States)

Lloyd, William G.; Davenport, Derek A.

1980-01-01

Discussed are the calculations of heats of combustions of some selected fossil fuel compounds such as some foreign shale oils and United States coals. Heating values for coal- and petroleum-derived fuel oils are also presented. (HM)

Measurement of the fuel temperature and the fuel-to-coolant heat transfer coefficient of Super Phenix 1 fuel elements

International Nuclear Information System (INIS)

Edelmann, M.

1995-12-01

A new measurement method for measuring the mean fuel temperature as well as the fuel-to-coolant heat transfer coefficient of fast breeder reactor subassemblies (SA) is reported. The method is based on the individual heat balance of fuel SA's after fast reactor shut-downs and uses only the plants normal SA outlet temperature and neutron power signals. The method was used successfully at the french breeder prototype Super Phenix 1. The mean SA fuel temperature as well as the heat transfer coefficient of all SPX SA's have been determined at power levels between 15 and 90% of nominal power and increasing fuel burn-up from 3 to 83 EFPD (Equivalent of Full Power-Days). The measurements also provided fuel and whole SA time constants. The estimated accuracy of measured fuel parameters is in the order of 10%. Fuel temperatures and SA outlet temperature transients were also calculated with the SPX1 systems code DYN2 for exactly the same fuel and reactor operating parameters as in the experiments. Measured fuel temperatures were higher than calculated ones in all cases. The difference between measured and calculated core mean values increases from 50 K at low power to 180 K at 90% n.p. This is about the double of the experimental error margins. Measured SA heat transfer coefficients are by nearly 20% lower than corresponding heat transfer parameters used in the calculations. Discrepancies found between measured and calculated results also indicate that either the transient heat transfer in the gap between fuel and cladding (gap conductance) might not be exactly reproduced in the computer code or that the gap in the fresh fuel was larger than assumed in the calculations. (orig.) [de

TAPIR, Thermal Analysis of HTGR with Graphite Sleeve Fuel Elements

International Nuclear Information System (INIS)

Weicht, U.; Mueller, W.

1983-01-01

1 - Nature of the physical problem solved: Thermal analysis of a reactor core containing internally and/or externally gas cooled prismatic fuel elements of various geometries, rating, power distribution, and material properties. 2 - Method of solution: A fuel element in this programme is regarded as a sector of a fuelled annulus with graphite sleeves of any shape on either side and optional annular gaps between fuel and graphite and/or within the graphite. It may have any centre angle and the fuelled annulus may become a solid cylindrical rod. Heat generation in the fuel is assumed to be uniform over the cross section and peripheral heat flux into adjacent sectors is ignored. Fuel elements and coolant channels are treated separately, then linked together to fit a specified pattern. 3 - Restrictions on the complexity of the problem: Maxima of: 50 fuel elements; 50 cooled channels; 25 fuel geometries; 25 coolant channel geometries; 10 axial power distributions; 10 graphite conductivities

Calculation of the fuel temperature field under heat release and heat conductance transient conditions

International Nuclear Information System (INIS)

Kazakov, E.K.; Chernukhina, G.M.

1974-01-01

Results of calculation of the temperature distribution in an annular fuel element at transient thermal conductivity and heat release values are given. The calculation has been carried out by the mesh technique with the third-order boundary conditions for the inner surface assumed and with heat fluxes and temperatures at the zone boundaries to be equal. Three variants of solving the problem of a stationary temperature field are considered for failed fuel elements with clad flaking or cracks. The results obtained show the nonuniformity of the fuel element temperature field to depend strongly on the perturbation parameter at transient thermal conductivity and heat release values. In case of can flaking at a short length, the core temperature rises quickly after flaking. While evaluating superheating, one should take into account the symmetry of can flaking [ru

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.

Calculation of Heat-Bearing Agent’s Steady Flow in Fuel Bundle

Science.gov (United States)

Amosova, E. V.; Guba, G. G.

2017-11-01

This paper introduces the result of studying the heat exchange in the fuel bundle of the nuclear reactor’s fuel magazine. The article considers the fuel bundle of the infinite number of fuel elements, fuel elements are considered in the checkerboard fashion (at the tops of a regular triangle a fuel element is a plain round rod. The inhomogeneity of volume energy release in the rod forms the inhomogeneity of temperature and velocity fields, and pressure. Computational methods for studying hydrodynamics in magazines and cores with rod-shape fuel elements are based on a significant simplification of the problem: using basic (averaged) equations, isobaric section hypothesis, porous body model, etc. This could be explained by the complexity of math description of the three-dimensional fluid flow in the multi-connected area with the transfer coefficient anisotropy, curved boundaries and technical computation difficulties. Thus, calculative studying suggests itself as promising and important. There was developed a method for calculating the heat-mass exchange processes of inter-channel fuel element motions, which allows considering the contribution of natural convection to the heat-mass exchange based on the Navier-Stokes equations and Boussinesq approximation.

Theoretical models to predict the transient heat transfer performance of HIFAR fuel elements under non-forced convective conditions

International Nuclear Information System (INIS)

Green, W.J.

1987-04-01

Simple theoretical models have been developed which are suitable for predicting the thermal responses of irradiated research fuel elements of markedly different geometries when they are subjected to loss-of-coolant accident conditions. These models have been used to calculate temperature responses corresponding to various non-forced convective conditions. Comparisons between experimentally observed temperatures and calculated values have shown that a suitable value for surface thermal emissivity is 0.35; modelling of the fuel element beyond the region of the fuel plate needs to be included since these areas account for approximately 25 per cent of the thermal power dissipated; general agreement between calculated and experimental temperatures for both transient and steady-state conditions is good - the maximum discrepancy between calculated and experimental temperatures for a HIFAR Mark IV/V fuel element is ∼ 70 deg C, and for an Oak Ridge Reactor (ORR) box-type fuel element ∼ 30 deg C; and axial power distribution does not significantly affect thermal responses for the conditions investigated. Overall, the comparisons have shown that the models evolved can reproduce experimental data to a level of accuracy that provides confidence in the modelling technique and the postulated heat dissipation mechanisms, and that these models can be used to predict thermal responses of fuel elements in accident conditions that are not easily investigated experimentally

Nuclear reactor fuel element

International Nuclear Information System (INIS)

D'Eye, R.W.M.; Shennan, J.V.; Ford, L.H.

1977-01-01

Fuel element with particles from ceramic fissionable material (e.g. uranium carbide), each one being coated with pyrolitically deposited carbon and all of them being connected at their points of contact by means of an individual crossbar. The crossbar consists of silicon carbide produced by reaction of silicon metal powder with the carbon under the influence of heat. Previously the silicon metal powder together with the particles was kneaded in a solvent and a binder (e.g. epoxy resin in methyl ethyl ketone plus setting agent) to from a pulp. The reaction temperature lies at 1750 0 C. The reaction itself may take place in a nitrogen atmosphere. There will be produced a fuel element with a high overall thermal conductivity. (DG) [de

Computer simulation of fuel element performance

Energy Technology Data Exchange (ETDEWEB)

Sukhanov, G I

1979-01-01

The review presents reports made at the Conference on the Bahaviour and Production of Fuel for Water Reactors on March 13-17, 1979. Discussed at the Conference are the most developed and tested calculation models specially evolved to predict the behaviour of fuel elements of water reactors. The following five main aspects of the problem are discussed: general conceptions and programs; mechanical mock-ups and their applications; gas release, gap conductivity and fuel thermal conductivity; analysis of nonstationary processes; models of specific phenomena. The review briefly describes the physical principles of the following models and programs: the RESTR, providing calculation of the radii of zones of columnar and equiaxial grains as well as the radius of the internal cavity of the fuel core; programs for calculation of fuel-can interaction, based on the finite elements method; a model predicting the behaviour of the CANDU-PHW fuel elements in transient conditions. General results are presented of investigations of heat transfer through a can-fuel gap and thermal conductivity of UO/sub 2/ with regard for cracking and gas release of the fuel. Many programs already suit the accepted standards and are intensively tested at present.

Improved lumped parameter for annular fuel element thermohydraulic analysis

International Nuclear Information System (INIS)

Duarte, Juliana Pacheco; Su, Jian; Alvim, Antonio Carlos Marques

2011-01-01

Annular fuel elements have been intensively studied for the purpose of increasing power density in light water reactors (LWR). This paper presents an improved lumped parameter model for the dynamics of a LWR core with annular fuel elements, composed of three sub-models: the fuel dynamics model, the neutronics model, and the coolant energy balance model. The transient heat conduction in radial direction is analyzed through an improved lumped parameter formulation. The Hermite approximation for integration is used to obtain the average temperature of the fuel and cladding and also to obtain the average heat flux. The volumetric heat generation in fuel rods was obtained with the point kinetics equations with six delayed neutron groups. The equations for average temperature of fuel and cladding are solved along with the point kinetic equations, assuming linear reactivity and coolant temperature in cases of reactivity insertion. The analytical development of the model and the numeric solution of the ordinary differential equation system were obtained by using Mathematica 7.0. The dynamic behaviors for average temperatures of fuel, cladding and coolant in transient events as well as the reactor power were analyzed. (author)

ELOCA: fuel element behaviour during high temperature transients

International Nuclear Information System (INIS)

Sills, H.E.

1979-03-01

The ELOCA computer code was developed to simulate the uniform thermal-mechanical behaviour of a fuel element during high-temperature transients such as a loss-of-coolant accident (LOCA). Primary emphasis is on the diametral expansion of the fuel sheath. The model assumed is a single UO2/zircaloy-clad element with axisymmetric properties. Physical effects considered by the code are fuel expansion, cracking and melting; variation, during the transient, of internal gas pressure; changing fuel/sheath heat transfer; thermal, elastic and plastic sheath deformation (anisotropic); Zr/H 2 O chemical reaction effects; and beryllium-assisted crack penetration of the sheath. (author)

Method of making a graphite fuel element having carbonaceous fuel bodies

International Nuclear Information System (INIS)

Miertschin, G.N.; Leary, D.F.

1977-01-01

Particulate nuclear fuel material, particulate carbon and pitch are combined with an additive which is effective to reduce the coke yield upon carbonization to mold a green fuel body. The additive may be polystyrene, a styrene-butadiene copolymer, an aromatic hydrocarbon having a molecular weight between about 75 and 300 or a saturated hydrocarbon polymer. The green fuel body is inserted in a complementary cavity within a porous nuclear fuel element body and heated in situ to decompose the pitch and additive, leaving a relatively close-fitting fuel body in the cavity

Sodium removal of fuel elements by vacuum distillation

International Nuclear Information System (INIS)

Buescher, E.; Haubold, W.; Jansing, W.; Kirchner, G.

1978-01-01

Cleaning of sodium-wetted core components can be performed by using either lead, moist nitrogen, or alcohol. The advantages of these methods for cleaning fuel elements without causing damage are well known. The disadvantage is that large amounts of radioactive liquids are formed during handling in the latter two cases. In this paper a new method to clean components is described. The main idea is to remove all liquid metal from the core components within a comparatively short period of time. Fuel elements removed from the reactor must be cooled because of high decay heat release. To date, vacuum distillation of fuel elements has not yet been applied

Heat transfer and pressure drop of the reactor fuel element with polyzonal spiral finning; Prelaz toplote i pad pritiska reaktorskog gorivnog elementa sa polizonalno-spiralnim orebrenjem

Energy Technology Data Exchange (ETDEWEB)

Oka, S; Becirspahic, S [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

1964-06-15

Heat transfer and pressure drop of the reactor fuel element with polyzonal spiral finning were investigated. Longitudinal and circumferential distributions of Sr-number of finnings in the fuel element are given. Dependences of St{sub kmin} and St{sub ksr} on the Re number are derived. The influence of gap between two fuel elements on the heat transfer, pressure drop is presented dependent on the Re number. The influence of mutual position of flow separators of two neighbouring fuel elements on the pressure drop and heat transfer is shown as well. Investigations were performed in the range of Re numbers from 15000 to 100000. Ispitivan je prelaz toplote i pad pritiska modela reaktorskog gorivnog elementa sa polizonalno-spiralnim orebrenjem. Dat je uzduznu i obimni raspored Sr-broja na orebrenju gorivnog elementa. Izvedene su zavisnosti St{sub kmin} i St{sub ksr} u funkciji od Re-broja. Pokazan je uticaj prekida izmedju dva gorivna elementa na prelaz toplote i pad pritiska u zavisnosti od Re-broja. Pokazan je uticaj medjusobnog polozaja razdeljivaca struje dva susedna gorivna elementa na pad pritiska i prelaz toplote. Ispitivanja su vrsena u oblasti Re-brojeva od 15000 do 100000 (author)

Impact of thermal conductivity models on the coupling of heat transport, oxygen diffusion, and deformation in (U, Pu)O nuclear fuel elements

Science.gov (United States)

Mihaila, Bogdan; Stan, Marius; Crapps, Justin; Yun, Di

2013-02-01

We study the coupled thermal transport, oxygen diffusion, and thermal expansion in a generic nuclear fuel rod consisting of a (U) fuel pellet separated by a helium gap from zircaloy cladding. Steady-state and time-dependent finite-element simulations with a variety of initial- and boundary-value conditions are used to study the effect of the Pu content, y, and deviation from stoichiometry, x, on the temperature and deformation profiles in this fuel element. We find that the equilibrium radial temperature and deformation profiles are most sensitive to x at small values of y. For larger values of y, the effects of oxygen and Pu content are equally important. Following a change in the heat-generation rate, the centerline temperature, the radial deformation of the fuel pellet, and the centerline deviation from stoichiometry track each other closely in (U,Pu)O, as the characteristic time scales of the heat transport and oxygen diffusion are similar. This result is different from the situation observed in the case of UO fuels.

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

International Nuclear Information System (INIS)

McCulloch, R.W.; Lovell, R.T.; Post, D.W.; Snyder, S.D.

1980-01-01

Variable width ribbon heating elements have been fabricated which provide a chopped cosine, variable heat flux profile for fuel rod simulators used in test loops by the Breeder Reactor Program Thermal Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor Core Flow Test Loop. Thermal, mechanical, and electrical design considerations result in the derivation of an analytical expression for the ribbon contours. From this, the ribbons are machined and wound on numerically controlled equipment. Postprocessing and inspection results in a wound, variable width ribbon with the precise dimensional, electrical, and mechanical properties needed for use in fuel pin simulators

KANDY - a numerical model to describe phenomena, which - in a heated and voided fuel element of an LMFBR - may occur

International Nuclear Information System (INIS)

Thurnay, K.

1984-02-01

Kandy is a model developed to describe the essential destructionphenomena of the fuel elements of an LMFBR. The fuel element is assumed to be a voided one, in which the heat generation is still going on. The main process to be modeled is the melting/bursting/evaporating of parts of the fuel pins and the subsequent dislocation of these materials in the coolant channel. The work presented summarizes the assumptions constituting the model, develops the corresponding equations of motion and describes the procedure, turning these into a system of difference-equations ready for coding. As a final part results of a testcase calculation with the Kandy-code are presentend and interpreted. (orig.) [de

Storage container for radioactive fuel elements

International Nuclear Information System (INIS)

1984-01-01

The interim storage cask for spent fuel elements or the glass moulds for high-level radioactive waste are made up of heat-resistant, reinforced concrete with chambers and highgrade steel lining. Cooling systems with natural air circulation are connected with the chambers. (HP) [de

Report on the disposal of radioactive wastes and spent fuel elements from Baden-Wuerttemberg

International Nuclear Information System (INIS)

2017-04-01

The report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg covers the following issues: legal framework for the nuclear disposal; producer of spent fuels and radioactive wastes in Baden- Report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg; low- and medium-level radioactive wastes (non heat generating radioactive wastes); spent fuels and radioactive wastes from waste processing (heat generating radioactive wastes); final disposal.

Transient and steady-state analyses of an electrically heated Topaz-II Thermionic Fuel Element

International Nuclear Information System (INIS)

El-Genk, M.S.; Xue, H.

1992-01-01

Transient and steady-state analyses of electrically heated, Thermionic Fuel Elements (TFEs) for Topaz-II space power system are performed. The calculated emitter and collector temperatures, load electric power and conversion efficiency are in good agreement with reported data. In this paper the effects or Cs pressure, thermal power input, and load resistance on the steady-state performance of the TFE are also investigated. In addition, the thermal response of the ZrH moderator during a startup transient and following a change in the thermal power input is examined

A combined wet/dry sipping cell for investigating failed triga fuel elements

International Nuclear Information System (INIS)

Boeck, H.; Gallhammer, H.; Hammer, J.; Israr, M.

1987-08-01

A sipping cell to detect failed triga fuel has been designed and constructed at the Atominstitut. The cell allows both wet- and dry sipping of one single standard triga fuel element. In the dry sipping method the fuel element may be electrically heated up to a maximum temperature of about 300 0 C to allow the detection of temperature dependent fission product release from the fuel element. 20 figs., 1 tab. (Author)

Determining reactor fuel elements broken by Cerenkov counting

International Nuclear Information System (INIS)

Guo Juhao; Dong Shiyuan; Feng Yuying

1996-01-01

The basis and method of determining fuel elements broken in a reactor by Cerenkov counting measured with liquid scintillation spectrometer are introduced. The radioactive characteristic of the radiation nuclides generating Cherenkov radiation in the primary water of 200 MW nuclear district heating reactor is analyzed. The activity of the activation products in the primary water and the fission products in the fuel elements are calculated. A feasibility of Cerenkov counting measure was analyzed. This method is simple and quick

FUDA MOD-2: a computer program for simulation the performance of fuel element validation exercise

International Nuclear Information System (INIS)

Chouhan, S.K.; Tripathi, R.M.; Prasad, P.N.; Chauhan, Ashok

2014-01-01

The PHWR fuel element performance is evaluated using the fuel analysis computer code FUDA MOD2. It is specifically written for performance simulation of UO 2 fuel pellet, located in zirconium alloy sheath operating under coolant pressure. For specific element power histories, the code investigates the variables and their interactions that govern fuel element performance. The input data requires pellet dimensions, element dimensions, sheath properties, heat transfer data, thermal hydraulic parameters of coolant, the inner filler gas composition, flux gradient and linear heat ratings (LHR) at different burn up. The output data generated by the code are radial temperature profile of fuel and sheath, fuel sheath-gap heat transfer coefficient, fission gas generated and released, fission gas pressure, sheath stress and strain for different burn-up zones. The code has been verified against literature data and post irradiation examinations carried out. It is also bench marked against various international fuel element simulation programmes available with water cooled reactors operating countries. The present paper describes the FUDA MOD2 code verification studies carried out using the literature data and post irradiation examination data. (author)

Heat removal in gas-cooled fuel rod clusters

International Nuclear Information System (INIS)

Rehme, K.

1975-01-01

For a thermo- and fluid-dynamic analysis of fuel rod cluster subchannels for gas-cooled breeder reactors, the following values must be verified: a) friction coefficient as flow parameter; b) Stanton number as heat transfer parameter; c) influence of spacers on friction coefficient and Stanton number; d) heat and mass exchange between subchannels with different temperatures. These parameters are established by combining results of single experiments and of integral experiments. Mention is made of further studies to be performed in order to determine the heat removal from gas-cooled fast breeder fuel elements. (HR) [de

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

International Nuclear Information System (INIS)

Roy, P.R.; Purushotham, D.S.C.; Majumdar, S.

1983-01-01

Three clusters, each containing 6 zircaloy-2 clad short length fuel elements of either MOX or UO 2 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/cm 2 ; 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 UO 2 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)

Numerical simulations of helium flow through prismatic fuel elements of very high temperature reactors

International Nuclear Information System (INIS)

Ribeiro, Felipe Lopes; Pinto, Joao Pedro C.T.A.

2013-01-01

The 4 th generation Very High Temperature Reactor (VHTR) most popular concept uses a graphite-moderated and helium cooled core with an outlet gas temperature of approximately 1000 deg C. The high output temperature allows the use of the process heat and the production of hydrogen through the thermochemical iodine-sulfur process as well as highly efficient electricity generation. There are two concepts of VHTR core: the prismatic block and the pebble bed core. The prismatic block core has two popular concepts for the fuel element: multihole and annular. In the multi-hole fuel element, prismatic graphite blocks contain cylindrical flow channels where the helium coolant flows removing heat from cylindrical fuel rods positioned in the graphite. In the other hand, the annular type fuel element has annular channels around the fuel. This paper shows the numerical evaluations of prismatic multi-hole and annular VHTR fuel elements and does a comparison between the results of these assembly reactors. In this study the analysis were performed using the CFD code ANSYS CFX 14.0. The simulations were made in 1/12 fuel element models. A numerical validation was performed through the energy balance, where the theoretical and the numerical generated heat were compared for each model. (author)

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

International Nuclear Information System (INIS)

McCulloch, R.W.; Post, D.W.; Lovell, R.T.; Snyder, S.D.

1981-04-01

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

Theoretical study of fuel element reliability in the BRIG-300 fast reactor

International Nuclear Information System (INIS)

Kulikov, I.S.; Nesterenko, V.B.; Tverkovkin, B.E.

1983-01-01

The theoretical results on studies of the reliability of cermet symmetrically heated fuel elements under conditions of the BRIG-300 fast gas cooled reactor are presented. The investigations have been conducted at the Nuclear Power Engineering Institute of the Byelorussian Academy of Sciences. Two variants of the fuel elements are considered :the fuel element with the gas gap between fuel and can and the fuel element with tight contact between cermet fuel and can. The estimated data on can resistance, swelling of the fuel rods and cans, strains and stresses in cans, change of the gap and its thermal coductivity during the reactor operation are obtained. The results of the analysis show that cermet fuel has sufficient reliability upon oparational conditions of the reactor with dissociating gas coolant in a steady-state regime

Whole-core damage analysis of EBR-II driver fuel elements following SHRT program

International Nuclear Information System (INIS)

Chang, L.K.; Koenig, J.F.; Porter, D.L.

1987-01-01

In the Shutdown Heat Removal Testing (SHRT) program in EBR-II, fuel element cladding temperatures of some driver subassemblies were predicted to exceed temperatures at which cladding breach may occur. A whole-core thermal analysis of driver subassemblies was performed to determine the cladding temperatures of fuel elemnts, and these temperatures were used for fuel element damage calculation. The accumulated cladding damage of fuel element was found to be very small and fuel element failure resulting from SHRT transients is unlikely. No element breach was noted during the SHRT transients. The reactor was immediately restarted after the most severe SHRT transient had been completed and no driver fuel breach has been noted to date. (orig.)

Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

International Nuclear Information System (INIS)

Verfondern, K.; Mueller, D.

1991-01-01

Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

Test of high temperature fuel element, (1)

International Nuclear Information System (INIS)

Akino, Norio; Shiina, Yasuaki; Nekoya, Shin-ichi; Takizuka, Takakazu; Emori, Koichi

1980-11-01

Heat transfer experiment to measure the characteristics of a VHTR fuel in the same condition of the reactor core was carried out using HTGL (High Temperature Helium Gas Loop) and its test section. In this report, the details of the test section, related problems of construction and some typical results are described. The newly developed heater with graphite heat transfer surface was used as a simulated fuel element to determine the heat transfer characteristics. Following conclusions were obtained; (1) Reynolds number between turbulent and transitional region is about 2600. (2) Reynolds number between transitional and laminar region is about 4800. (3) The laminarization phenomena have not been observed and are hardly occurred in annular tubes comparing with round tube. (4) Measured Nusselt numbers agree to the established correlations in turbulent and laminar regions. (author)

Radial heat conduction in a power reactor fuel element

International Nuclear Information System (INIS)

Ventura, M.A.

1998-01-01

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 UO 2 pellet and the clad and its associated uncertainty is included in the steady state model. (author) [es

Nuclear fuel element

International Nuclear Information System (INIS)

Penrose, R.T.; Thompson, J.R.

1976-01-01

A method of protecting the cladding of a nuclear fuel element from internal attack and a nuclear fuel element for use in the core of a nuclear reactor are disclosed. The nuclear fuel element has disposed therein an additive of a barium-containing material and the barium-containing material collects reactive gases through chemical reaction or adsorption at temperatures ranging from room temperature up to fuel element plenum temperatures. The additive is located in the plenum of the fuel element and preferably in the form of particles in a hollow container having a multiplicity of gas permeable openings in one portion of the container with the openings being of a size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles. The additive is comprised of elemental barium or a barium alloy containing one or more metals in addition to barium such as aluminum, zirconium, nickel, titanium and combinations thereof. 6 claims, 3 drawing figures

Fuel element loading system

International Nuclear Information System (INIS)

Arya, S.P; s.

1978-01-01

A nuclear fuel element loading system is described which conveys a plurality of fuel rods to longitudinal passages in fuel elements. Conveyor means successively position the fuel rods above the longitudinal passages in axial alignment therewith and adapter means guide the fuel rods from the conveyor means into the longitudinal passages. The fuel elements are vibrated to cause the fuel rods to fall into the longitudinal passages through the adapter means

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

Science.gov (United States)

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

2016-01-01

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

Experience of developing the imitators of the fuel element for the WWER reactors

International Nuclear Information System (INIS)

Balashov, S.M.; Boltenko, Eh.A.; Vinogradov, V.A.

1998-01-01

Peculiarities of designs of fuel elements imitators for the WWER-type reactors of nominal capacity and with single-ended current feed positioning are considered. The data on the filler heat conductivity and the results of tests and application of the fuel elements imitators at various testing facilities are presented. The possibility of equipping one of the non operating WWER reactors with the fuel element imitators for conduct of large-scale experiment is indicated

Transient heat conduction in a pebble fuel applying fractional model

International Nuclear Information System (INIS)

Gomez A, R.; Espinosa P, G.

2009-10-01

In this paper we presents the equation of thermal diffusion of temporary-fractional order in the one-dimensional space in spherical coordinates, with the objective to analyze the heat transference between the fuel and coolant in a fuel element of a Pebble Bed Modular Reactor. The pebble fuel is the heterogeneous system made by microsphere constitutes by U O, pyrolytic carbon and silicon carbide mixed with graphite. To describe the heat transfer phenomena in the pebble fuel we applied a constitutive law fractional (Non-Fourier) in order to analyze the behaviour transient of the temperature distribution in the pebble fuel with anomalous thermal diffusion effects a numerical model is developed. (Author)

Process for the production of fuel combined articles for addition in block shaped high temperature fuel elements

International Nuclear Information System (INIS)

Hrovat, M.; Rachor, L.

1976-01-01

There is provided a process for the production of fuel compacts consisting of an isotropic, radiation-resistant graphite matrix of good heat conductivity having embedded therein coated fuel and/or fertile particles for insertion into high temperature fuel elements by providing the coated fuel and/or fertile particles with an overcoat of molding mixture consisting of graphite powder and a thermoplastic resin binder. The particles after the overcoating are provided with hardener and lubricant only on the surface and subsequently are compressed in a die heated to a constant temperature of about 150 0 C, hardened and discharged therefrom as finished compacts

Validation of structural design of JHR fuel element

International Nuclear Information System (INIS)

Brisson, S.; Miras, G.; Le Bourdonnec, L.; Lemoine, P.; Anselmet, M.C.; Marelle, V.

2010-01-01

The validation of the structural design of the Jules Horowitz Reactor fuel element was made by the Finite Element Method, starting from the Computer Aided Design. The JHR fuel element is a cylindrical assembly of three sectors composed of eight rolled fuel plates. A roll-swaging process is used to join the fuel plates to three aluminium stiffeners. The hydraulic gap between each plate is 1.95 mm. The JHR fuel assembly is fastened at both ends to the upper and lower endfittings by riveting. The main stresses are essentially thermal loads, imposed on the fuel zone of the plates. These thermal loads result from the nuclear heat flux (W/cm 2 ). The mechanical loads are mainly hydraulic thrust forces. The average coolant velocity is 15 m/s. Seismic effects are also studied. The fuel assembly is entirely modelled by thin shells. The model takes into account asymmetric thermal loads which often appear in Research Reactors. The mechanics of the fuel plates vary in function of the burn up. These mechanical properties are derived from the data sets used in the MAIA code, and the validity of the structure is demonstrable at throughout the life of the fuel. Results concerning displacement are compared to functional criteria, while results concerning stress are compared to RCC-MX criteria. The results of this analysis show that the mechanical and geometrical integrity of the JHR fuel elements is respected for Operating Categories 1 and 2. This paper presents the methodology of this demonstration for the results obtained. (author)

Once-through reactor with smooth hollow fuel elements for a manoeuvrable nuclear power plant

International Nuclear Information System (INIS)

Smolin, V.N.; Esikov, V.I.; Mityaev, Yu.I.; Vasil'ev, S.A.

1980-01-01

Studied are results of experimental investigation of the top capacity of hollow fuel element imitators, which has been carried out to substantiate possibilities of creation of nuclear energetic once-through type installation with subcritical parameters of heat carrier and smooth hollow fuel elements designed for operation in the regulatory regime of loading. Principal technological scheme of such NPP and approximate calculation characteristics of the reactor are presented. The investigation has been made using closed circular installation. Electro-heated tubes out of corrosion-resistant steel with the wall thickness variable as to the length are used as fuel element imitators. The experiments have been conducted using imitators with inner diameter of 14.6; 18 and 22 mm and the length of 6000 mm under the pressure (P) from 7.85 to 15.7 MPa, mass rate from 250 to 2000 kg/(m 2 xs); heat-carrier temperature at the entrance from 80 deg C up to the temperature 10-15 deg C lower than saturation temperature under the given pressure. The experiments have shown that under the pressure of 7.85-9.8 MPa in the heat carrier mass rates of 250-350 kg/(cm 2 xs) deteriorated regime of heat output appears only balance mass vapor content somewhat exceeding 1. It follows from the analysis of experimental data that the following parameters of heat-carrier are optimum: (from the viewpoint of maximum capacity in crisis-free regime of fuel element work) the pressure of 6-9 MPa; mass rate - 250-300 kg/(m 2 xs) at fuel element inner diameter of 25-20 mm respectively. Conclusion is made, that the investigations conducted create precondition for manoeuvrable NPP with once-through reactor

Some aspects of statistic evaluation of fast reactor fuel element reliability

International Nuclear Information System (INIS)

Proshkin, A.A.; Likhachev, Yu.I.; Tuzov, A.N.; Zabud'ko, L.M.

1980-01-01

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)O 2 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

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

Energy Technology Data Exchange (ETDEWEB)

Spasojevic, D; Vehauc, A [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

1969-06-15

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,{phi}); the thermal conductivity of the fuel and cladding material dependent on temperature, {lambda} = {lambda}(t), {lambda}{sub k} = {lambda}{sub k} (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)

Heat transfer investigations within dry spent fuel casks

International Nuclear Information System (INIS)

Nitsche, F.

1986-07-01

For studying the heat transfer processes and predicting the maximum spent fuel element surface temperature in a spent fuel assembly (SFA) transported in a dry cask, model experiments have been performed with a gas-filled model cask containing a simplified electrically heated model of a WWER-type SFA with 90 fuel elements. The temperature distribution of the SFA model is measured for different heat rates under vacuum in the model cask, and under normal pressure and overpressure (0.1 ... 0.7 MPa) for several cooling gases (air, argon, helium) in order to separately investigate heat transfer processes by radiation and convection/conduction. The measuring results were compared with the calculations. Computer programmes as well as simplified calculation methods for temperature prediction were developed and checked. The results obtained are also useful for thermal analyses in the field of the dry storage of SFAs in a cask or can. Specifically it was found that: The heat removal from the SFA can be considerably improved by increasing the internal cask pressure or by using helium as coolant. The radiant heat exchange in the SFA model can be calculated with sufficient accuracy by means of a computer programme developed in 1978 or by means of a simplified analytical representation shown in the final report. Both methods are directly applicable to the original SFA and useful in order to approximately calculate the maximum SFE surface temperature under normal pressure, if the fraction of heat transferred by radiation is allowed for. For the calculation of the total heat transfer a computer programme was developed and verified, which completely permits the temperature prediction of the SFA model in dependence on heat rate, type of gaseous coolant and coolant pressure. This computer programme can be directly applied to the original SFA for the calculation of the maximum SFE surface temperature

Prediction of the thermal behavior of a particle spherical fuel element using GITT

International Nuclear Information System (INIS)

Pessoa, C.V.; Oliveira, Claudio L. de; Jian, Su

2008-01-01

In this work, the transient and steady state heat conduction in a spherical fuel element of a pebble-bed high temperature were studied. This pebble element is composed by a particulate region with spherical inclusions, the fuel UO 2 particles, dispersed in a graphite matrix. A convective heat transfer by helium occurs on the outer surface of the fuel element. The two-energy equation model for the case of pure conduction was applied to this particulate spherical element, generating two macroscopic temperatures, respectively, of the inclusions and of the matrix. The transient analysis was carried out by using the Generalized Integral Transform Technique (GITT) that requires low computational efforts and allows a fast evaluation of the two macroscopic transient temperatures of the particulate region. The solution by GITT leads to a system of ordinary differential equations with the unknown transformed potentials. The mechanical properties (thermal conductivity and specific heat) of the materials were supposed not to depend on the temperature and to be uniform in each region. (author)

Laser pulse heating of nuclear fuels for simulation of reactor power

Indian Academy of Sciences (India)

Laser applications; nuclear fuel elements; nuclear safety. ... accident (LOCA) and reactivity initiated accident (RIA), a laser pulse heating system is under ... As a prelude to work on irradiated nuclear fuel specimens, pilot studies on unirradiated ...

Fuel coolant interaction experiment by direct electrical heating method

International Nuclear Information System (INIS)

Takeda, Tsuneo; Hirano, Kenmei

1979-01-01

In the PCM (Power Cooling Mismatch) experiments, the FCI (Fuel Coolant Interaction) test is one of necessary tests in order to predict various phenomena that occur during PCM in the core. A direct electrical heating method is used for the FCI tests for fuel pellet temperature of over 1000 0 C. Therefore, preheating is required before initiating the direct electrical heating. The fuel pin used in the FCI tests is typical LWR fuel element, which is surrounded by coolant water. It is undersirable to heat up the coolant water during preheating of the fuel pin. Therefore, a zirconia (ZrO 2 ) pellet which is similar to a UO 2 pellet in physical and chemical properties is used. Electric property (electric conductivity) of ZrO 2 is particularly suitable for direct electrical heating as in the case of UO 2 . In this experiment, ZrO 2 pellet (melting point 2500 0 C) melting was achieved by use of both preheating and direct electrical heating. Temperature changes of coolant and fuel surface, as well as the pressure change of coolant water, were measured. The molten fuel interacted with the coolant and generated shock waves. A portion of this molten fuel fragmented into small particles during this interaction. The peak pressure of the observed shock wave was about 35 bars. The damaged fuel pin was photographed after disassembly. This report shows the measured coolant pressure changes and the coolant temperature changes, as well as photographs of damaged fuel pin and fuel fragments. (author)

Development of 3D dynamic gap element for simulation of asymmetric fuel behavior

International Nuclear Information System (INIS)

Kim, Hyochan; Yang, Yongsik; Koo, Yanghyun; Kang, Changhak; Lee, Sunguk; Yang, Dongyol

2014-01-01

The accurate modeling of heat transfer across the gap between fuel pellets and the protective cladding is essential to understanding the fuel performance, including cladding stress and behavior under irradiated conditions. To establish a heat transfer model through a gap in the fuel performance code, the gap conductance based on the Ross and Stoute model was employed in most previous works. In this model, the gap conductance that determines the temperature gradient within the gap is a function of gap thickness, which is dependent on mechanical behavior. Recently, many researchers have been developing fuel performance codes based on the finite element method (FE) to calculate the temperature, stress, and strain in 2D or 3D. The gap conductance model for FE can be a challenging issue in terms of convergence and nonlinearity because the elements that are positioned in a gap have a different gap conductance, and the boundary conditions of the gap vary at each iteration step. In this paper, the specified 3D dynamic gap element has been proposed and implemented to simulate asymmetric thermo-mechanical fuel behavior. A thermo-mechanical 3D finite element module incorporating a gap element has been implemented using FORTRAN77. To evaluate the proposed 3D gap element, the missing pellet surface (MPS), which results in an asymmetric heat transfer in the pellet and cladding, was simulated. As a result, the maximum temperature of a pellet for the MPS problem calculated with the specified 3D gap element is much higher than the temperature calculated with a uniform gap conductance model that a multidimensional fuel performance code employs. The results demonstrate that a 3D simulation is essential to evaluate the temperature and stress of the pellet and cladding for an asymmetric geometry simulation. (author)

The advanced neutron source three-element-core fuel grading

International Nuclear Information System (INIS)

Gehin, J.C.

1995-01-01

The proposed Advanced Neutron Source (ANS) pre-conceptual design consists of a two-element 330 MW f nuclear reactor fueled with highly-enriched uranium and is cooled, moderated, and reflected with heavy water. Recently, the ANS design has been changed to a three-element configuration in order to permit a reduction of the enrichment, if required, while maintaining or improving the thermal-hydraulic margins. The core consists of three annular fuel elements composed of involute-shaped fuel plates. Each fuel plate has a thickness of 1.27 mm and consists of a fuel meat region Of U 3 Si 2 -Al (50% enriched in one case that was proposed) and an aluminum filler region between aluminum cladding. The individual plates are separated by a 1.27 mm coolant channel. The three element core has a fuel loading of 31 kg of 235 U which is sufficient for a 17-day fuel cycle. The goal in obtaining a new fuel grading is to maximize important temperature margins. The limits imposed axe: (1) Limit the temperature drop over the cladding oxide layer to less than 119 degrees C to avoid oxide spallation. (2) Limit the fuel centerline temperature to less than 400 degrees C to avoid fuel damage. (3) Limit the cladding wall temperature to less than the coolant. incipient-boiling temperature to avoid coolant boiling. Other thermal hydraulic conditions, such as critical heat flux, are also considered

The role of a fuel element and its cladding in water cooled reactor dynamics

International Nuclear Information System (INIS)

Randles, J.

1963-10-01

To clarify the role of fuel element cladding in water reactor dynamics, the heat diffusion and transfer equations are solved in slab geometry for (a) an oscillatory fission power, (b) an oscillatory coolant temperature. From the resulting transfer functions a clear description of the effect of the cladding on the heat flow is obtained. A Mercury autocode programme for evaluating the transfer functions is described. In addition to the slab element, the heat diffusion equations are also solved for a cylindrical system exposed to an oscillatory fission power. The solutions are expressed as transfer functions and are obtainable numerically from another autocode programme. Both of these programmes are used to obtain the power out/ power in transfer function for a typical cylindrical and slab UO 2 fuel pellet clad in zircaloy. The results give a further indication of the effect of the cladding heat capacity over a wide frequency range. It is shown also that the effect of the geometrical difference between a slab and cylindrical fuel element is unimportant provided the surface to volume ratio of the fuel is the same in each case. (author)

The role of a fuel element and its cladding in water cooled reactor dynamics

Energy Technology Data Exchange (ETDEWEB)

Randles, J [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1963-10-15

To clarify the role of fuel element cladding in water reactor dynamics, the heat diffusion and transfer equations are solved in slab geometry for (a) an oscillatory fission power, (b) an oscillatory coolant temperature. From the resulting transfer functions a clear description of the effect of the cladding on the heat flow is obtained. A Mercury autocode programme for evaluating the transfer functions is described. In addition to the slab element, the heat diffusion equations are also solved for a cylindrical system exposed to an oscillatory fission power. The solutions are expressed as transfer functions and are obtainable numerically from another autocode programme. Both of these programmes are used to obtain the power out/ power in transfer function for a typical cylindrical and slab UO{sub 2} fuel pellet clad in zircaloy. The results give a further indication of the effect of the cladding heat capacity over a wide frequency range. It is shown also that the effect of the geometrical difference between a slab and cylindrical fuel element is unimportant provided the surface to volume ratio of the fuel is the same in each case. (author)

Nuclear fuel element

International Nuclear Information System (INIS)

Grossman, L.N.; Levin, H.A.

1975-01-01

A nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smallr than the size of the particles. The container is preferably held in the spring in the plenum of the fuel element. (E.C.B.)

Predicted HIFAR fuel element temperatures for postulated loss-of-coolant accidents

International Nuclear Information System (INIS)

Green, W.J.

1987-04-01

A two-dimensional theoretical heat transfer model of a HIFAR Mark IV/Va fuel element has been developed and validated by comparing predicted thermal performances with experimental temperature responses obtained from irradiated fuel elements during simulated accident conditions. Full details of the model's development and its verification have been reported elsewhere. In this report, the model has been further used to ascertain acceptable limits of fuel element decay power for the start of two specific LOCAs which have been identified by the Regulatory Bureau of the AAEC. For a single fuel element which is positioned within a fuel load/unload flask and is not subjected to any forced convective air cooling, the model indicates that fission product decay powers must not exceed 1.86 kW if fuel surface temperatures are not to exceed 450 deg C. In the case of a HIFAR core LOCA in which the complete inventory of heavy water is lost, it is calculated that the maximum fission product decay power of a central element must not exceed 1.1 kW if fuel surface temperatures are not to exceed 450 deg C anywhere in the core

Prospect of Uranium Silicide fuel element with hypostoichiometric (Si ≤3.7%)

International Nuclear Information System (INIS)

Suripto, A.; Sardjono; Martoyo

1996-01-01

An attempt to obtain high uranium-loading in silicide dispersion fuel element using the fabrication technology applicable nowadays can reach Uranium-loading slightly above 5 gU/cm 3 . It is difficult to achieve a higher uranium-loading than that because of fabricability constraints. To overcome those difficulties, the use of uranium silicide U 3 Si based is considered. The excess of U is obtained by synthesising U 3 Si 2 in Si-hypostoichiometric stage, without applying heat treatment to the ingot as it can generate undesired U 3 Si. The U U will react with the matrix to form U al x compound, that its pressure is tolerable. This experiment is to consider possibilities of employing the U 3 Si 2 as nuclear fuel element which have been performed by synthesising U 3 Si 2 -U with the composition of 3.7 % weigh and 3 % weigh U. The ingot was obtained and converted into powder form which then was fabricated into experimental plate nuclear fuel element. The interaction between free U and Al-matrix during heat-treatment is the rolling phase of the fuel element was observed. The study of the next phase will be conducted later

Operational requirements of spherical HTR fuel elements and their performance

International Nuclear Information System (INIS)

Roellig, K.; Theymann, W.

1985-01-01

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 MWel eq 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)

Storage ponds for fuel elements of nuclear reactors

International Nuclear Information System (INIS)

Kumpf, H.

1981-01-01

Heat exchangers are inserted in storage ponds for fuel elements of nuclear reactors, so that the heat to be removed is given up to an external coolant, without any radio-activity being emitted. The heat exchanger is a hollow body, which is connected to an air cooler, which works with a cooling circuit with natural circulation. A cooling pipe is enclosed in the hollow body, which forms a cooling circuit with forced flow with an open pond. One therefore obtains two successive separating walls for the external coolant. (orig.) [de

Nuclear fuel elements

International Nuclear Information System (INIS)

Nakai, Keiichi

1983-01-01

Purpose: To decrease the tensile stresses resulted in a fuel can as well as prevent decladding of fuel pellets into the bore holes by decreasing the inner pressure within the nuclear fuel element. Constitution: A fuel can is filled with hollow fuel pellets, inserted with a spring for retaining the hollow fuel pellets with an appropriate force and, thereafter, closely sealed at the both ends with end plugs. A cylindrical body is disposed into the bore holes of the hollow fuel pellets. Since initial sealing gases and/or gaseous nuclear fission products can thus be excluded from the bore holes where the temperature is at the highest level, the inner pressure of the nuclear fuel element can be reduced to decrease the tensile strength resulted to the fuel can. Furthermore, decladding of fuel pellets into the bore holes can be prevented. (Moriyama, K.)

Nuclear fuel element

International Nuclear Information System (INIS)

Hirayama, Satoshi; Kawada, Toshiyuki; Matsuzaki, Masayoshi.

1980-01-01

Purpose: To provide a fuel element for reducing the mechanical interactions between a fuel-cladding tube and the fuel element and for alleviating the limits of the operating conditions of a reactor. Constitution: A fuel element having mainly uranium dioxide consists of a cylindrical outer pellet and cylindrical inner pellet inserted into the outer pellet. The outer pellet contains two or more additives selected from aluminium oxide, beryllium oxide, magnesium oxide, silicon oxide, sodium oxide, phosphorus oxide, calcium oxide and iron oxide, and the inner pellet contains nuclear fuel substance solely or one additive selected from calcium oxide, silicon oxide, aluminium oxide, magnesium oxide, zirconium oxide and iron oxide. The outer pellet of the fuel thus constituted is reduced in mechanical strength and also in the mechanical interactions with the cladding tube, and the plastic fluidity of the entire pellet is prevented by the inner pellet increased in the mechanical strength. (Kamimura, M.)

Process and equipment for locating defective fuel rods of a reactor fuel element

International Nuclear Information System (INIS)

Jester, A.; Honig, H.

1977-01-01

By this equipment, well-known processes for determining defective fuel rods of a reactor fuel element are improved in such a fashion that defective fuel rods can be located individually, so that it is possible to replace them. The equipment consists of a cylindrical test vessel open above, which accommodates the element to be tested, so that an annular space is left between the latter's external circumference and the wall of the vessel, and so that the fuel rods project above the vessel. A bell in the shape of a frustrum of a cone is inverted over the test vessel, which has an infra-red measuring equipment at a certain distance above the tops of the fuel rods. The fuel element to be tested together with the test vessel and hood are immersed in a basin full of water, which displaces water by means of gas from the hood. The post-shutdown heat increases the temperature in the water space of the test vessel, which is stabilised at 100 0 C. In each defective fuel rod the water which has penetrated the defective fuel rod previously, or does so now, starts to boil. The steam rising in the fuel rod raises the temperature of the defective fuel rod compared to all the sound ones. The subsequent measurement easily determines this. Where one can expect interference with the measurement by appreciable amounts of gamma rays, the measuring equipment is removed from the path of radiation by mirror deflection in a suitably shaped measuring hood. (FW) [de

Critical heat flux tests for a 12 finned-element assembly

Energy Technology Data Exchange (ETDEWEB)

Yang, J., E-mail: Jun.Yang@cnl.ca; Groeneveld, D.C.; Yuan, L.Q.

2017-03-15

Highlights: • CHF tests for a 12 finned-fuel-element assembly at highly subcooled conditions. • Test approach to maximize experimental information and minimize heater failures. • Three series of tests were completed in vertical upward light water flow. • Bundle simulators of two axial power profiles and three heated lengths were tested. • Results confirm that the prediction method predicts lower CHF values than measured. - Abstract: An experimental study was undertaken to provide relevant data to validate the current critical heat flux (CHF) prediction method of the NRU driver fuel for safety analysis, i.e., to confirm no CHF occurrence below the predicted values. The NRU driver fuel assembly consists of twelve finned fuel elements arranged in two rings – three in the inner ring and nine in the outer ring. To satisfy the experimental objective tests at very high heat fluxes, very high mass velocities, and high subcoolings were conducted where the CHF mechanism is the departure from nucleate boiling (DNB). Such a CHF experiment can be very difficult, costly and time consuming since failure of the heating surface due to rupture or melting (physical burnout) is expected when the DNB type of CHF is reached. A novel experimental approach has been developed to maximize the amount of relevant experimental information on safe operating conditions in the tests, and to minimize any possible heater failures that inherently accompany the CHF occurrence at these conditions. Three series of tests using electrically heated NRU driver fuel simulators with three heated lengths and two axial power profiles (or axial heat flux distribution (AFD)) were completed in vertical upward light water flow. Each series of tests covered two mass flow rates, several heat flux levels, and local subcoolings that bound the ranges of interest for the analysis of postulated slow loss-of-regulation accident (LORA) and loss-of-flow accident (LOFA) scenarios. Tests for each mass flow rate of

Thermal behaviour of pressure tube under fully and partially voided heating conditions using 19 pin fuel element simulator

International Nuclear Information System (INIS)

Yadav, Ashwini K.; Kumar, Ravi; Gupta, Akhilesh; Chatterjee, B.; Mukhopadhya, D.; Lele, H.G.

2011-01-01

In a nuclear reactor temperature can rise drastically during LOCA due to failure of heat transportation system and subsequently leads to mechanical deformations like sagging, ballooning and breaching of pressure tube. To understand the phenomenon an experiment has been carried out using 19 pin fuel element simulator. Main purpose of the experiment was to trace temperature profiles over the pressure tube, calandria tube and clad tubes of 220 MWe Indian Pressurised Heavy Water Reactor (IPHWR). The symmetrical heating of pressure tube of 1 m length was done through resistance heating of 19 pins under 13.5 kW power using a rectifier and the variation of temperatures over the circumference of pressure tube (PT), calandria tube (CT) and clad tubes were measured. The sagging of pressure tube was initiated at 460 deg C temperature and highest temperature attained was 650 deg C. The highest temperature attained by clad tubes was 680 deg C (over outer ring) and heat is dissipated to calandria vessel mainly due to radiation and natural convection. Again to simulate partially voided conditions, asymmetrical heating of pressure was carried out by injecting 8 kW power to upper 8 pins of fuel simulator. A maximum temperature difference of 295 deg C was observed over the circumference of pressure tube which highlights the magnitude of thermal stresses and its role in breaching of pressure tube under partially voided conditions. Integrity of pressure tube was retained during both symmetrical and asymmetrical heatup conditions. (author)

Calculation of Distribution Dynamics of Inhomogeneous Temperature Field in Range of Fuel Elements by Using FreeFem++

Science.gov (United States)

Amosova, E. V.; Shishkin, A. V.

2017-11-01

This article introduces the result of studying the heat exchange in the fuel element of the nuclear reactor fuel magazine. Fuel assemblies are completed as a bundle of cylindrical fuel elements located at the tops of a regular triangle. Uneven distribution of fuel rods in a nuclear reactor’s core forms the inhomogeneity of temperature fields. This article describes the developed method for heat exchange calculation with the account for impact of an inhomogeneous temperature field on the thermal-physical properties of materials and unsteady effects. The acquired calculation results are used for evaluating the tolerable temperature levels in protective case materials.

A collapse mode of failure in powder-filled fuel elements

International Nuclear Information System (INIS)

Feraday, M.A.; Chalder, G.H.

1964-01-01

Two swaged fuel elements containing crushed, fused UO 2 powder were irradiated in a pressurized water loop at high heat ratings (∫Kdθ = 48 w/cm). The fuel elements were 2.0 cm in diameter and were sheathed in nickel-free Zircaloy--2 of 0.038 cm thickness. One element failed when the sheath ruptured at the top of a longitudinal ridge in the sheath after a burn-up of approximately 2550 MWd/TeU. No evidence was found that outgassing of the UO 2 contributed to the failure. Dimensional and structural changes observed in the fuel elements led to the conclusion that ridging of the sheath resulted from the action of coolant pressure on the diametral clearance formed by sintering and shrinkage of the UO 2 . Failure resulted due to severe local deformation accompanying one or more power cycles following ridge formation. (author)

Analysis of the nine-point finite difference approximation for the heat conduction equation in a nuclear fuel element

International Nuclear Information System (INIS)

Kadri, M.

1983-01-01

The time dependent heat conduction equation in the x-y Cartesian geometry is formulated in terms of a nine-point finite difference relation using a Taylor series expansion technique. The accuracy of the nine-point formulation over the five-point formulation has been tested and evaluated for various reactor fuel-cladding plate configurations using a computer program. The results have been checked against analytical solutions for various model problems. The following cases were considered in the steady-state condition: (a) The thermal conductivity and the heat generation were uniform. (b) The thermal conductivity was constant, the heat generation variable. (c) The thermal conductivity varied linearly with the temperature, the heat generation was uniform. (d) Both thermal conductivity and heat generation vary. In case (a), approximately, for the same accuracy, 85% fewer grid points were needed for the nine-point relation which has a 14% higher convergence rate as compared to the five-point relation. In case (b), on the average, 84% fewer grid points were needed for the nine-point relation which has a 65% higher convergence rate as compared to the five-point relation. In case (c) and (d), there is significant accuracy (91% higher than the five-point relation) for the nine-point relation when a worse grid was used. The numerical solution of the nine-point formula in the time dependent case was also more accurate and converges faster than the numerical solution of the five-point formula for all comparative tests related to heat conduction problems in a nuclear fuel element

Review of fuel element development for nuclear rocket engines

International Nuclear Information System (INIS)

Taub, J.M.

1975-06-01

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 2500 0 C. 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 UO 2 , pyrocarbon-coated UC 2 , 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.)

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

Fluid pressure method for recovering fuel pellets from nuclear fuel elements

International Nuclear Information System (INIS)

John, C.D. Jr.

1979-01-01

A method is described for removing fuel pellets from a nuclear fuel element without damaging the fuel pellets or fuel element sheath so that both may be reused. The method comprises holding the fuel element while a high pressure stream internally pressurizes the fuel element to expand the fuel element sheath away from the fuel pellets therein so that the fuel pellets may be easily removed

Measuring heat transfer through TR-0 reactor fuel element

International Nuclear Information System (INIS)

Nemec, V.; Turzik, Z.; Vitek, M.

1977-05-01

The time course of temperatures of the peripheral and the central fuel pins of the TR-O reactor was studied during moderator temperature changes using a model. The formula T=Tsub(e)+(Tsub(o)-Tsub(e)).exp(-t/tsub(e)) applies, where T is the pin temperature, Tsub(o) the initial pin temperature, Tsub(e) is the steady-state bath temperature, tsub(e) the time constant of temperature equilibration and t the time required for a temperature change from value Tsub(o) to T. For the bath level height H=1 m the tsub(e) value for the central pin was determined to be 1.05 hours, for the peripheral pin 0.96 hour; for level height H=2 m the values were 2.1 and 2.12 hours, respectively. The dependence found will allow correcting the experimental results in measurements with heated moderator for fuel temperature changes. (Ha)

Fuel element

International Nuclear Information System (INIS)

Armijo, J.S.

1976-01-01

A fuel element for nuclear reactors is proposed which has a higher corrosion resisting quality in reactor operations. The zirconium alloy coating around the fuel element (uranium or plutonium compound) has on its inside a protection layer of metal which is metallurgically bound to the substance of the coating. As materials are namned: Alluminium, copper, niobium, stainless steel, and iron. This protective metallic layer has another inner layer, also metallurgically bound to its surface, which consists usually of a zirconium alloy. (UWI) [de

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

Science.gov (United States)

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

Fuel element for high-temperature nuclear power reactors

International Nuclear Information System (INIS)

Schloesser, J.

1974-01-01

The fuel element of the HTGR consists of a spherical graphite body with a spherical cavity. A deposit of fissile material, e.g. coated particles of uranium carbide, is fixed to the inner wall using binders. In addition to the fissile material, there are concentric deposits of fertile material, e.g. coated thorium carbide particles. The remaining cavity is filled with a graphite mass, preferably graphite powder, and the filling opening with a graphite stopper. At the beginning of the reactor operation, the fissile material layer provides the whole power. With progressing burn-up, the energy production is taken over by the fertile layer, which provides the heat production until the end of burn-up. Due to the relatively small temperature difference between the outer wall of the outer graphite body and the maximum fuel temperature, the power of the fuel element can be increased. (DG) [de

Hydrogen in CANDU fuel elements

International Nuclear Information System (INIS)

Sejnoha, R.; Manzer, A.M.; Surette, B.A.

1995-01-01

Unirradiated and irradiated CANDU fuel cladding was tested to compare the role of stress-corrosion cracking and of hydrogen in the development of fuel defects. The results of the tests are compared with information on fuel performance in-reactor. The role of hydriding (deuteriding) from the coolant and from the fuel element inside is discussed, and the control of 'hydrogen gas' content in the element is confirmed as essential for defect-free fuel performance. Finally, implications for fuel element design are discussed. (author)

Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

International Nuclear Information System (INIS)

Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

1992-11-01

This report discusses three furnace heating tests which were conducted with irradiated, HT9-clad and U-19wt.%Pu-l0wt.%Zr-alloy fuel, Mk-V-type fuel elements in the Alpha-Gamma Hot Cell Facility at Argonne National Laboratory, Illinois. In general, very significant safety margins for fuel-element cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results will be given, as well as discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction found in high-temperature testing of irradiated metallic fuel elements

Technique for mass-spectrometric determination of moisture content in fuel elements and fuel element claddings

International Nuclear Information System (INIS)

Kurillovich, A.N.; Pimonov, Yu.I.; Biryukov, A.S.

1988-01-01

A technique for mass-spectroimetric determination of moisture content in fuel elements and fuek claddings in the 2x10 -4 -1.5x10 -2 g range is developed. The relative standard deviation is 0.13. A character of moisture extraction from oxide uranium fuels in the 20-700 deg C temperature range is studied. Approximately 80% of moisture is extracted from the fuels at 300 deg C. The moisture content in fuel elements with granular uranium oxide fuels is measured. Dependence of fuel element moisture content on conditions of hot vacuum drying is shown. The technique permits to optimize the fuel element fabrication process to decrease the moisture content in them. 4 refs.; 3 figs.; 2 tabs

Rack for nuclear fuel elements

International Nuclear Information System (INIS)

Rubinstein, H.J.; Gordon, C.B.; Robison, A.; Clark, P.M.

1977-01-01

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

Fuel elements (uranium clad with zirconium) produced by co-extrusion; Les elements combustibles uranium gaine de zirconium obtenus par coextrusion

Energy Technology Data Exchange (ETDEWEB)

Montagne, R.; Winogradzki, A.; Sauve, C.; Buffet, J. [Commissariat a l' energie atomique et aux energies alternatives - CEA, Departement de Metallurgie et de Chimie Appliquee (France)

1959-07-01

In this paper a description is given of a process for making fuel elements for atomic reactors. Contact of the most intimate possible kind is achieved between the fuel and the sheath by the simultaneous extrusion of the two elements. Genuine welding between the two metals is thus effected. This can be subsequently improved by a heat-treatment which causes diffusion. Tests made on these co-extruded elements are described in the paper. Reprint of a paper published in Revue de Metallurgie, LV, no. 11, 1958.

Heat transfer from the roughened surface of gas cooled fast breeder reactor fuel element

International Nuclear Information System (INIS)

Tang, I.M.

1979-01-01

The temperature distributions and the augmentation of heat transfer performance by artificial roughening of a gas cooled fast breeder reactor (GCFR) fuel rod cladding are studied. Numerical solutions are based on the axisymmetric assumption for a two-dimensional model for one rib pitch of axial distance. The local and axial clad temperature distributions are obtained for both the rectangular and ramp rib roughened surface geometries. The transformation of experimentally measured convective heat transfer coefficients, in terms of Stanton number, into GCFR values is studied. In addition, the heat transfer performance of a GCFR fuel rod cladding roughened surface design is evaluated. Approximate analytical solution for correlating an average Stanton number is also obtained and satisfactorily compared with the corresponding numerical result for a GCFR design. The analytical correlation is useful in assessing roughened surface heat transfer performance in scoping studies and conceptual design

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

International Nuclear Information System (INIS)

Nastase, D.; Olteanu, G.; Ioan, M.; Pauna, E.

2013-01-01

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)

Fuel element

International Nuclear Information System (INIS)

Kennedy, S.T.

1982-01-01

A nuclear reactor fuel element wherein a stack of nuclear fuel is prevented from displacement within its sheath by a retainer comprising a tube member which is radially expanded into frictional contact with the sheath by means of a captive ball within a tapered bore. (author)

Influence of fuel-cladding system deviations from the model of continuous cylinders on the parameters of WWER fuel element working ability

International Nuclear Information System (INIS)

Scheglov, A.

1994-01-01

In the programs of fuel rod computation, fuel and cladding are usually presented in the form of coaxial cylinders, which can change their sizes, mechanical and thermal-physical properties. The real fuel element has some typical deviations from this continuous coaxial cylinders (CCC) model as: axial asymmetry of fuel-cladding system (due to the oval form of the cladding, cracking and other type of fuel pallet damage, axial asymmetry of the volumetric heat release), gaps between the pallets (and heat release peaking in fuel near the gap), chambers in the pallets. As a result of these deviations actual fuel rod parameters of working ability - temperature, stresses, thermal fluxes relieved from the cladding, geometry changes - in some locations can greatly vary from the ones calculated according to CCC model. The influence of these deviations is extremely important while calculating the fuel rod, because they are a part of the mechanical excess coefficient. The author reviews the influence of these factors using specific examples. He applies his own two-dimensional codes based on the Finite Elements Method for calculations of temperature fields, stresses and deformation in the fuel rod elements. It is shown that consideration of these deviations, as a rule, leads to the increase of the maximum fuel temperature in the WWER pellets (characterized by a large central hole), temperature of the cladding, thermal flux, relieved by the coolant from the cladding, and stresses in the cladding. It is necessary to consider these factors for both validation of the fuel element working ability and interpretation of the experimental results. 4 tabs., 3 figs., 5 refs

Influence of fuel-cladding system deviations from the model of continuous cylinders on the parameters of WWER fuel element working ability

Energy Technology Data Exchange (ETDEWEB)

Scheglov, A [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

1994-12-31

In the programs of fuel rod computation, fuel and cladding are usually presented in the form of coaxial cylinders, which can change their sizes, mechanical and thermal-physical properties. The real fuel element has some typical deviations from this continuous coaxial cylinders (CCC) model as: axial asymmetry of fuel-cladding system (due to the oval form of the cladding, cracking and other type of fuel pallet damage, axial asymmetry of the volumetric heat release), gaps between the pallets (and heat release peaking in fuel near the gap), chambers in the pallets. As a result of these deviations actual fuel rod parameters of working ability - temperature, stresses, thermal fluxes relieved from the cladding, geometry changes - in some locations can greatly vary from the ones calculated according to CCC model. The influence of these deviations is extremely important while calculating the fuel rod, because they are a part of the mechanical excess coefficient. The author reviews the influence of these factors using specific examples. He applies his own two-dimensional codes based on the Finite Elements Method for calculations of temperature fields, stresses and deformation in the fuel rod elements. It is shown that consideration of these deviations, as a rule, leads to the increase of the maximum fuel temperature in the WWER pellets (characterized by a large central hole), temperature of the cladding, thermal flux, relieved by the coolant from the cladding, and stresses in the cladding. It is necessary to consider these factors for both validation of the fuel element working ability and interpretation of the experimental results. 4 tabs., 3 figs., 5 refs.

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

International Nuclear Information System (INIS)

Bae, Jun Ho; Rho, Gyu Hong; Park, Joo Hwan

2009-10-01

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

Nuclear fuel element

International Nuclear Information System (INIS)

Mogard, J.H.

1977-01-01

A nuclear fuel element is disclosed for use in power producing nuclear reactors, comprising a plurality of axially aligned ceramic cylindrical fuel bodies of the sintered type, and a cladding tube of metal or metal alloys, wherein said cladding tube on its cylindrical inner surface is provided with a plurality of slightly protruding spacing elements distributed over said inner surface

Conjugate heat transfer simulations of advanced research reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Piro, M.H.A., E-mail: pirom@aecl.ca; Leitch, B.W.

2014-07-01

Highlights: • Temperature predictions are enhanced by coupling heat transfer in solid and fluid zones. • Seven different cases are considered to observe trends in predicted temperature and pressure. • The seven cases consider high/medium/low power, flow, burnup, fuel material and geometry. • Simulations provide temperature predictions for performance/safety. Boiling is unlikely. • Simulations demonstrate that a candidate geometry can enhance performance/safety. - Abstract: The current work presents numerical simulations of coupled fluid flow and heat transfer of advanced U–Mo/Al and U–Mo/Mg research reactor fuels in support of performance and safety analyses. The objective of this study is to enhance predictions of the flow regime and fuel temperatures through high fidelity simulations that better capture various heat transfer pathways and with a more realistic geometric representation of the fuel assembly in comparison to previous efforts. Specifically, thermal conduction, convection and radiation mechanisms are conjugated between the solid and fluid regions. Also, a complete fuel element assembly is represented in three dimensional space, permitting fluid flow and heat transfer to be simulated across the entire domain. Seven case studies are examined that vary the coolant inlet conditions, specific power, and burnup to investigate the predicted changes in the pressure drop in the coolant and the fuel, clad and coolant temperatures. In addition, an alternate fuel geometry is considered with helical fins (replacing straight fins in the existing design) to investigate the relative changes in predicted fluid and solid temperatures. Numerical simulations predict that the clad temperature is sensitive to changes in the thermal boundary layer in the coolant, particularly in simultaneously developing flow regions, while the temperature in the fuel is anticipated to be unaffected. Finally, heat transfer between fluid and solid regions is enhanced with

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

International Nuclear Information System (INIS)

Bowditch, F.H.; Mogford, D.J.

1987-02-01

This report describes experiments that have been carried out at the Winfrith Establishment of the United Kingdom Atomic Energy Authority to determine the critical heat flux characteristics of pressurized water reactor fuel elements over an unusually wide range of coolant flow conditions that are relevant to both normal and fault conditions of reactor operation. The experiments were carried out in the TITAN loop using an electrically heated bundle of 25 rods of 9.5 mm diameter on a 12.7 mm pitch fitted with plain grids in order to provide a generic base for code validation. The fully tabulated experimental data for critical heat flux, pressure drop and sub-channel mixing are encompassed by ranges of pressure between 20 and 160 Bar, coolant flow between 150 and 3600 Kg/m 2 s, and coolant inlet temperature between 150 and 320 0 C. The results of the experiments are compared with predicted data based upon several established critical heat flux correlations. It is concluded that the extrapolation of some correlations to conditions beyond their intended range of application can lead to dangerous over estimates of critical heat flux, but the Winfrith WSC-2 and the EPRI NP-2609 correlations perform well over the whole data range and correlate all data with RMS errors of 9% and 6% respectively. (author)

Fuel Element Experience at the Halden Boiling Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Aas, S. [OECD Halden Reactor Project, Halden (Norway); Videm, K.; Hanevik, A. [Institutt for Atomenergi, Kjeller (Norway)

1968-04-15

The penalty for neutron absorbing materials is higher for a reactor moderated with heavy water than one with light water. As Zircaloy and enriched uranium were not readily available in 1954 when the design of the first fuel charge for HBWR was frozen, fuel elements of natural uranium metal clad in a specially developed aluminium alloy (A 1 0.3% Fe, 0.03% Si) were used. The temperature was limited to 150 Degree-Sign C and with this limitation the general behaviour of the elements was good. In I960, in another effort to maintain a good neutron economy, a couple of elements with as thin cladding as 0.25 mm A1S1 316, stainless steel with an unsegmented length of 2 m supported by wire grid spacers were tested. These elements with 1.5% enriched UO{sub 2} behaved satisfactorily at 150'C. Elements of a rather similar construction failed due to stress corrosion during the later operation at 230 'C. The reason for the different behaviour is probably the higher stresses in the cladding, due to the increased pressure, possibly combined with a short period with a high chloride content in the heavy water. The second fuel core with 1.5% enriched UO{sub 2} clad in Zircaloy-2 was installed in order to permit an increase in temperature to 230 Degree-Sign C and in power from 5 to 20 MW(th). The maximum burnup obtained is 11000 MWd/t and the maximum heat rating 375 W/cm with no fracture failure and practically no change in appearance according to the post-irradiation examination. One element was deliberately taken to burn-out conditions by throttling the water flow. After a series of burn-outs, the element finally failed because of over-temperature. The successful use of aluminium cladding at 150 Degree-Sign C mitiated an effort for making aluminium alloys suitable for normal power reactor operation. Promising properties were found for an alloy (designated IFA 3 aluminium) with A1 10% Si, 1% Ni, 1% Mg, 0.3% Fe + Ti. Despite increase in corrosion rate under heat transfer conditions

Eutectic penetration times in irradiated EBR-II driver fuel elements

International Nuclear Information System (INIS)

Betten, P.R.; Bottcher, J.H.; Seidel, B.R.

1983-01-01

The experimental test procedure employed the use of a high-temperature furnace which heated pre-irradiated elements to temperature and maintained the environment until element-cladding breach occurred. Pre-irradiated elements of the Mark-II design were first encapsulated in a close-fitting sealed tube that was instrumented with a pressure transducer at the top of the tube and a thermocouple at the element's top-of-fuel axial location. The volume of the capsule was evacuated in order to better identify the pressure pulse which would occur on breach and to minimize contaminants. Next, a three-zone fast-recovery furnace was heated and an axial temperature profile, similar to that experienced in the EBR-II core, was established. The encapsulated element was then quickly inserted into the furnace and remained there until clad breach occurred. The element was then removed from the furnace immediately. Visual and metallurgical examination of the rupture site was done later. A total of seven elements were tested in the above manner

Monitoring arrangement for vented nuclear fuel elements

International Nuclear Information System (INIS)

Campana, R.J.

1981-01-01

In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180 0 rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements

Instrumentation of fuel elements and fuel plates

International Nuclear Information System (INIS)

Durand, J.P.; Fanjas, Y.

1993-01-01

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)

Instrumentation of fuel elements and fuel plates

International Nuclear Information System (INIS)

Durand, J.P.; Fanjas, Y.

1994-01-01

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)

Nuclear fuel element leak detection system

International Nuclear Information System (INIS)

John, C.D. Jr.

1978-01-01

Disclosed is a leak detection system integral with a wall of a building used to fabricate nuclear fuel elements for detecting radiation leakage from the nuclear fuel elements as the fuel elements exit the building. The leak detecting system comprises a shielded compartment constructed to withstand environmental hazards extending into a similarly constructed building and having sealed doors on both ends along with leak detecting apparatus connected to the compartment. The leak detecting system provides a system for removing a nuclear fuel element from its fabrication building while testing for radiation leaks in the fuel element

Strength analysis of fast gas cooled reactor fuel element in conditions of fuel-cladding interraction and non-uniform azimuthal heating

International Nuclear Information System (INIS)

Kulikov, I.S.; Tverkovkin, B.E.

1984-01-01

The technique and the PRORT mathematical program in FORTRAN language for determining mechanical properties of a fuel element with motionless fuel-cladding interaction taking into account circular temperature non-uniformity in gas-cooled fast reactor conditions are proposed. The calculation results of the fuel element of dissociating gas cooled fast reactor are presented for seven cross-sections over the height of the core. The obtained data testify to appreciable swelling of Cr16Ni15Mo3Nb steel fuel cladding in the conditions of dissociating gas cooled fast reactor through the allowance for the effect of stresses on this essential parameter shows, that its value is lower in comparison with swelling, wherein stresses are not taken into account

Nuclear Fuel elements

International Nuclear Information System (INIS)

Hirakawa, Hiromasa.

1979-01-01

Purpose: To reduce the stress gradient resulted in the fuel can in fuel rods adapted to control the axial power distribution by the combination of fuel pellets having different linear power densities. Constitution: In a fuel rod comprising a first fuel pellet of a relatively low linear power density and a second fuel pellet of a relatively high linear power density, the second fuel pellet is cut at its both end faces by an amount corresponding to the heat expansion of the pellet due to the difference in the linear power density to the adjacent first fuel pellet. Thus, the second fuel pellet takes a smaller space than the first fuel pellet in the fuel can. This can reduce the stress produced in the portion of the fuel can corresponding to the boundary between the adjacent fuel pellets. (Kawakami, Y.)

Spacer for supporting fuel element boxes

International Nuclear Information System (INIS)

Wild, E.

1979-01-01

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

Unified fuel elements development for research reactors

International Nuclear Information System (INIS)

Vatulin, A.; Stetsky, Y.; Dobrikova, I.

1998-01-01

Square cross-section rod type fuel elements have been developed for russian pool-type research reactors. new fuel elements can replace the large nomenclature of tubular fuel elements with around, square and hexahedral cross-sections and to solve a problem of enrichment reduction. the fuel assembly designs with rod type fuel elements have been developed. The overall dimensions of existing the assemblies are preserved in this one. the experimental-industrial fabricating process of fuel elements, based on a joint extrusion method has been developed. The fabricating process has been tested in laboratory conditions, 150 experimental fuel element samples of the various sizes were produced. (author)

Improved lumped models for transient combined convective and radiative cooling of a two-layer spherical fuel element

International Nuclear Information System (INIS)

Silva, Alice Cunha da; Su, Jian

2013-01-01

The High Temperature Gas cooled Reactor (HTGR) is a fourth generation thermal nuclear reactor, graphite-moderated and helium cooled. The HTGRs have important characteristics making essential the study of these reactors, as well as its fuel element. Examples of these are: high thermal efficiency,low operating costs and construction, passive safety attributes that allow implication of the respective plants. The Pebble Bed Modular Reactor (PBMR) is a HTGR with spherical fuel elements that named the reactor. This fuel element is composed by a particulate region with spherical inclusions, the fuel UO2 particles, dispersed in a graphite matrix and a convective heat transfer by Helium happens on the outer surface of the fuel element. In this work, the transient heat conduction in a spherical fuel element of a pebble-bed high temperature reactor was studied in a transient situation of combined convective and radiative cooling. Improved lumped parameter model was developed for the transient heat conduction in the two-layer composite sphere subjected to combined convective and radiative cooling. The improved lumped model was obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of the two-layer spherical fuel element was analyzed to illustrate the applicability of the proposed lumped model, with respect to die rent values of the Biot number, the radiation-conduction parameter, the dimensionless thermal contact resistance, the dimensionless inner diameter and coating thickness, and the dimensionless thermal conductivity. It was shown by comparison with numerical solution of the original distributed parameter model that the improved lumped model, with H2,1/H1,1/H0,0 approximation yielded significant improvement of average temperature prediction over the classical lumped model. (author)

Automated Fuel Element Closure Welding System

International Nuclear Information System (INIS)

Wahlquist, D.R.

1993-01-01

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

Research of heat releasing element of an active zone of gaseous nuclear reactor with pumped through nuclear fuel - uranium hexafluoride (UF6)

International Nuclear Information System (INIS)

Batyrbekov, G.; Batyrbekov, E.; Belyakova, E.; Kunakov, S.; Koltyshev, S.

1996-01-01

The purpose of the offered project is learning physics and substantiation of possibility of creation gaseous nuclear reactor with pumped through nuclear fuel-hexafluoride of uranium (Uf6).Main problems of this work are'. Determination of physic-chemical, spectral and optical properties of non-equilibrium nuclear - excited plasma of hexafluoride of uranium and its mixtures with other gases. Research of gas dynamics of laminar, non-mixing two-layer current of gases of hexafluoride of uranium and helium at availability and absence of internal energy release in hexafluoride of uranium with the purpose to determinate a possibility of isolation of hexafluoride of uranium from walls by inert helium. Creation and research of gaseous heat releasing element with pumped through fuel Uf6 in an active zone of research nuclear WWR-K reactor. Objects of a research: Non-equilibrium nuclear - excited plasma of hexafluoride of uranium and its mixtures with other gases. With use of specially created ampoules will come true in-reactor probe and spectral diagnostics of plasma. Calculations of kinetics with the account of main elementary processes proceeding in it, will be carried out. Two-layer non-mixed streams of hexafluoride of uranium and helium at availability and absence of internal energy release. Conditions of obtaining and characteristics of such streams will be investigated. Gaseous heat releasing element with pumped through fuel - Uf6 in an active zone of nuclear WWR-K reactor

Block fuel element for gas-cooled high temperature reactors

International Nuclear Information System (INIS)

Hrovat, M.F.

1978-01-01

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

Thermal design and analysis of the HTGR fuel element vertical carbonizing and annealing furnace

International Nuclear Information System (INIS)

Llewellyn, G.H.

1977-06-01

Computer analyses of the thermal design for the proposed HTGR fuel element vertical carbonizing and annealing furnace were performed to verify its capability and to determine the required power input and distribution. Although the furnace is designed for continuous operation, steady-state temperature distributions were obtained by assuming internal heat generation in the fuel elements to simulate their mass movement. The furnace thermal design, the analysis methods, and the results are discussed herein

Study on the Efficient Disintegration of HTGR Fuel Elements by Electrochemical Method

International Nuclear Information System (INIS)

Piao Nan; Chen Ji; Xiao Cuiping; We Mingfen; Che Jing

2014-01-01

The spent fuel elements in High- temperature gas-cooled reactor (HTGR) have a special structure, so the head-end process of the spent fuel reprocessing is different from the process of water reactor spent fuel. The first step of head-end process of the HTGR spent fuel reprocessing process is disintegration of the graphite matrix and separation of the coated fuel particles. Electrochemical method with nitrate solution as an electrolyte for fuel element disintegration has been conducted by the Institute of Nuclear and New Energy Technology in Tsinghua University. This method allows a total disintegration of graphite matrix, while still preserving the integrity of TRISO particles. The influences of the pretreatment methods such as heating oxidation of graphite, hydrothermal and oxidants oxidation were investigated in the present work. The experimental results showed that there were no significant effects on increasing the disintegration rate when pretreatment methods were used ahead of electrochemical disintegration. This phenomenon indicated that the fuel elements which were calcined at 1073 K and pressed under 300 MPa are too compact to be broken by these pretreatment methods. And the electrochemical disintegration is an effective but slow method in breaking the graphite matrix. (author)

Nuclear reactor fuel element assemblies

International Nuclear Information System (INIS)

Krawiec, D.M.; Bevilacqua, F.

1974-01-01

The fuel elements of each fuel element group are separated from each other by means of a multitude of thin, intersecting plates in the from of grid strips. Flow deflectors near the surface of the fuel elements are used in order to make the coolant flow more turbulent. They are designed as vanes and arranged at a distance on the grid strips. Each deflector vane has two arms stretching in opposite directions, each one into a neighbouring channel. In outward direction, the deflector vanes are converging. The strips with the vanes can be put on the supporting grid of the fuel elements. The vane structure can be reinforced by providing distortions in the strip material near the vanes. (DG) [de

Heating subsurface formations by oxidizing fuel on a fuel carrier

Science.gov (United States)

Costello, Michael; Vinegar, Harold J.

2012-10-02

A method of heating a portion of a subsurface formation includes drawing fuel on a fuel carrier through an opening formed in the formation. Oxidant is supplied to the fuel at one or more locations in the opening. The fuel is combusted with the oxidant to provide heat to the formation.

Method of measuring distance between fuel element

International Nuclear Information System (INIS)

Urata, Megumu.

1991-01-01

The distance between fuel elements contained in a pool is measured in a contactless manner even for a narrow distance less than 1 mm. That is, the equipment for measuring the distance between spent fuel elements of a spent fuel assembly in a nuclear reactor comprises a optical fiber scope, a lens, an industrial TV camera and a monitor TV. The top end of the optical fiber scope is inserted between fuel elements to be measured. The state thereof is displayed on the TV screen to measure the distance between the fuel elements. The measured results are compared with a previously formed calibration curve to determine the value between the fuel elements. Then, the distance between the fuel elements can be determined in the pool of a power plant without dismantling the fuel assembly, to investigate the state of the bending and estimate the fuel working life. (I.S.)

A Study of the Temperature Distribution in UO2 Reactor Fuel Elements

International Nuclear Information System (INIS)

Devold, I.

1968-05-01

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 UO 2 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

Increased burnup of fuel elements

International Nuclear Information System (INIS)

Ahlf, J.

1983-01-01

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.) [de

Pre-irradiation testing of experimental fuel elements

International Nuclear Information System (INIS)

Basova, B.G.; Davydov, E.F.; Dvoretskij, V.G.; Ivanov, V.B.; Syuzev, V.N.; Timofeev, G.A.; Tsykanov, V.A.

1979-01-01

The problems of testing of experimental fuel elements of nuclear reactors on the basis of complex accountancy of the factors defining operating capacity of the fuel elements are considered. The classification of the parameters under control and the methods of initial technological testing, including testing of the fuel product, cladding and fished fuel element, is given. The requirements to the apparatus used for complex testing are formulated. One of the possible variants of representation of the information obtained in the form of the input certificate of a single fuel element under study is proposed. The processing flowsheet of the gathered information using the computer is given. The approach under consideration is a methodological basis of investigation of fuel element operating life at the testing stage of the experimental fuel elements

Methodology for verification of heat transfer crisis in the nuclear fuel assemblies

International Nuclear Information System (INIS)

Sharaevsky, I. G.; Sharaevskaya, E. I.; Domashev, E. D.; Arkhypov, A. P.; Kolochko, V. N.

2003-01-01

Reliable operation of water-water type nuclear energy units and design of new generation reactors are not to be provided with wide application of best estimate ThermalHydraulic (TH) codes. It is accepted to consider that up-to-date versions of the codes are featured not only by wide range of NPPs equipment modeling and high ergonomic characteristics of realized in the codes interfaces but comprehensive substantiation of its governing component viz correlations and closure relations systems The pointed correlations and closure relations provide mathematical restraint of the main differential equations system which are necessary for adequate description of the main classes of two-phase flow TH regimes. The principal fact is that without physically justificated correlations and adequate closure relations first of all concerning heat transfer crisis at boiling (DNB) the acceptable reliability of numerical solutions cannot be guaranteed by the codes. But the significant part of realized in the codes correlations mainly on heat transfer crisis are based on the experimental data obtained more than 30 years ago for cylindrical channels. It is known that for TH reliability calculations of the WWERs core with rod fuel elements, such correlations can be applied with caution as it give significantly conservative values of critical heat flux especially at under pressure accident regimes. Moreover because of irregularity of the flow TH parameters on fuel rod elements cross-section distribution the heat transfer crisis regimes are originated only in separate 'hot' cells. Additionally it should be underlined that realized in the codes correlations and closure relations do not consider possibility occurring in the steam generating channels high frequency oscillation instability which poses a threat to the reactor safety. The high frequency oscillations can bring to the fuel elements destruction at heat fluxes much less than the critical ones. Now this type of oscillation

Fuel element transport container

International Nuclear Information System (INIS)

Benna, P.; Neuenfeldt, W.

1979-01-01

The reprocessing system includes a large number of waterfilled ponds next to each other for the intermediate storage of fuel elements from LWR's. The fuel element transport device is allocated to a middle pond. The individual ponds are separated from each other by walls, and are only accessible from the middle pond via narrow passages. The transport device includes a telescopic running rail for a trolley with a grab device for the fuel element. The running rail is supported in turn by a second trolley, which can be moved by wheels on rails. Part of the drive of the first trolley is arranged on the second one. Using this transport device, adjacent ponds can be served through the passage openings. (DG) [de

Thermal model of spent fuel transport cask

International Nuclear Information System (INIS)

Ahmed, E.E.M.; Rahman, F.A.; Sultan, G.F.; Khalil, E.E.

1996-01-01

The investigation provides a theoretical model to represent the thermal behaviour of the spent fuel elements when transported in a dry shipping cask under normal transport conditions. The heat transfer process in the spent fuel elements and within the cask are modeled which include the radiant heat transfer within the cask and the heat transfer by thermal conduction within the spent fuel element. The model considers the net radiant method for radiant heat transfer process from the inner most heated element to the surrounding spent elements. The heat conduction through fuel interior, fuel-clad interface and on clad surface are also presented. (author) 6 figs., 9 refs

Fuel element tomography by gammametry

International Nuclear Information System (INIS)

Simonet, G.; Pineira, T.

1982-03-01

As from transversal gamma determinations of a cylindrical fuel element, the TOMOGAM program reconstitutes the distribution of fission products in a section. This direct, fast and non destructive method, makes it possible to have access to the behaviour of the fuel at any time: - the soluble fission products in the matrix represent the fuel itself and the distribution of the fissions, - the migrating elements inform on the temperature reached in accordance with the permitted powers, - the volatile nuclides build up in particular points where physical-chemical phenomena of fuel-cladding interaction are liable to corrode the latter. Hence, gamma spectrometry extends its possibilities of analysis relative to the performance of reactor elements [fr

Nuclear fuels and development of nuclear fuel elements

International Nuclear Information System (INIS)

Sundaram, C.V.; Mannan, S.L.

1989-01-01

Safe, reliable and economic operation of nuclear fission reactors, the source of nuclear power at present, requires judicious choice, careful preparation and specialised fabrication procedures for fuels and fuel element structural materials. These aspects of nuclear fuels (uranium, plutonium and their oxides and carbides), fuel element technology and structural materials (aluminium, zircaloy, stainless steel etc.) are discussed with particular reference to research and power reactors in India, e.g. the DHRUVA research reactor at BARC, Trombay, the pressurised heavy water reactors (PHWR) at Rajasthan and Kalpakkam, and the Fast Breeder Test Reactor (FBTR) at Kalpakkam. Other reactors like the gas-cooled reactors operating in UK are also mentioned. Because of the limited uranium resources, India has opted for a three-stage nuclear power programme aimed at the ultimate utilization of her abundant thorium resources. The first phase consists of natural uranium dioxide-fuelled, heavy water-moderated and cooled PHWR. The second phase was initiated with the attainment of criticality in the FBTR at Kalpakkam. Fast Breeder Reactors (FBR) utilize the plutonium and uranium by-products of phase 1. Moreover, FBR can convert thorium into fissile 233 U. They produce more fuel than is consumed - hence, the name breeders. The fuel parameters of some of the operating or proposed fast reactors in the world are compared. FBTR is unique in the choice of mixed carbides of plutonium and uranium as fuel. Factors affecting the fuel element performance and life in various reactors e.g. hydriding of zircaloys, fuel pellet-cladding interaction etc. in PHWR and void swelling; irradiation creep and helium embrittlement of fuel element structural materials in FBR are discussed along with measures to overcome some of these problems. (author). 15 refs., 9 tabs., 23 figs

Nuclear reactor fuel element splitter

International Nuclear Information System (INIS)

Yeo, D.

1976-01-01

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

Burnup measurements of leader fuel elements

International Nuclear Information System (INIS)

Henriquez, C; Navarro, G; Pereda, C

2000-01-01

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

Device for manipulating a nuclear reactor fuel element in a fuel element pond containing water

International Nuclear Information System (INIS)

Jabsen, F.S.

1977-01-01

Using this device a fuel element can be manipulated inside a water filled storage pond for inspection purposes. A transport arrangement which is normally situated above such a pond is modified for this purpose. A crane bridge runs on rails on the upper edge of the pond. A type of trolley runs transversely to the direction of travel of the bridge between 2 wide flange supports forming the crane support. During movement this trolley moves a submerged combination of periscope and TV camera pendant from it at about half the pond height horizontally along the crane support. 2 vehicles move between these on 4 rollers each, on the under flanges of the crane support at spacings of about one fuel element length. A pendant arm of the same length as the periscope dips vertically into the pond from each vehicle. There is a bar of about fuel element length resting on the lower ends of both arms. The surface of a fuel element lying on this bar can be inspected through the periscope on longitudinal travel of the trolley. The bar with the fuel element can be rotated 90 0 downwards into a vertical position after removal of one or more rotating kingpins and release of a rope hanging on the end away from the kingpin. The rope is actuated by a winch on the crane support. The bar has vertical plates at both ends to hold the fuel element in its vertical position. (HP) [de

Study of fuel cell powerplant with heat recovery

Science.gov (United States)

King, J. M.; Grasso, A. P.; Clausi, J. V.

1975-01-01

It was shown that heat can be recovered from fuel cell power plants by replacing the air-cooled heat exchangers in present designs with units which transfer the heat to the integrated utility system. Energy availability for a 40-kW power plant was studied and showed that the total usable energy at rated power represents 84 percent of the fuel lower heating value. The effects of design variables on heat availability proved to be small. Design requirements were established for the heat recovery heat exchangers, including measurement of the characteristics of two candidate fuel cell coolants after exposure to fuel cell operating conditions. A heat exchanger test program was defined to assess fouling and other characteristics of fuel cell heat exchangers needed to confirm heat exchanger designs for heat recovery.

Heaters to simulate fuel pins for heat transfer tests in single-phase liquid-metal-flow

International Nuclear Information System (INIS)

Casal, V.; Graf, E.; Hartmann, W.

1976-09-01

The development of heaters for thermal simulation of the fuel elements of liquid metal cooled fast breeder reactors (SNR) is reported. Beginning with the experimental demands various heating methods are discussed for thermodynamic investigations of the heat transfer in liquid metals. Then a preferred heater rod is derived to simulate the fuel pins of a SNR. Finally it is reported on the fabrication and the operation practice. (orig.) [de

Study of fuel element characteristic of SM and SMP (SM-PRIMA) fuel assemblies

International Nuclear Information System (INIS)

Klinov, A.V.; Kuprienko, V.A.; Lebedev, V.A.; Makhin, V.M.; Tuchnin, L.M.; Tsykanov, V.A.

1999-01-01

The paper discusses the techniques and results of reactor tests and post-reactor investigations of the SM reactor fuel elements and fuel elements developed in the process of designing the specialized PRIMA test reactor with the SM reactor fuel elements used as a prototype and which are referred to as the SMP fuel elements. The behavior of fuel elements under normal operating conditions and under deviation from normal operating conditions was studied to verify the calculation techniques, to check the calculation results during preparation of the SM reactor safety substantiation report and to estimate the possibility of using such fuel elements in other projects. During tests of fuel rods under deviation from normal operating conditions their advantages were shown over fuel elements, the components of which were produced using the Al-based alloys. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Saurwein, J.J.

1977-08-01

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

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.

System for assembling nuclear fuel elements

International Nuclear Information System (INIS)

1980-01-01

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.)

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

International Nuclear Information System (INIS)

Suresh Varma, P.V.; Gupta, Amit; Amit, K.; Bhatt, R.B.; Afzal, Mohd.; Panakkal, J.P.; Kamath, H.S.

2009-02-01

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

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

Science.gov (United States)

Najeeb, Umair

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

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.

A Study of the Temperature Distribution in UO{sub 2} Reactor Fuel Elements

Energy Technology Data Exchange (ETDEWEB)

Devold, I

1968-05-15

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 UO{sub 2} 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.

TRIP: a finite element computer program for the solution of convection heat transfer problems

International Nuclear Information System (INIS)

Slagter, W.; Roodbergen, H.A.

1976-01-01

The theory and use of the finite element code TRIP are described. The code calculates temperature distributions in three-dimensional continua subjected to convection heat transfer. A variational principle for transport phenomena is applied to solve the convection heat transfer problem with temperature and heat flux boundary conditions. The finite element discretization technique is used to reduce the continuous spatial solution into a finite number of unknowns. The method is developed in detail to determine temperature distributions in coolant passages of fuel rod bundles which are idealized by hexahedral elements. The development of the TRIP code is discussed and the listing of the program is given in FORTRAN IV. An example is given to illustrate the validity and practicality of the method

Inert matrix fuel in dispersion type fuel elements

Energy Technology Data Exchange (ETDEWEB)

Savchenko, A.M. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)]. E-mail: sav@bochvar.ru; Vatulin, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Morozov, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Sirotin, V.L. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Dobrikova, I.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kulakov, G.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Ershov, S.A. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kostomarov, V.P. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Stelyuk, Y.I. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)

2006-06-30

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg{sup -1} (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Inert matrix fuel in dispersion type fuel elements

Science.gov (United States)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1984-01-01

The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

Coherence of reactor design and fuel element design

International Nuclear Information System (INIS)

Vom Scheidt, S.

1995-01-01

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.) [de

Characteristics and behaviour of the PHENIX fuel element

International Nuclear Information System (INIS)

Delpeyroux, P.; Balloffet, Y.; Blanchard, P.; Courcon, P.; Jallade, M.; Millet, P.; Rousseau, J.; Carteret, Y.; Coulon, P.

1977-01-01

The Phenix reactor has been in regular industrial operation for two years and has functioned very satisfactorily thanks in particular to the very good behaviour of the fuel element. A brief description is given of the fuel element and the operating conditions which were set for the fuel at the time of start-up (50000 MWd/t). The surveillance scheme is then described with the examinations in the hot laboratory on the basis of which it was possible to achieve the nominal specific burn-up and then to clear the Phenix fuel for a specific burn-up of 60000 MWd/t or 7 at.%. The behaviour of the mixed oxide (U, Pu)O 2 is quite normal and conforms to predictions as regards the heat conditions, swelling and fission gas release. The corrosion reaction between the oxide and the clad is progressing slowly and affects only small thicknesses of cladding. The mechanical integrity of the clad under thermal stresses and the stresses produced by swelling and fission gas pressure do not pose any special problem. The present limitation of the irradiation level is essentially based on the permissible deformations due to swelling and irradiation creep in the fuel pin cladding and in the hexagonal tube. This corresponds to damage to the steel of the order of 80 dpa. The mechanical behaviour of the bundle of pins, its interaction with the hexagonal tube and the thermohydraulic consequences of the deformations are all satisfactory to date. The absence of fuel failures is also worth noting; the only burst can detected to date did not affect either the operation of the fuel assembly or the performance of the reactor [fr

Nuclear fuel element end fitting

International Nuclear Information System (INIS)

Jabsen, F.S.

1979-01-01

A typical embodiment of the invention has an array of sockets that are welded to the intersections of the plates that form the upper and lower end fittings of a nuclear reactor fuel element. The sockets, which are generally cylindrical in shape, are oriented in directions that enable the longitudinal axes of the sockets to align with the longitudinal axes of the fuel rods that are received in the respective sockets. Detents impressed in the surfaces of the sockets engage mating grooves that are formed in the ends of the fuel rods to provide for the structural integrity of the fuel element

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

International Nuclear Information System (INIS)

Languasco, J.

1989-01-01

The peruvian research reactor RP-10 is composed of a compound nucleus of boxes containing fuel plates which are cooled with light water in order to remove heat produced by fission of uranium atoms. However from a certainty viewpoint, it exists certain restrictions to design the cooling system. The most admissible caloric flux of 90.3 watts/cm 2 is deflux of 90.3 watts/cm 2 is determined on the basis of these thermic restrictions when cooling speed is 409 cm/sec permitting at least 24 fuel elements(boxes) within the nucleus. On the basis of restrictions of load loss in the nucleus, it would be permitted at least 18 fuel elements, but this quantity breaks thermic restrictions for this reason, 24 boxes in the nucleus will be the minimum number of elements

Method for inspecting nuclear reactor fuel elements

International Nuclear Information System (INIS)

Jabsen, F.S.

1979-01-01

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

Integral nuclear fuel element assembly

International Nuclear Information System (INIS)

Schluderberg, D. C.

1985-01-01

An integral nuclear fuel element assembly utilizes longitudinally finned fuel pins. The continuous or interrupted fins of the fuel pins are brazed to fins of juxtaposed fuel pins or directly to the juxtaposed fuel pins or both. The integrally brazed fuel assembly is designed to satisfy the thermal and hydraulic requirements of a fuel assembly lattice having moderator to fuel atom ratios required to achieve high conversion and breeding ratios

A non-linear, finite element, heat conduction code to calculate temperatures in solids of arbitrary geometry

International Nuclear Information System (INIS)

Tayal, M.

1987-01-01

Structures often operate at elevated temperatures. Temperature calculations are needed so that the design can accommodate thermally induced stresses and material changes. A finite element computer called FEAT has been developed to calculate temperatures in solids of arbitrary shapes. FEAT solves the classical equation for steady state conduction of heat. The solution is obtained for two-dimensional (plane or axisymmetric) or for three-dimensional problems. Gap elements are use to simulate interfaces between neighbouring surfaces. The code can model: conduction; internal generation of heat; prescribed convection to a heat sink; prescribed temperatures at boundaries; prescribed heat fluxes on some surfaces; and temperature-dependence of material properties like thermal conductivity. The user has a option of specifying the detailed variation of thermal conductivity with temperature. For convenience to the nuclear fuel industry, the user can also opt for pre-coded values of thermal conductivity, which are obtained from the MATPRO data base (sponsored by the U.S. Nuclear Regulatory Commission). The finite element method makes FEAT versatile, and enables it to accurately accommodate complex geometries. The optional link to MATPRO makes it convenient for the nuclear fuel industry to use FEAT, without loss of generality. Special numerical techniques make the code inexpensive to run, for the type of material non-linearities often encounter in the analysis of nuclear fuel. The code, however, is general, and can be used for other components of the reactor, or even for non-nuclear systems. The predictions of FEAT have been compared against several analytical solutions. The agreement is usually better than 5%. Thermocouple measurements show that the FEAT predictions are consistent with measured changes in temperatures in simulated pressure tubes. FEAT was also found to predict well, the axial variations in temperatures in the end-pellets(UO 2 ) of two fuel elements irradiated

Modelling of pressurized water reactor fuel, rod time dependent radial heat flow with boundary element method; Modeliranje spremenljivega radijalnega toplotnega toka tlacnovodne gorivne palice z metodo robnih elementov

Energy Technology Data Exchange (ETDEWEB)

Sarler, B [Institut Jozef Stefan, Ljubljana (Yugoslavia)

1987-07-01

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

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1981-01-01

An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

Process and device for cooling of nuclear reactor fuel elements enclosed in a transport container

International Nuclear Information System (INIS)

Stiefel, M.

1986-01-01

In order to remove the post-decay heat of the fuel elements contained in them, transport containers for burnt-up fuel elements can be connected to a water cooling circuit. In order to avoid thermal shocks, a tenside forming foam and air are introduced into the cooling circuit before its entry into the transport container in the direction of flow. The tenside and air continue to be supplied until the temperature inside the transport container has fallen below the temperature at which the foam is destroyed. By adding tenside and air, a two phase mixture is produced, which foams greatly when it enters the transport container and which cools the fuel elements so as to protect them.(orig./HP) [de

Fast breeder fuel element development

International Nuclear Information System (INIS)

Marth, W.; Muehling, G.

1983-08-01

This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

Finite element method programs to analyze irradiation behavior of fuel pellets

International Nuclear Information System (INIS)

Yamada, Rayji; Harayama, Yasuo; Ishibashi, Akihiro; Ono, Masao.

1979-09-01

For the safety assessment of reactor fuel, it is important to grasp local changes of fuel pins due to irradiation in a reactor. Such changes of fuel result mostly from irradiation of fuel pellets. Elasto-plastic analysis programs based on the finite element method were developed to analyze these local changes. In the programs, emphasis is placed on the analysis of cracks in pellets; the interaction between cracked-pellets and cladding is not taken into consideration. The two programs developed are FEMF3 based on a two-dimensional axially symmetric model (r-z system) and FREB4 on a two-dimensional plane model (r-theta system). It is discussed in this report how the occurrence and distribution of cracks depend on heat rate of the fuel pin. (author)

Unification of fuel elements for research reactors

International Nuclear Information System (INIS)

Vatulyn, A.V.; Stetskyi, Y.A.; Dobrikova, I.V.

1997-01-01

To the purpose of fuel elements unification the possibility of rod fuel assembly (FA) using in the cores of research reactors have been considered in this paper. The calculation results of geometric, hydraulic and thermotechnical parameters of rod assembly are submitted. Several designs of finned square fuel element and fuel assembly are proposed on base of analysis of rod FA characteristics in compare of tube ones. The fuel elements specimens and the model assembly are manufactured. The developed designs are the basis for further optimization after neutron-physical calculations of cores. (author)

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

International Nuclear Information System (INIS)

Walker, J.R.; de Vaal, J.W.; Arimescu, V.I.; McGrady, T.G.; Wong, C.

1992-04-01

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

Control device for can failures of liquid cooled nuclear reactor fuel elements

International Nuclear Information System (INIS)

Althaus, D.; Mohm, F.; Nyhof, M.

1974-01-01

Checking of the fuel or breeding elements of, e.g., sodium-cooled reactors is done by detecting fission products in the coolant, with a flushing gas like argon removing the fission products from the coolant and carrying them to a detector system. In order to increase the escape rate of the fission products from the elements, these are lifted by a hoisting unit into a pit reaching down below the coolant level and are then heated. Heating is achieved by the decay heat or by an additional heating in the receiving pit which at this point is thermically insulated from the exterior. The flushing gas is blown radially into the receiving pit from below. The rising bubbles take along the fission products to a scintillation counter mounted on the head of the receiving vessel. This vessel may have double walls with coolant flowing through the interspace. (DG) [de

Core conversion from rod to plate type fuel elements in research reactors

International Nuclear Information System (INIS)

Khattab, M.S.; Mina, A.R.

1997-01-01

Core thermalhydraulic analysis have been performed for rod and plate types fuel elements without altering the core bundles square grid spacer (68 mm, side) and coolant mass flow rate. The U O 2 -Mg, 10% enrichment rod type fuel elements are replaced by the MTR plate type, U-Al alloy of 20% enrichment. Coolant mass flux increased from 2000 kg/m 2 S to 5000 kg/m 2 S. Reactor power could be upgraded from 2 to 10 MW without significantly altering the steady state, thermal-hydraulic safety margins. Fuel, clad and coolant transient temperatures are determined inside the core hot channel during flow coast down using paret code. Residual heat removal system of 20% coolant capacity is necessary for upgrading reactor power to encounter the case of pumps off at 10 MW nominal operation. 6 figs., 2 tabs

Apparatus for locating defective nuclear fuel elements

International Nuclear Information System (INIS)

Lawrie, W.E.

1979-01-01

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

Hydriding failure in water reactor fuel elements

International Nuclear Information System (INIS)

Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

1980-01-01

Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

Nuclear fuel elements

International Nuclear Information System (INIS)

Ainsworth, K.F.

1979-01-01

A nuclear fuel element is described having a cluster of nuclear fuel pins supported in parallel, spaced apart relationship by transverse cellular braces within coaxial, inner and outer sleeves, the inner sleeve being in at least two separate axial lengths, each of the transverse braces having a peripheral portion which is clamped peripherally between the ends of the axial lengths of the inner sleeve. (author)

LEU fuel element produced by the Egyptian fuel manufacturing pilot plant

International Nuclear Information System (INIS)

Zidan, W.I.

2000-01-01

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 (UF 6 , 19.75% U 235 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)

Reproduction of the RA reactor fuel element fabrication

International Nuclear Information System (INIS)

Novakovic, M.

1961-12-01

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 [sr

Storage device for a long nuclear reactor fuel element and/or a long nuclear reactor fuel element part

International Nuclear Information System (INIS)

Vogt, M.; Schoenwitz, H.P.; Dassbach, W.

1986-01-01

The storage device can be erected in a dry storage room for new fuel elements and also in a storage pond for irradiated fuel elements. It consists of shells, which are arranged vertically and which have a lid. A suspension for the fuel element is provided on the underside of the lid, which acts as a support against squashing or bending in case of vertical forces acting (earthquake). (DG) [de

Chilean fuel elements fabrication progress report

International Nuclear Information System (INIS)

Baeza, J.; Contreras, H.; Chavez, J.; Klein, J.; Mansilla, R.; Marin, J.; Medina, R.

1993-01-01

Due to HEU-LEU core conversion necessity for the Chilean MTR reactors, the Fuel Elements Plant is being implemented to LEU nuclear fuel elements fabrication. A glove box line for powder-compact processing designed at CCHEN, which supposed to operate under an automatic control system, is at present under initial tests. Results of first natural uranium fuel plates manufacturing runs are shown

Theoretical analysis of the temperature changes and resultant loss of fuel integrity in the IEA-R1 research reactor fuel elements following a loss of coalant accident

International Nuclear Information System (INIS)

Garone, J.G.M.

1983-01-01

The IEA-R1 core following a loss of coolant accident (LOCA) is analysed. THe AIRLOCA code was used to calculate fuel temperatures, heat generation due to fission product decay and convective and radiative heat transfer from the fuel elements to the surrounding air both during and following the loss of coolant. The influence of certain critical parameters, such as log time, specific power was studied in detail. Representative results are presented and suggestions made to ensure that fuel integrity is maintained following a LOCA. (Author) [pt

Report on the disposal of radioactive wastes and spent fuel elements from Baden-Wuerttemberg; Bericht ueber die Entsorgung von radioaktiven Abfaellen und abgebrannten Brennelementen aus Baden-Wuerttemberg

Energy Technology Data Exchange (ETDEWEB)

NONE

2017-04-15

The report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg covers the following issues: legal framework for the nuclear disposal; producer of spent fuels and radioactive wastes in Baden- Report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg; low- and medium-level radioactive wastes (non heat generating radioactive wastes); spent fuels and radioactive wastes from waste processing (heat generating radioactive wastes); final disposal.

Critical evaluation of the experiments and mathematical models for the determination of fission product release from the spherical fuel elements in cases of core heating accidents in modular HTR's

International Nuclear Information System (INIS)

Bailly, H.W.

1987-01-01

In this work, the thermal behaviour of modular reactors in cases of core heating accidents and the physical phenomena relevant for a release of radioactive materials from HTR fuel elements are explained as far as is necessary for understanding the work. The present mathematical models by which the release of radioactive materials from HTR fuel elements due to diffusion or breaking particles in cases of core heating accidents are also described, examined and evaluated with regard to their applicability to module reactors. The experiments used to verify the mathematical models are also evaluated. The mathematical models are in nearly all cases computer programs, which describe the complicated process of releasing radioactive materials quantitative mathematically. One should point out that these models are constantly being developed further, in line with the increasing amount of knowledge. To conclude the work, proposals are made for improving the certainty of information from experiments and mathematical models to determine the release behaviour of modular reactors. (orig./GL) [de

Nuclear fuel elements

International Nuclear Information System (INIS)

Obara, Hiroshi.

1981-01-01

Purpose: To suppress iodine release thereby prevent stress corrosion cracks in fuel cans by dispersing ferrous oxide at the outer periphery of sintered uranium dioxide pellets filled and sealed within zirconium alloy fuel cans of fuel elements. Constitution: Sintered uranium dioxide pellets to be filled and sealed within a zirconium alloy fuel can are prepared either by mixing ferric oxide powder in uranium dioxide powder, sintering and then reducing at low temperature or by mixing iron powder in uranium dioxide powder, sintering and then oxidizing at low temperature. In this way, ferrous oxide is dispersed on the outer periphery of the sintered uranium dioxide pellets to convert corrosive fission products iodine into iron iodide, whereby the iodine release is suppressed and the stress corrosion cracks can be prevented in the fuel can. (Moriyama, K.)

Soybean-derived biofuels and home heating fuels.

Science.gov (United States)

Mushrush, George W; Wynne, James H; Willauer, Heather D; Lloyd, Christopher L

2006-01-01

It is environmentally enticing to consider replacing or blending petroleum derived heating fuels with biofuels for many reasons. Major considerations include the soaring worldwide price of petroleum products, especially home heating oil, the toxicity of the petroleum-derived fuels and the environmental damage that leaking petroleum tanks afford. For these reasons, it has been suggested that domestic renewable energy sources be considered as replacements, or at the least, as blending stocks for home heating fuels. If recycled soy restaurant cooking oils could be employed for this purpose, this would represent an environmental advantage. Renewable plant sources of energy tend to be less toxic than their petroleum counterparts. This is an important consideration when tank leakage occurs. Home fuel oil storage tanks practically always contain some bottom water. This water environment has a pH value that factors into heating fuel stability. Therefore, the question is: would the biofuel help or exacerbate fuel stability and furnace maintenance issues?

International experience in conditioning spent fuel elements

International Nuclear Information System (INIS)

Ashton, P.

1991-04-01

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) [de

Analysis of gamma heating at TRIGA mark reactor core Bandung using plate type fuel

International Nuclear Information System (INIS)

Setiyanto; Tukiran Surbakti

2016-01-01

In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0.87 W/g), but very low value for Lazy Susan position (lest then 0.11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. (author)

Fuel element box inspection device

International Nuclear Information System (INIS)

Ortmayer, R.M.; Pick, W.

1985-01-01

The invention concerns a device for inspecting the outer geometry of a long fuel element box by measuring the surface contours over its longitudinal crossection and along its length by sensors. These are kept in a sledge which can be moved along the fuel element guide in a slot guide. The measurement signals reach an evaluation device outside the longitudinal box. (orig./HP) [de

About calculation results of heat transfer in the fuel assembly clusters cooled by water with supercritical parameters

International Nuclear Information System (INIS)

Grabezhnaya, V.A.

2008-01-01

Paper reviews the numerical investigation into the heat transfer in the supercritical water cooled fuel assemblies on the basis of the various commercial codes. The turbulence available models specified in the codes describe adequately the experimental data in tubes within the range of flow temperatures away from the pseudocritical point, as well as under high mass velocities. There are k-ε type turbulence models that show qualitatively the local acceleration (slowdown) of the heat transfer in tubes, but they fail to describe the mentioned phenomena quantitatively. To determine the effect of grid spacers on the suppression of the heat transfer local slowdown and on the heat transfer acceleration in fuel assemblies and to ensure more accurate calculation of the fuel element cladding maximum temperature one should perform a number of the experiments making use of the fuel assembly models [ru

In-pile tests of HTGR fuel particles and fuel elements

International Nuclear Information System (INIS)

Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

1985-01-01

Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

Fuel element structure - design, production and operational behaviour

International Nuclear Information System (INIS)

Pott, G.; Dietz, W.

1985-01-01

The lectures held at the meeting of the fuel element section of the Kerntechnische Gesellschaft gives a survey of developments in fuel element structure design for PWR-type, BWR-type and fast breeder reactors. For better utilization of the fuel, concepts have been developed for re-usable, removable and thus repairable fuel elements. Furthermore, the manufacturing methods for fuel element structures were refined to achieve better quality and more efficient manufacturing methods. Statements on the dimensional behaviour and on the mechanical stability of fuel element structures in normal and accident operation could be made on the basis of post-irradiation inspections. Finally, the design, manufacture and irradiation behaviour of graphite reflectors in HTGR-type reactors are described. The 12 lectures have been recorded in the data base separately. (RF) [de

Heat planning for fossil-fuel-free district heating areas with extensive end-use heat savings

DEFF Research Database (Denmark)

Harrestrup, Maria; Svendsen, S.

2014-01-01

is a theoretical investigation of the district heating system in the Copenhagen area, in which heat conservation is related to the heat supply in buildings from an economic perspective. Supplying the existing building stock from low-temperature energy resources, e.g. geothermal heat, might lead to oversized......The Danish government plans to make the Danish energy system to be completely free of fossil fuels by 2050 and that by 2035 the energy supply for buildings and electricity should be entirely based on renewable energy sources. To become independent from fossil fuels, it is necessary to reduce...... the energy consumption of the existing building stock, increase energy efficiency, and convert the present heat supply from fossil fuels to renewable energy sources. District heating is a sustainable way of providing space heating and domestic hot water to buildings in densely populated areas. This paper...

Catalogue of fuel elements - 1. addendum October 1958

International Nuclear Information System (INIS)

Even, A.

1957-01-01

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

Quality assurance of fuel elements

International Nuclear Information System (INIS)

Hoerber, J.

1980-01-01

The quality assurance activities for reactor fuel elements are based on a quality assurance system which implies the requirements resulting from the specifications, regulations of the authorities, national standards and international rules and regulations. The quality assurance related to production of reactor fuel will be shown for PWR fuel elements in all typical fabrication steps as conversion into UO 2 -powder, pelletizing, rodmanufacture and assembling. A wide range of destructive and nondestructive techniques is applied. Quality assurance is not only verified by testing techniques but also by process monitoring by means of parameter control in production and testing procedures. (RW)

Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Bibilashvili, Yu K; Nekrasova, G A; Sukhanov, G I

1989-01-01

In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified.

Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Nekrasova, G.A.; Sukhanov, G.I.

1989-01-01

In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Butterfield, C.E.; Waite, E.

1982-01-01

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)

The behaviour of spherical HTR fuel elements under accident conditions

Energy Technology Data Exchange (ETDEWEB)

Schenk, W; Naoumidis, A [Institute for Reactor Material, KFA Juelich (Germany)

1985-07-01

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 UO{sub 2}-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.

Fuel element transfer cask modelling using MCNP technique

International Nuclear Information System (INIS)

Rosli Darmawan

2009-01-01

Full text: 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. (author)

Fuel Element Transfer Cask Modelling Using MCNP Technique

International Nuclear Information System (INIS)

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.

Fuel elements for LWR power plants

International Nuclear Information System (INIS)

Roepenack, H.

1977-01-01

About five times more expensive than the fabrication of a fuel element is the enriched uranium contained therein; soon the monthly interest charges for the uranium value of a fuel element reload will account for five percent of the fabrication costs, and much more expensive than all this together can it be if reactor operation has to be interrupted because of damaged elements. Thus, quality assurance comes first. (orig.) [de

Fuel element for a nuclear reactor

International Nuclear Information System (INIS)

Linning, D.L.

1977-01-01

An improvement of the fuel element for a fast nuclear reactor described in patent 15 89 010 is proposed which should avoid possible damage due to swelling of the fuel. While the fuel element according to patent 15 89 010 is made in the form of a tube, here a further metal jacket is inserted in the centre of the fuel rod and the intermediate layer (ceramic uranium compound) is provided on both sides, so that the nuclear fuel is situated in the centre of the annular construction. Ceramic uranium or plutonium compounds (preferably carbide) form the fuel zone in the form of circular pellets, which are surrounded by annular gaps, so that gaseous fission products can escape. (UWI) [de

Nuclear fuel elements and assemblies

International Nuclear Information System (INIS)

Saito, Shozo; Maki, Hideo.

1982-01-01

Purpose: To facilitate the attainment of the uranium enrichment or gadolinia enrichment of a pellet filled in a fuel element. Constitution: The axial length of a pellet filled in a fuel element is set to predetermined sizes according to the uranium enrichment factor, gadolinia enrichment or their combination. Thus, the uranium enrichment factor or gadolinia enrichment can be identified by attaining the axial length of the pellet by using such a pellt. (Kamimura, M.)

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.

Reactor fuel element and fuel assembly

International Nuclear Information System (INIS)

Okada, Seiji; Ishida, Tsuyoshi; Ikeda, Atsuko.

1997-01-01

A mixture of fission products and burnable poisons is disposed at least to a portion between MOX pellets to form a burnable poison-incorporated fuel element without mixing burnable poisons to the MOX pellets. Alternatively, a mixture of materials other than the fission products and burnable poisons is formed into disks, a fuel lamination portion is divided into at least to two regions, and the ratio of number of the disks of the mixture relative to the volume of the region is increased toward the lower portion of the fuel lamination portion. With such a constitution, the axial power distribution of fuels can be made flat easily. Alternatively, the thickness of the disk of the mixture is increased toward the lower region of the fuel lamination portion to flatten the axial power distribution of the fuels in the same manner easily. The time and the cost required for the manufacture are reduced, and MOX fuels filled with burnable poisons with easy maintenance and control can be realized. (N.H.)

The development of fuel elements for boiling water reactors

International Nuclear Information System (INIS)

Holzer, R.; Kilian, P.

1984-01-01

The longevity of today's standard fuel elements constitutes a sound basis for designing advanced fuel elements for higher discharge burnups. Operating experience as well as postirradiation examinations of discharged fuel elements indicate that the technical limits have not reached by far. However, measures to achieve an economic and reliable fuel cycle are not restricted to the design of fuel elements, but also extend into such fields as fuel management and the mode of reactor operation. Fuel elements can be grouped together in zones in the core as a function of burnup and reactivity. The loading scheme can be aligned to this approach by concentrating on typical control rod positions. Reloads can also be made up of two sublots of fuel elements with different gadolinium contents. Longer cycles, e.g., of eighteen instead of twelve months, are easy to plan reactivitywise by increasing the quantity to be replaced from at present one quarter to one third. In fuel elements designed for higher burnups, the old scheme of reloading one quarter of the fuel inventory can be retained. The measures already introduced or in the planning stage incorporate a major potential for technical and economic optimization of the fuel cycle in boiling water reactors. (orig.) [de

Determining the elemental composition of fuels by bomb calorimetry and the inverse correlation of HHV with elemental composition

Energy Technology Data Exchange (ETDEWEB)

Bech, Niels; Jensen, Peter Arendt; Dam-Johansen, Kim [Department of Chemical and Biochemical Engineering, CHEC Research Centre, Technical University of Denmark, Soeltofts Plads, Building 229, DK-2800 Kongens Lyngby (Denmark)

2009-03-15

This article presents a method to obtain a simplified elemental analysis of an organic sample in which oxygen, nitrogen, and sulphur are lumped. The method uses a bomb calorimeter, water, and ash measurements combined with a numerical procedure based on a generalised equation for predicting higher heating value. By analysing pure organic substances, literature data, and fuels it is demonstrated that the method can provide hydrogen estimates within {+-}0.7% daf. and carbon and sum of oxygen, nitrogen, and sulphur estimates within {+-}2% daf. for fuels containing less than 90% ash db., 2% nitrogen daf., and 1% daf. sulphur. (author)

Milling Behavior of Matrix Graphite Powders with Different Binder Materials in HTGR Fuel Element Fabrication: I. Variation in Particle Size Distribution

International Nuclear Information System (INIS)

Lee, Young Woo; Cho, Moon Sung

2011-01-01

The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a matrix graphite powder properly prepared and pressed into a spherical shape or a cylindrical compact finally heat-treated at about 1900 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, overcoating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. In order to develop a fuel compact fabrication technology, it is important to develop a technology to prepare the matrix graphite powder (MGP) with proper characteristics, which has a strong influence on further steps and the material properties of fuel element. In this work, the milling behavior of matrix graphite powder mixture with different binder materials and their contents was investigated by analyzing the change in particle size distribution with different milling time

Nuclear reactor fuel element with a cluster of parallel fuel pins

International Nuclear Information System (INIS)

Macfall, D.; Butterfield, C.E.; Butterfield, R.S.

1977-01-01

An improvement of the design of nuclear reactor fuel elements is described and illustrated by the example of a gas-cooled, graphite-moderated nuclear reactor. The fuel element has a cluster of parallel fuel pins with an outer can of structure material and an inner sleeve, as well as tie bars and spacing devices for all of these parts. The fuel element designed according to the invention allows lasy assembling and disassembling before and after use. During use, no relative axial motions are possible; nevertheless, the graphite sleeve is at no time subject to tensile stress: the individual parts are held in position from below by a single holding device. (UWI) [de

Transfer of heat to fluidized-solids beds

Energy Technology Data Exchange (ETDEWEB)

1952-10-16

The improvement in the method described and claimed in patent application 14,363/47 (136,186) for supplying heat to a dense turbulent mass of solid fluidized by a gas flowing upwardly therethrough and subjected to a high temperature in a treating zone, by heat transfer through heat-transfer surfaces of heat-transfer elements in contact with the said turbulent mass of finely divided solid and heated by means of a fluid heating medium, including burning fuels comprising contacting the said heat-transfer surfaces with a fuel and a combustion supporting gas under such conditions that the combustion of the fuel is localized in the heat-transfer element near the point of entry of the fuel and combustion-supporting gas and a substantial temperature gradient is maintained along the path of said fuel combustion-supporting gas and combustion products through the said heat-transfer element.

Capacity of the equipment family SICOM to inspect fuel elements; Capacidad de los equipos familia SICOM para inspeccionar elementos de combustible

Energy Technology Data Exchange (ETDEWEB)

Sanchez Siguero, A.; Sola, A.

2013-07-01

To check the status where the fuel assemblies are after has been operating in the core of nuclear plants, inspections have been conducted to carry out an improvement in the behavior of alloys used in pods of fuel, the control of corrosion of these pods because of heat, reducing the transfer of heat due to the oxide and with the support of visual inspections monitor the physical integrity of the fuel elements.

Finite-elements modeling of radiant heat transfers between mobile surfaces; Modelisation par elements finis de transferts radiatifs entre surfaces mobiles

Energy Technology Data Exchange (ETDEWEB)

Daurelle, J V; Cadene, V; Occelli, R [Universite de Provence, 13 - Marseille (France)

1997-12-31

In the numerical modeling of thermal industrial problems, radiant heat transfers remain difficult to take into account and require important computer memory and long computing time. These difficulties are enhanced when radiant heat transfers are coupled with finite-elements diffusive heat transfers because finite-elements architecture is complex and requires a lot of memory. In the case of radiant heat transfers along mobile boundaries, the methods must be optimized. The model described in this paper concerns the radiant heat transfers between diffuse grey surfaces. These transfers are coupled with conduction transfers in the limits of the diffusive opaque domain. 2-D and 3-D geometries are analyzed and two configurations of mobile boundaries are considered. In the first configuration, the boundary follows the deformation of the mesh, while in the second, the boundary moves along the fixed mesh. Matter displacement is taken into account in the term of transport of the energy equation, and an appropriate variation of the thermophysical properties of the transition elements between the opaque and transparent media is used. After a description of the introduction of radiative limit conditions in a finite-elements thermal model, the original methods used to optimize calculation time are explained. Two examples of application illustrate the approach used. The first concerns the modeling of radiant heat transfers between fuel rods during a reactor cooling accident, and the second concerns the study of heat transfers inside the air-gap of an electric motor. The method of identification of the mobile surface on the fixed mesh is described. (J.S.) 12 refs.

Finite-elements modeling of radiant heat transfers between mobile surfaces; Modelisation par elements finis de transferts radiatifs entre surfaces mobiles

Energy Technology Data Exchange (ETDEWEB)

Daurelle, J.V.; Cadene, V.; Occelli, R. [Universite de Provence, 13 - Marseille (France)

1996-12-31

In the numerical modeling of thermal industrial problems, radiant heat transfers remain difficult to take into account and require important computer memory and long computing time. These difficulties are enhanced when radiant heat transfers are coupled with finite-elements diffusive heat transfers because finite-elements architecture is complex and requires a lot of memory. In the case of radiant heat transfers along mobile boundaries, the methods must be optimized. The model described in this paper concerns the radiant heat transfers between diffuse grey surfaces. These transfers are coupled with conduction transfers in the limits of the diffusive opaque domain. 2-D and 3-D geometries are analyzed and two configurations of mobile boundaries are considered. In the first configuration, the boundary follows the deformation of the mesh, while in the second, the boundary moves along the fixed mesh. Matter displacement is taken into account in the term of transport of the energy equation, and an appropriate variation of the thermophysical properties of the transition elements between the opaque and transparent media is used. After a description of the introduction of radiative limit conditions in a finite-elements thermal model, the original methods used to optimize calculation time are explained. Two examples of application illustrate the approach used. The first concerns the modeling of radiant heat transfers between fuel rods during a reactor cooling accident, and the second concerns the study of heat transfers inside the air-gap of an electric motor. The method of identification of the mobile surface on the fixed mesh is described. (J.S.) 12 refs.

TRIGA fuel element burnup determination by measurement and calculation

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.; Persic, A.; Jeraj, R.

2000-01-01

To estimate the accuracy of the fuel element burnup calculation different factors influencing the calculation were studied. To cover different aspects of burnup calculations, two in-house developed computer codes were used in calculations. The first (TRIGAP) is based on a one-dimensional two-group diffusion approximation, and the second (TRIGLAV) is based on a two-dimensional four-group diffusion equation. Both codes use WIMSD program with different libraries forunit-cell cross section data calculation. The burnup accumulated during the operating history of the TRIGA reactor at Josef Stefan Institute was calculated for all fuel elements. Elements used in the core during this period were standard SS 8.5% fuel elements, standard SS 12% fuel elements and highly enriched FLIP fuel elements. During the considerable period of operational history, FLIP and standard fuel elements were used simultaneously in mixed cores. (authors)

Post-irradiation examination of the first SAP clad UO{sub 2} fuel elements irradiated in the X-7 organic loop

Energy Technology Data Exchange (ETDEWEB)

MacDonald, R. D.; Aspila, K.

1962-02-15

Seven fuel elements composing the first in-reactor test at Chalk River of SAP sheathing were irradiated in the X-7 organic loop. Activity, denoting a fuel failure, was detected in the loop coolant immediately after reactor start up; the fuel string was consequently removed from the loop nine hours later. Leak tests disclosed that five of the seven elements were defective. Inspection of the specimens showed essentially no change in element dimensions. Practically no organic fouling film was observed on the surface of the SAP cladding; organic coolant was found inside four of the defective elements. The appearance of the UO{sub 2} fuel was consistent with the irradiation time and the heat ratings achieved during the test. (author)

An improved assembly for the transport of fuel elements

International Nuclear Information System (INIS)

Myers, G.

1979-01-01

An improved assembly for the transport and storage of radioactive nuclear fuel elements is described. The fuel element transport canister is of the type in which the fuel elements are submerged in liquid with a self regulating ullage system, so that the fuel elements are always submerged in the liquid even when the assembly is used in one orientation during loading and another orientation during transportation. (UK)

Development of the Fuel Element Database of PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Nurhayati Ramli; Naim Syauqi Hamzah; Nurfazila Husain; Yahya Ismail; Mat Zin Mat Husin; Mohd Fairus Abd Farid

2015-01-01

Since June 28th, 1982, the PUSPATI TRIGA Reactor (RTP) operates safely with an accumulated energy release of about 17,200 MWhr, which corresponds to about 882 g of uranium burn-up. The reactor core has been reconfigured 15th times. Presently, there are 111 TRIGA fuel elements in the core, which 66 of the fuel elements are from the initial criticality while the rest of the fuel elements have been added to compensate the uranium consumption. As 59 % of the fuel elements are older than 30 years old, it is necessary to put the history of every fuel element in a database for easy access of the fuel element movement, inspection results history and integrity status. This paper intends to describe how the fuel element database is developed and related formulae used in determining the RTP fuel element elongation. (author)

Nuclear fuel element

International Nuclear Information System (INIS)

Iwano, Yoshihiko.

1993-01-01

Microfine cracks having a depth of less than 10% of a pipe thickness are disposed radially from a central axis each at an interval of less than 100 micron over the entire inner circumferential surface of a zirconium alloy fuel cladding tube. For manufacturing such a nuclear fuel element, the inside of the cladding tube is at first filled with an electrolyte solution of potassium chloride. Then, electrolysis is conducted using the cladding tube as an anode and the electrolyte solution as a cathode, and the inner surface of the cladding tube with a zirconium dioxide layer having a predetermined thickness. Subsequently, the cladding tube is laid on a smooth steel plate and lightly compressed by other smooth steel plate to form microfine cracks in the zirconium dioxide layer on the inner surface of the cladding tube. Such a compressing operation is continuously applied to the cladding tube while rotating the cladding tube. This can inhibit progress of cracks on the inner surface of the cladding tube, thereby enabling to prevent failure of the cladding tube even if a pellet/cladding tube mechanical interaction is applied. Accordingly, reliability of the nuclear fuel elements is improved. (I.N.)

Determination of the hydrogen content of fuel elements

International Nuclear Information System (INIS)

Soare, M.; Petriu, F.; Toma, V.

1995-01-01

A new method and apparatus are reported for determination of the total hydrogen content by measurements on as-manufactured fuel elements, heated at prescribed temperature values between 200 degrees C and 600 degrees C. The method is based on the catalytic oxidation of the organic compounds and transformation of the hydrogen in the equivalent water quantity which is analysed by a special infrared detector. Different types of measurements for determination of the hydrogen content from graphite coating, UO 2 pellets and filling gas are presented. Also, experimental observation regarding water release and graphite thermal decomposition kinetic are discussed. (author)

Fuel elements of research reactors in China

International Nuclear Information System (INIS)

Zhou Yongmao; Chen Dianshan; Tan Jiaqiu

1987-01-01

This paper describes the current status of design, fabrication of fuel elements for research reactors in China, emphasis is placed on the technology of fuel elements for the High Flux Engineering Test Reactor (HFETR). (author)

Fuel element

International Nuclear Information System (INIS)

Hirose, Yasuo.

1982-01-01

Purpose: To increase the plenum space in a fuel element used for a liquid metal cooled reactor. Constitution: A fuel pellet is secured at one end with an end plug and at the other with a coil spring in a tubular container. A mechanism for fixing the coil spring composed of a tubular unit is mounted by friction with the inner surface of the tubular container. Accordingly, the recoiling force of the coil spring can be retained by fixing mechanism with a small volume, and since a large amount of plenum space can be obtained, the internal pressure rise in the cladding tube can be suppressed even if large quantities of fission products are discharged. (Kamimura, M.)

Fabrication of the Spent Fuel Elements Rack on the ISFSF

International Nuclear Information System (INIS)

Slamet Wiranto; Sigit Purwanto; Safrul, H.

2004-01-01

The Interim Storage For Spent Fuel elements (ISFSF) was designed to be able to store the 33 spent fuel element racks with capacity of 1386 of normal spent fuel elements and 2 racks for 36 of defected ones. Until now, only 9 out of 33 racks of normal spent fuel elements and lout of 2 racks of defected fuel elements are available. Five of them have suffered from corrosion so that they are not fulfilled the requirements of the spent fuel elements storage anymore. Meanwhile, the spent fuel storage racks in the reactor are almost full. It means, the transfer of the spent fuel from reactor spent fuel storage to the ISFSF pool are compulsory needed. Therefore, it is necessary to provide the new ISFSF spent fuel storage rack with better material and fabrication method than the old one. In this design all materials consist of SS 316 L that are welded with the Argon TIG-welding. Right now there has been one new spent fuel storage rack fabricated with capacity of 42 normal spent fuel elements. (author)

TWODEE-2/MOD3, 2-D Time-Dependent Fuel Elements Thermal Analysis after PWR LOCA

International Nuclear Information System (INIS)

Lauben, G. N.

2001-01-01

1 - Description of problem or function: WREM-TOODEE2 is a two- dimensional, time-dependent, fuel-element thermal analysis program. Its primary purpose is to evaluate fuel-element thermal response during post-LOCA refill and reflood in a pressurized water reactor (PWR). 2 - Method of solution: TOODEE2 calculations are carried out in a two-dimensional mesh region defined in slab or cylindrical geometry by orthogonal grid lines. Coordinates which form order pairs are labeled x-y in slab geometry, and those in cylindrical geometry are labeled r-z for the axisymmetric case and r-theta for the polar case. Conduction and radiation are the only heat transfer mechanisms assumed within the boundaries of the mesh region. Convective and boiling heat transfer mechanisms are assumed at the boundaries. The program numerically solves the two-dimensional, time-dependent, heat conduction equation within the mesh region. 3 - Restrictions on the complexity of the problem: WREM-TOODEE2 considers only axisymmetric geometry although the equations for slab and polar geometry are included in the program

Nonstationary temperature field in the fuel element; Nestacionarno temperatursko polje u gorivnom elementu

Energy Technology Data Exchange (ETDEWEB)

Vehauc, A; Spasojevic, D [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1970-03-15

Nonstationary temperature field in the fuel element was examined for spatial and time distribution of the specific power generated in the fuel element. Analytical method was developed for calculating the temperature variation in the fuel element of a nuclear reactor for a typical shape of the heat generation function. The method is based on series expansion of the temperature field by self functions and application of Laplace transformation in time coordinate. For numerical calculation of the temperature distribution a computer code was developed based on the proposed method and applied on the ZUSE-Z-23 computer. Razmatrano je nestacionarno temperatursko polje u preseku sipke gorivnog elementa za slucaj prostorne i vremenske raspodele specificne generacije snage u gorivnom elementu. Razradjen je analii postupak odredjivanja promene temperature u gorivu nuklearnog reaktora za tipican oblik funkcije generacije toplote. Postupak se zasniva na razvoju temperaturskog polja po sopstvenim funkcijama i primeni Laplasove transformacije po vremenskoj koordinati. Za efektivno nalazenje temperaturskog polja, postupak je programiran za digitalnu racunsku masinu ZUSE-Z-23 (author)

Memory list for the ordering of nuclear fuel elements with UO2 fuel

International Nuclear Information System (INIS)

1977-01-01

The memory list will help to simplify and speed up the technical procedure of fuel element supply for nuclear reactors. Operators of nuclear power plants take great interest in the latest state of thechnology, if sufficiently tested, being applied with regard to material, manufacturing and testing methods. In order to obtain an unlimited availability of the nuclear plant in the future, this application of technology should be taken care of when designing and producing fuel elements. When ordering fuel elements special attention should be drawn to the interdependence of reactor and fuel element with reqard to design and construction, about which, howevers, no further details are given. When ordering fuel elements the operator give the producer all design data of the reactor core and the fuel elements as well as the planned operation mode. He also hands in the respective graphs and the required conditions for design so that a correct and detailed offer can be supplied. An exemplary extent of supply is shown in the given memory list. The regulations required herefore on passing technical material to the fuel element producers have to be established by agreements made by the customer. The order to be given should be itemized as follows: requirements, quality controland quality assurance, warranties and conditions, limits and extent of supply, terms of delivery. (orig./HP) [de

Fuel element cladding state change mathematical model for a WWER-1000 plant operated in the mode of varying loading

Directory of Open Access Journals (Sweden)

S. N. Pelykh

2010-09-01

Full Text Available Main features of a fuel element cladding state change mathematical model for a WWER-1000 reactor plant operated in the mode of varying loading are listed. The integrated model is based on the energy creep theory, uses the finite element method for imultaneous solution of the fuel element heat conduction and mechanical deformation equa-tions. Proposed mathematical model allows us to determine the influence of the WWER-1000 regime parameters and fuel assembly design characteristics on the change of cladding properties under different loading conditions of normal operation, as well as the cladding limiting state at variable loading depending on the length, depth and number of cycles.

Fundamental aspects of nuclear reactor fuel elements

International Nuclear Information System (INIS)

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

Shielded regeneration heating element for a particulate filter

Science.gov (United States)

Gonze, Eugene V [Pinckney, MI; Ament, Frank [Troy, MI

2011-01-04

An exhaust system includes a particulate filter (PF) that is disposed downstream from an engine. The PF filters particulates within an exhaust from the engine. A heating element heats particulate matter in the PF. A catalyst substrate or a flow converter is disposed upstream from said heating element. The catalyst substrate oxidizes the exhaust prior to reception by the heating element. The flow converter converts turbulent exhaust flow to laminar exhaust flow prior to reception by the heating element.

REACTOR FUEL ELEMENTS TESTING CONTAINER

Science.gov (United States)

Whitham, G.K.; Smith, R.R.

1963-01-15

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Analysis of heat transfer from fuel rods with externally attached thermocouples

International Nuclear Information System (INIS)

Gill, C.R.; Coddington, P.

1988-05-01

This paper describes the development of 2 and 3 dimensional finite element heat conduction models to simulate the behaviour of the external thermocouples attached to the LOFT fuel rods during the blowdown phase of a large break loss-of-coolant accident. To establish the model and determine the thermal coupling between the thermocouple and the fuel rod extensive use was made of two series of experiments performed at INEL in the LOFT Test Support Facility (LTSF). These experiments were high pressure reflood experiments with fluid conditions 'typical' of those seen during the bottom-up flow period of the LOFT experiments. (author)

Sensitivity analysis for heat diffusion in a fin on a nuclear fuel element

International Nuclear Information System (INIS)

Tito, Max Werner de Carvalho

2001-11-01

The modern thermal systems generally present a growing complexity, as is in the case of nuclear power plants. It seems that is necessary the use of complex computation and mathematical tools in order to increase the efficiency of the operations, reduce costs and maximize profits while maintaining the integrity of its components. The use of sensitivity calculations plays an important role in this process providing relevant information regarding the resultant influence of variation or perturbation of its parameters as the system works. This technique is better known as sensitivity analysis and through its use makes possible the understanding of the effects of the parameters, which are fundamental for the project preparation, and for the development of preventive and corrective handling measurements of many pieces of equipment of modern engineering. The sensitivity calculation methodology is based generally on the response surface technique (graphic description of the functions of interest based in the results obtained from the system parameter variation). This method presents a lot of disadvantages and sometimes is even impracticable since many parameters can cause alterations or perturbations to the system and the model to analyse it can be very complex as well. The utilization of perturbative methods result appropriate as a practical solution to this problem especially in the presence of complex equations. Also it reduces the resultant computational calculus time considerably. The use of these methods becomes an essential tool to simplify the sensitivity analysis. In this dissertation, the differential perturbative method is applied in a heat conduction problem within a thermal system, made up of a one-dimensional circumferential fin on a nuclear fuel element. The fins are used to extend the thermal surfaces where convection occurs; thus increasing the heat transfer to many thermal pieces of equipment in order to obtain better results. The finned claddings are

Fuel element store

International Nuclear Information System (INIS)

Wieser, R.

1987-01-01

The spherical fuel elements are stored dry in cans. The cans themselves are stacked in parallel storage shafts, which are combined into a rectangular storage space. The storage space is made earthquake-proof by surrounding it with concrete. It consists of a ceiling assembled from several steel parts, which is connected to the floor by support elements. A cooling air ventilation station supplies the individual storage shaft and therefore the cans with cooling air via incoming and outgoing pipes. (DG) [de

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)

Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

International Nuclear Information System (INIS)

Walton, J.T.

1992-11-01

This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code

Nuclear fuel element

International Nuclear Information System (INIS)

Hirama, H.

1978-01-01

A nuclear fuel element comprises an elongated tube having upper and lower end plugs fixed to both ends thereof and nuclear fuel pellets contained within the tube. The fuel pellets are held against the lower end plug by a spring which is supported by a setting structure. The setting structure is maintained at a proper position at the middle of the tube by a wedge effect caused by spring force exerted by the spring against a set of balls coacting with a tapered member of the setting structure thereby wedging the balls against the inner wall of the tube, and the setting structure is moved free by pushing with a push bar against the spring force so as to release the wedge effect

Irradiation of a CANDU UOsub(2) fuel element with twenty-three machined slits cut through the zircaloy sheath

International Nuclear Information System (INIS)

DaSilva, R.L.

1984-09-01

A CANDU fuel element was purposely defected, exposing a minimum UOsub(2) fuel stack area of 272 mmsup(2), by machining 23 longitudinal slits through the Zircaloy-4 sheathing. The element was then irradiated in the X-2 loop of the NRX reactor for a period of 14.64 effective full power days at a linear heat rating of 48 kW/m to investigate the relationship between fission product release and UOsub(2) oxidation behaviour in an element with minimal fuel-to-gap fission gas trapping. The fission product releases, as measured by on-line gamma-ray spectroscopy, revealed that the noble gases and radioiodines are both released from the UOsub(2) fuel matrix directly to the coolant via simple diffusion kinetics, and that their diffusivities in hyperstoichiometric UOsub(2) are approximately equal. The oxidation of UOsub(2) to the higher states UOsub(2+x), Usub(4)Osub(9) and Usub(3)Osub(8), was accompanied by substantial fuel swelling and sheath deformation preferentially located in the lower powered end of the element. The spalling and erosion behaviour of the fuel pellets was correlated to the rate of fuel oxidation

Structural analysis of reactor fuel elements

International Nuclear Information System (INIS)

Weeks, R.W.

1977-01-01

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

A Critical Heat Generation for Safe Nuclear Fuels after a LOCA

Directory of Open Access Journals (Sweden)

Jae-Yong Kim

2014-01-01

Full Text Available This study applies a thermo-elasto-plastic-creep finite element procedure to the analysis of an accidental behavior of nuclear fuel as well as normal behavior. The result will be used as basic data for the robust design of nuclear power plant and fuels. We extended the range of mechanical strain from small or medium to large adopting the Hencky logarithmic strain measure in addition to the Green-Lagrange strain and Almansi strain measures, for the possible large strain situation in accidental environments. We found that there is a critical heat generation after LOCA without ECCS (event category 5, under which the cladding of fuel sustains the internal pressure and temperature for the time being for the rescue of the power plant. With the heat generation above the critical value caused by malfunctioning of the control rods, the stiffness of cladding becomes zero due to the softening by high temperature. The weak position of cladding along the length continuously bulges radially to burst and to discharge radioactive substances. This kind of cases should be avoid by any means.

Computation fluid dynamic modelling of natural convection heat flow in unpumped molten salt fuel tubes

International Nuclear Information System (INIS)

Leefe, S.; Jackson-Laver, P.; Scott, I.R.

2015-01-01

Use of static molten salt nuclear fuel in simple tubes was discarded in 1949 without considering how convection could affect its utility. This poster describes CFD studies showing that such tubes are practical as fuel elements in essentially conventional fuel assemblies. They can achieve power densities above 250kW per liter of fuel salt (higher than PWR's) and do so without causing the tube wall to heat to dangerous levels. This discovery enables the achievement of the many benefits of molten salt fuel while utilizing the highly developed technology, regulatory, non proliferation and safety benefits of current fuel assembly technology. (author)

Nuclear reactor fuel element assembly spacer grid and method of making

International Nuclear Information System (INIS)

Chetter, J.

1975-01-01

A cellular fuel element assembly spacer grid is described which provides for resilient bracing of fuel pins in the cells of the grid by bow spring locating members projecting inside the cells of the grid to hold the fuel pins against opposed rigid stops also projecting inside the cells of the grid. The grid comprises two tiers each formed from intersecting strip members defining cells which are penetrated by the fuel pins and arranged parallel to one another but spaced apart. The bow spring locating members extend longitudinally between the two tiers and have end ferrules which are a sliding fit on locating members which extend longitudinally from the facing inner edges of the strip members forming the two tiers. The grid tiers are fabricated individually by heat bonding the intersecting strip members prior to assembling the tiers into the spacer grid. (U.S.)

Further developments of PWR and BWR fuel elements

International Nuclear Information System (INIS)

Sofer, G.A.; Busselman, G.J.; Federico, L.J.

1988-01-01

The performance, safety, and economy of nuclear power plants in inluenced very decisively by the quality of their fuel elements. This is why quality assurance in fuel fabrication has been a factor of great importance from the outset. Operating experince and more stringent performance requirements have resulted in a continuous process of further development of fuel elements, which has been reflected also in lower and lower failure rates and increasingly higher burn-ups. Next to further development also innovation has been an important factor contributing to the present high quality level of fuel elements, which also has allowed fuel cycle costs to be decreased quite considerably. (orig.) [de

Premiering SAFE for Safety Added Fuel Element - 15020

International Nuclear Information System (INIS)

Bhowmik, P.K.; Shamim, J.A.; Suh, K.Y.; Suh, K.S.

2015-01-01

The impact of the Fukushima accident has been the willingness to implement passive safety measures in reactor design and to simplify reactor design itself. Within this framework, a new fuel element, named SAFE (Safety Added Fuel Element) based on the concept of accident tolerant fuel, is presented. SAFE is a new type of fuel element cooled internally and externally by light water and with stainless steel as the cladding material. The removal of boron may trigger a series of changes which may simplify the system greatly. A simplified thermal analysis of SAFE shows that the fuel centerline temperature is well below the maximal limit during the normal operation of the plant

Commercial Aspect of Research Reactor Fuel Element Production

International Nuclear Information System (INIS)

Susanto, B.G; Suripto, A

1998-01-01

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

Fuel-clad heat transfer coefficient of a defected fuel rod

International Nuclear Information System (INIS)

Bruet, M.; Stora, J.P.

1976-01-01

A special rod has been built with a stack of UO 2 pellets inside a thick zircaloy clad. The atmosphere inside the fuel rod can be changed and particularly the introduction of water is possible. The capsule was inserted in the Siloe pool reactor in a special device equipped with a neutron flux monitor. The fuel centerline temperature and the temperature at a certain radius of the clad were recorded by two thermocouples. The temperature profiles in the fuel and in the cladding have been calculated and then the heat transfer coefficient. In order to check the proper functioning of the device, two runs were successively achieved with a helium atmosphere. Then the helium atmosphere inside the fuel rod was removed and replaced by water. The heat transfer coefficients derived from the measurements at low power level are in agreement with the values given by the model based on thermal conductivity. However, for higher power levels, the heat transfer coefficients become higher than those based on the calculated gap

Convective parameters in fuel elements for research nuclear reactors

International Nuclear Information System (INIS)

Lopez Martinez, C.D.

1992-01-01

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)

Nuclear criticality assessment of LEU and HEU fuel element storage

International Nuclear Information System (INIS)

Pond, R.B.; Matos, J.E.

1984-01-01

Criticality aspects of storing LEU (20%) and HEU (93%) fuel elements have been evaluated as a function of 235 U loading, element geometry, and fuel type. Silicide, oxide, and aluminide fuel types have been evaluated ranging in 235 U 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)

Development of a system for rapid discharge of spherical fuel elements as a diversitary afterheat removal system for pebble-bed HTR-type reactors

International Nuclear Information System (INIS)

Phlippen, P.W.

1982-07-01

Owing to its spherical fuel elements the pebble bed high temperature reactor provides the possibility to remove these fuel elements rapidly from the reactor for the purpose of after-heat removal and cooling in situations of danger and to collect them in easily cooled tanks. The paper investigates and represents fields of problems such as critically behaviour of core and fuel element collecting tanks, emission time of the core, thermodynamics in the vessel etc. by example of the PNP-500 reactor concept. A selection for the construction proposal was made from the performance possibilities of the three necessary main components prestressed-concrete vessel closure, fuel element guide and fuel element storage with cooling system. The proposal includes a prestressed concrete vessel closure opening by hydraulics as well as three annular fuel element storage modules cooled with the containment air by natural convection. (orig.) [de

Heat and Mass Transfer during Hydrogen Generation in an Array of Fuel Bars of a BWR Using a Periodic Unit Cell

Directory of Open Access Journals (Sweden)

H. Romero-Paredes

2012-01-01

Full Text Available This paper presents, the numerical analysis of heat and mass transfer during hydrogen generation in an array of fuel cylinder bars, each coated with a cladding and a steam current flowing outside the cylinders. The analysis considers the fuel element without mitigation effects. The system consists of a representative periodic unit cell where the initial and boundary-value problems for heat and mass transfer were solved. In this unit cell, we considered that a fuel element is coated by a cladding with steam surrounding it as a coolant. The numerical simulations allow describing the evolution of the temperature and concentration profiles inside the nuclear reactor and could be used as a basis for hybrid upscaling simulations.

Attempt to produce silicide fuel elements in Indonesia

International Nuclear Information System (INIS)

Soentono, S.; Suripto, A.

1991-01-01

After the successful experiment to produce U 3 Si 2 powder and U 3 Si 2 -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 U 3 Si 2 -Al fuel elements, having similar specifications to the ones of U 3 O 8 -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.)

Fabrication of fuel elements on the basis of increased concentration fuel composition

International Nuclear Information System (INIS)

Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

2004-01-01

As a part of Russian Program RERTR Reduced Enrichment for Research and Test Reactors), at NCCP, Inc. jointly with the State Scientific Centre VNIINM the mastering in industrial environment of design and fabrication process of fuel elements (FE) with increased concentration fuel compositions is performed. Fuel elements with fuel composition on the basis of dioxide uranium with nearly 4 g/cm 3 fuel concentration have been produced thus confirming the principal possibility of fuel enrichment reduction down to 20% for research reactors which were built up according to the projects of the former USSR, by increasing the oxide fuel concentration in fuel assemblies (FAs). The form and geometrical dimensions of FEs and FAs shall remain unchanged, only uranium mass in FA shall be increased. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Ravndal, F

1967-12-15

The design, fabrication, irradiation results, and post-irradiation examination for three superheater test fuel elements are described. During the spring of 1966 these clusters, each consisting of six fuel rods, were successfully exposed in the superheater loop No. 5 in the R2 reactor for a maximum of 24 days at a maximum outer cladding surface temperature of {approx} 650 deg C. During irradiation the linear heat rating of the rods was in the range 400-535 W/cm. The diameter of the UO{sub 2} pellets was 11.5 and 13.0 mm; the wall thickness of the 20/25 Nb and 20/35 cladding was in every case 0.4 mm. The diametrical gap between fuel and cladding was one of the main parameters and was chosen to be 0.05, 0.07 and 0.10 mm. These experiments, to be followed by one high cladding temperature irradiation ({approx} 750 deg C) and one long time irradiation ({approx} 6000 MWd/tU), were carried out to demonstrate the operational capability of short superheater test fuel rods at steady and transient operational environments for the Marviken superheater fuel elements and also to provide confirmation of design criteria for the same fuel elements.

Design and operational behaviour of the SNR-reactor fuel element structure

International Nuclear Information System (INIS)

Dietz, W.; Toebbe, H.

1985-01-01

The fuel element and core concept of a fast breeder reactor is described by the example of the SNR 300 (1st core), and the requirements made on the fuel elements with respect to burnup and neutron dose are listed for existing and projected plants. Irradiation experiments carried out and operational experience gained with fuel elements show that the residence time of the fuel elements is influenced mainly by the stability of shape of the fuel element components. The requirements made with reference to neutron loading for future advanced high-performance fuel elements can not be anticipated from the present state of experience. Besides optimization of fuel element design and checking-out of the limits of operation by PFADFINDERELEMENTE elements, R and D work for the improvement of fuel element materials is also necessary. (orig.) [de

Criteria for the selection of graphites for HTR integral block fuel elements

International Nuclear Information System (INIS)

Knowles, A.N.

1980-01-01

This paper is concerned with the special requirements for integral block fuel elements of the type first used in the Fort St. Vrain reactor. The main idea of these elements is that the carrier block and separate graphite clad fuel pins are combined into a single monolith. This combination leads to lower fabrication costs and some improvement in the thermal performance (lower temperature difference between fuel and the surface of heat transfer into the coolant). The advent of block fuel for HTRs of the Fort St. Vrain type has placed a fresh emphasis on the selection of graphite for block manufacture in respect of physical properties. This is because the temperature distributions typical of such fuelled blocks lead to shutdown stresses close to the maximum the graphite can sustain without damage. Figures presented in this paper suggest that the physical properties of the graphite can play a relatively large part in reducing such stress levels and that guidance on the key requirements for suitable specifications is therefore particularly needed by the manufacturers of fuel block graphites. While graphites for fuel blocks have this special need for combinations of physical properties which lead to low thermal and shrinkage stresses, the other characteristics must also receive attention. A low graphite cost combined with good homogeneity in the brick, so that waste minimized, are still necessary, while isotropy is also very important

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

International Nuclear Information System (INIS)

Nickel, H.

1990-05-01

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

Analysis of neutronic parameters related to reduction in fuel rod diameter for Angra-1 reactor fuel elements

International Nuclear Information System (INIS)

Faria, Eduardo F.; Sadde, Luciano M.; Sakai, Massao; Gomes, Sydney da S.

2000-01-01

The actual fuel element design for Angra-1 PWR satisfies in a very conservative way the design limits established for the critical heat flux as well as for the energy stored in the fuel rod. However, that is not an optimized design under neutronic considerations. The conservative ratio of the H and U atomic densities gives rise to a harder neutron spectrum which reduces its reactivity. In this report, a reduction in fuel rod diameters has been analyzed, keeping however the same rod pitch for geometrical compatibility reasons. By increasing the H/U ratio it is possible to obtain a net gain in reactivity. The optimized diameter in its turn should not jeopardize the reactor safety requirements. The actual trends of the nuclear industry is to extend the cycles and the enrichment by using advanced fuel design. It must be emphasized that this design change gives rise to economical advantages, for example, reduced costs for uranium utilization and enrichment with a net gain in reactivity. (author)

Improved nuclear fuel element

International Nuclear Information System (INIS)

Klepfer, H.H.

1974-01-01

A nuclear fuel element is described which comprises: 1) an elongated clad container, 2) a layer of high lubricity material being disposed in and adjacent to the clad container, 3) a low neutron capture cross section metal liner being disposed in the clad container and adjacent to the layer, 4) a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, 5) an enclosure integrally secured and sealed at each end of the container, and a nuclear fuel material retaining means positioned in the cavity. (author)

Fabrication technology of spherical fuel element for HTR-10

International Nuclear Information System (INIS)

He Jun; Zou Yanwen; Liang Tongxiang; Qiu Xueliang

2002-01-01

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

Study on heat transfer and hydraulic model of spiral-fin fuel rods based on equivalent annulus method

International Nuclear Information System (INIS)

Zhang Dan; Liu Changwen; Lu Jianchao

2011-01-01

Tight lattice fuel assembly usually adopts spiral-fin fuel elements. Compared with the traditional PWR fuel rods, the closely packed and spiral fin spacers make the heat transfer and hydraulic phenomena in sub-channels very complicated, and: there was no suitable model and correlation to study it. This paper studied the effect of spiral spacers on the channel geometry in the equivalent annulus and physical performance based on the Rehme equivalent annulus methods, and the heat transfer of the spiral fin fuel rods and hydraulic model were obtained. The new model was verified with the traditional one, and the verification showed that two new models agreed well, which could provide certain theoretical explanation to the effect of the spiral spacer on the thermal hydraulics. (authors)

Carbon deposition on 20/25/Nb steel using an electrically heated AGR fuel pin

International Nuclear Information System (INIS)

Blanchard, A.; Campion, P.

1980-01-01

The radiolysis of carbon dioxide in gas-cooled reactors leads to the production of active species capable of reacting with the graphite moderator to form carbon monoxide with a resultant gradual loss of moderator. In the early days of gas-cooled reactor design, the intention was to allow the carbon monoxide concentration to increase and use this reaction product to inhibit the initial radiolysis of the carbon dioxide. Exploratory irradiation experiments using 4 to 7% carbon monoxide revealed that low density deposits ranging in colour from light grey through brown to black were found in the temperature range 470 to 600 K. In view of the fact that this type of deposition could adversely affect heat transfer processes in both fuel channels and heat exchangers, together with the fact that carbon monoxide was not sufficiently powerful as a graphite oxidation inhibitor, methane was selected as the primary inhibitor for the AGR series of power stations. This paper describes some carbon deposition experiments using an electrically heated 'dummy fuel element' linked to a recirculating carbon dioxide irradiation loop in which carbon monoxide concentration, methane concentration, fuel pin temperature and the chemical nature of the fuel pin surface were varied. (author)

MTR fuel element burn-up measurements by the reactivity method

International Nuclear Information System (INIS)

Zuniga, A.; Cuya, T.R.; Ravnik, M.

2003-01-01

Fuel element burn-up was measured by the reactivity method in the 10 MW Peruvian MTR reactor RP-10. The main purpose of the experiment was testing the reactivity method for an MTR reactor as the reactivity method was originally developed for TRIGA reactors. The reactivity worth of each measured fuel element was measured in its original core position in order to measure the burn-up of the fuel elements that were part of the experimental core. The burn-up of each measured fuel element was derived by interpolating its reactivity worth from the reactivity worth of two reference fuel elements of known burn-up, whose reactivity worth was measured in the position of the measured fuel element. The accuracy of the method was improved by separating the reactivity effect of burn-up from the effect of the position in the core. The results of the experiment showed that the modified reactivity method for fuel element burn-up determination could be applied also to MTR reactors. (orig.)

Design of fuel element for RA10

International Nuclear Information System (INIS)

Estevez, Esteban A.; Markiewicz, Mario; Gerding, Roberto

2012-01-01

The RA-10 reactor is an open pool multipurpose reactor. It is intended for radioisotopes production, fuel irradiation and use of neutron beam experiments. The nominal configuration core consists of 19 fuel elements (FE) and 6 in-core irradiation positions. With regard to the FE, although both conceptual design and manufacturing technology are similar to the already developed and qualified by CNEA (MTR fuel flat plate), the conditions imposed by the new reactor on FE's are more demanding that previous supplies. Here it should be mentioned the magnitude of the hydrodynamic forces acting on the FE caused by coolant flow through the core (upward) and mainly by the high coolant velocity between fuel plates (greater than 5 times than those currently in operation). Moreover, the high power density results in higher heat flux in fuel plates and greater temperature gradient. As a result of these increased demands present during irradiation, and in order to maintain a high level of reliability, it is necessary carry out some modifications in the mechanical design of the FE (with respect to the so-called ECBE design or s tandard ) . Design verification is performed through analytical and code calculations, and hydrodynamic tests on a full-scale prototype. This article describes the design of the FE for RA 10 reactor, with special emphasis on those aspects that represent innovations in the traditional design (ECBE). It also presents the functional requirements, design criteria and design limits established according to the reactor operational states (author)

CARA, new concept of advanced fuel element for HWR

International Nuclear Information System (INIS)

Florido, P.C.; Crimello, R.O.; Bergallo, J.E.; Marino, A.C.; Delmastro, D.F.; Brasnarof, D.O.; Gonzalez, J.H.

1999-01-01

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)

Fission product retention in TRISO coated UO2 particle fuels subjected to HTR simulated core heating tests

International Nuclear Information System (INIS)

Baldwin, C.A.; Kania, M.J.

1991-01-01

Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbounded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 deg. C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 deg. C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 deg. C and above may exist. (author). 6 refs, 6 figs, 4 tabs

Fission product retention in TRISCO coated UO2 particle fuels subjected to HTR simulated core heating tests

International Nuclear Information System (INIS)

Baldwin, C.A.; Kania, M.J.

1990-11-01

Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 degree C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 degree C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 degree C and above may exist. 6 refs., 6 figs., 4 tabs

A high temperature heating device for the study of fission product release from nuclear fuel

International Nuclear Information System (INIS)

Svedkauskaite-Le Gore, Jolanta; Kivel, Niko; Guenther-Leopold, Ines

2010-01-01

At the Paul Scherrer Institute a high temperature inductive heating furnace, which can heat fuel samples up to 2300 deg. C, has been developed in order to study the release of fission products. The furnace can be directly connected to an inductively coupled plasma mass spectrometer for online monitoring of the released elements and does not require their trapping before measurement. This paper describes the design of the inductive heating furnace, discusses its operating parameters, limitations and illustrates foreseen applications. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-03-01

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

Residual heat use generated by a 12 kW fuel cell in an electric vehicle heating system

International Nuclear Information System (INIS)

Colmenar-Santos, Antonio; Alberdi-Jiménez, Lucía; Nasarre-Cortés, Lorenzo; Mora-Larramona, Joaquín

2014-01-01

A diesel or gasoline vehicle heating is produced by the heat of the engine coolant liquid. Nevertheless, electric vehicles, due to the fact that electric motor transform directly electricity into mechanical energy through electromagnetic interactions, do not generate this heat so other method of providing it has to be developed. This study introduces the system developed in a fuel cell electric vehicle (lithium-ion battery – fuel cell) with residual heat use. The fuel cell electric vehicle is driven by a 12 kW PEM (proton exchange membrane) fuel cell. This fuel cell has an operating temperature around 50 °C. The residual heat generated was originally wasted by interaction with the environment. The new developed heating system designed integrates the heat generated by the fuel cell into the heating system of the vehicle, reducing the global energy consumption and improving the global efficiency as well. - Highlights: • Modification of heating system was done by introducing the residual heat from fuel cell. • Maximum heat achieved by the heating radiator of 9.27 kW. • Reduction of the heat dissipation by the fuel cell cooling system 1.5 kW. • Total efficiency improvement of 20% with an autonomy increase of 21 km

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

International Nuclear Information System (INIS)

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

1999-01-01

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 3 O 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

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.

HTGR fuel element structural design consideration

International Nuclear Information System (INIS)

Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

1987-01-01

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. Probabilistic stress analysis techniques coupled with probabilistic 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 consistant with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the U.S.A. is discussed in the context of stress analysis uncertainty and structural criteria development. (author)

HTGR fuel element structural design considerations

International Nuclear Information System (INIS)

Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

1986-09-01

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

Thermal conductivity of U–Mo/Al dispersion fuel. Effects of particle shape and size, stereography, and heat generation

International Nuclear Information System (INIS)

Cho, Tae Won; Sohn, Dong-Seong; Kim, Yeon Soo

2015-01-01

This paper describes the effects of particle sphericity, interfacial thermal resistance, stereography, and heat generation on the thermal conductivity of U–Mo/Al dispersion fuel. The ABAQUS finite element method (FEM) tool was used to calculate the effective thermal conductivity of U–Mo/Al dispersion fuel by implementing fuel particles. For U–Mo/Al, the particle sphericity effect was insignificant. However, if the effect of the interfacial thermal resistance between the fuel particles and Al matrix was considered, the thermal conductivity of U–Mo/Al was increased as the particle size increases. To examine the effect of stereography, we compared the two-dimensional modeling and three-dimensional modeling. The results showed that the two-dimensional modeling predicted lower than the three-dimensional modeling. We also examined the effect of the presence of heat sources in the fuel particles and found a decrease in thermal conductivity of U–Mo/Al from that of the typical homogeneous heat generation modeling. (author)

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

Science.gov (United States)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are discussed. The authors demonstrated success in reaching desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and define a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Statistical estimation of fast-reactor fuel-element lifetime

International Nuclear Information System (INIS)

Proshkin, A.A.; Likhachev, Yu.I.; Tuzov, A.N.; Zabud'ko, L.M.

1980-01-01

On the basis of a statistical analysis, the main parameters having a significant influence on the theoretical determination of fuel-element lifetimes in the operation of power fast reactors in steady power conditions are isolated. These include the creep and swelling of the fuel and shell materials, prolonged-plasticity lag, shell-material corrosion, gap contact conductivity, and the strain diagrams of the shell and fuel materials obtained for irradiated materials at the corresponding strain rates. By means of deeper investigation of these properties of the materials, it is possible to increase significantly the reliability of fuel-element lifetime predictions in designing fast reactors and to optimize the structure of fuel elements more correctly. The results of such calculations must obviously be taken into account in the cost-benefit analysis of projected new reactors and in choosing the optimal fuel burnup. 9 refs

Fuel Composition and Performance Analysis of Endothermically Heated Fuels for Pulse Detonation Engines

Science.gov (United States)

2009-03-01

exchanger was constructed on an inner 2 in Inconel 625 schedule 10 pipe and an outer 2 ½ in Inconel 600 schedule 40 pipe 0.91 m (36 in) in length. The...switched to positions two and three for the remainder of the experiments. 46 The detonation tubes are fabricated from inconel and include heat...and four. Fuel Heating System 47 The fuel heating system centers around two pairs of inconel heat exchangers. The first pair was developed in

Analysis of pre-heated fuel combustion and heat-emission dynamics in a diesel engine

Science.gov (United States)

Plotnikov, S. A.; Kartashevich, A. N.; Buzikov, S. V.

2018-01-01

The article explores the feasibility of diesel fuel pre-heating. The research goal was to obtain and analyze the performance diagrams of a diesel engine fed with pre-heated fuel. The engine was tested in two modes: at rated RPMs and at maximum torque. To process the diagrams the authors used technique developed by the Central Diesel Research Institute (CDRI). The diesel engine’s heat emission curves were obtained. The authors concluded that fuel pre-heating shortened the initial phase of the combustion process and moderated the loads, thus making it possible to boost a diesel engine’s mean effective pressure.

Apparatus and method for assembling fuel elements

International Nuclear Information System (INIS)

Arya, S.P.

1978-01-01

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

Upgraded HFIR Fuel Element Welding System

International Nuclear Information System (INIS)

Sease, John D.

2010-01-01

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.

Advancements in the behavioral modeling of fuel elements and related structures

International Nuclear Information System (INIS)

Billone, M.C.; Montgomery, R.O.; Rashid, Y.R.; Head, J.L.

1989-01-01

An important aspect of the design and analysis of nuclear reactors is the ability to predict the behavior of fuel elements in the adverse environment of a reactor system. By understanding the thermomechanical behavior of the different materials which constitute a nuclear fuel element, analysis and predictions can be made regarding the integrity and reliability of fuel element designs. The SMiRT conference series, through the division on fuel elements and the post-conference seminars on fuel element modeling, provided technical forums for the international participation in the exchange of knowledge concerning the thermomechanical modeling of fuel elements. This paper discusses the technical advances in the behavioral modeling of fuel elements presented at the SMiRT conference series since its inception in 1971. Progress in the areas of material properties and constitutive relationships, modeling methodologies, and integral modeling approaches was reviewed and is summarized in light of their impact on the thermomechanical modeling of nuclear fuel elements. 34 refs., 5 tabs

Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

International Nuclear Information System (INIS)

Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

1993-01-01

Three furnace heating tests were conducted with irradiated, HT9-clad and U-19wt%Pu-10wt%Zr-alloy, EBR-II Mk-V-type fuel elements to evaluate the behavior that could be expected during a loss-of-flow event in the reactor. In general, very significant safety margins for cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results are presented, as are discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction that were found in these tests. (orig.)

Safety assessment for the CANFLEX-NU fuel bundles with respect to the 37-element fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Suk, H. C.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-11-01

The KAERI and AECL have jointly developed an advanced CANDU fuel, called CANFLEX-NU fuel bundle. CANFLEX 43-element bundle has some improved features of increased operating margin and enhanced safety compared to the existing 37-element bundle. Since CANFLEX fuel bundle is designed to be compatible with the CANDU-6 reactor design, the behaviour in the thermalhydraulic system will be nearly identical with 37-element bundle. But due to different element design and linear element power distribution between the two bundles, it is expected that CANFLEX fuel behaviour would be different from the behaviour of the 37-element fuel. Therefore, safety assessments on the design basis accidents which result if fuel failures are performed. For all accidents selected, it is observed that the loading of CANFLEX bundle in an existing CANDU-6 reactor would not worsen the reactor safety. It is also predicted that fission product release for CANFLEX fuel bundle generally is lower than that for 37-element bundle. 3 refs., 2 figs., 2 tabs. (Author)

Detection and location of leaking TRIGA fuel elements

International Nuclear Information System (INIS)

Bouchey, G.D.; Gage, S.J.

1970-01-01

Several TRIGA facilities have experienced difficulty resulting from cladding failures of aluminum clad TRIGA fuel elements. Recently, at the University of Texas at Austin reactor facility, fission product releases were observed during 250 kW operation and were attributed to a leaking fuel element. A rather extensive testing program has been undertaken to locate the faulty element. The used sniffer device is described, which provides a quick, easily constructed, and extremely sensitive means of locating leaking fuel elements. The difficulty at The University of Texas was compounded by extremely low levels and the sporadic nature of the releases. However, in the more typical situation, in which a faulty element consistently releases relatively large quantities of fission gas, such a device should locate the leak with little difficulty

Features of spherical uranium-graphite HTGR fuel elements control

International Nuclear Information System (INIS)

Kreindlin, I.I.; Oleynikov, P.P.; Shtan, A.S.

1985-01-01

Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described

Features of spherical uranium-graphite HTGR fuel elements control

Energy Technology Data Exchange (ETDEWEB)

Kreindlin, I I; Oleynikov, P P; Shtan, A S

1985-07-01

Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described.

Experimental study on heat pipe heat removal capacity for passive cooling of spent fuel pool

International Nuclear Information System (INIS)

Xiong, Zhenqin; Wang, Minglu; Gu, Hanyang; Ye, Cheng

2015-01-01

Highlights: • A passively cooling SFP heat pipe with an 8.2 m high evaporator was tested. • Heat removed by the heat pipe is in the range of 3.1–16.8 kW. • The heat transfer coefficient of the evaporator is 214–414 W/m 2 /K. • The heat pipe performance is sensitive to the hot water temperature. - Abstract: A loop-type heat pipe system uses natural flow with no electrically driven components. Therefore, such a system was proposed to passively cool spent fuel pools during accidents to improve nuclear power station safety especially for station blackouts such as those in Fukushima. The heat pipe used for a spent fuel pool is large due to the spent fuel pool size. An experimental heat pipe test loop was developed to estimate its heat removal capacity from the spent fuel pool during an accident. The 7.6 m high evaporator is heated by hot water flowing vertically down in an assistant tube with a 207-mm inner diameter. R134a was used as the potential heat pipe working fluid. The liquid R134a level was 3.6 m. The tests were performed for water velocities from 0.7 to 2.1 × 10 −2 m/s with water temperatures from 50 to 90 °C and air velocities from 0.5 m/s to 2.5 m/s. The results indicate significant heat is removed by the heat pipe under conditions that may occur in the spent fuel pool

End-of-season heating fuel report

International Nuclear Information System (INIS)

1992-01-01

The year-end report notes that the 1991-92 heating season had lower average oil prices (retail home heating fuel) than the past two winters and prices remained relatively stable throughout the season. This year, the heating season average was $.87 per gallon, $1.05 for kerosene, and $1.33 for propane

Evaluation of gap heat transfer model in ELESTRES for CANDU fuel element under normal operating conditions

International Nuclear Information System (INIS)

Lee, Kang Moon; Ohn, Myung Ryong; Im, Hong Sik; Choi, Jong Hoh; Hwang, Soon Taek

1995-01-01

The gap conductance between the fuel and the sheath depends strongly on the gap width and has a significant influence on the amount of initial stored energy. The modified Ross and Stoute gap conductance model in ELESTRES is based on a simplified thermal deformation model for steady-state fuel temperature calculations. A review on a series of experiments reveals that fuel pellets crack, relocate, and are eccentrically positioned within the sheath rather than solid concentric cylinders. In this paper, the two recently-proposed gap conductance models (offset gap model and relocated gap model) are described and are applied to calculate the fuel-sheath gap conductances under experimental conditions and normal operating conditions in CANDU reactors. The good agreement between the experimentally-inferred and calculated gap conductance values demonstrates that the modified Ross and Stoute model was implemented correctly in ELESTRES. The predictions of the modified Ross and Stoute model provide conservative values for gap heat transfer and fuel surface temperature compared to the offset gap and relocated gap models for a limiting power envelope. 13 figs., 3 tabs., 16 refs. (Author)

Quality control in the fuel elements production process

International Nuclear Information System (INIS)

Katanic-Popovic, J.; Spasic, Z.; Djuricis, Lj.

1977-01-01

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)

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

International Nuclear Information System (INIS)

Khafizov, R.R.; Poplavskij, V.M.; Rachkov, V.I.; Sorokin, A.P.; Ashurko, Yu.M.; Volkov, A.V.; Ivanov, E.F.; Privezentsev, V.V.

2015-01-01

The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m 2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling [ru

Investigation Status of Heat Exchange while Boiling Hydrocarbon Fuel

Directory of Open Access Journals (Sweden)

D. S. Obukhov

2006-01-01

Full Text Available The paper contains analysis of heat exchange investigations while boiling hydrocarbon fuel. The obtained data are within the limits of the S.S. Kutateladze dependence proposed in 1939. Heat exchange at non-stationary heat release has not been investigated. The data for hydrocarbon fuel with respect to critical density of heat flow are not available even for stationary conditions.

Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

International Nuclear Information System (INIS)

Shimooke, Takanori; Yoshida, Kazuo

1979-05-01

The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

Hot fuel examination facility element spacer wire-wrap machine

International Nuclear Information System (INIS)

Tobias, D.A.; Sherman, E.K.

1989-01-01

Nondestructive examinations of irradiated experimental fuel elements conducted in the Argonne National Laboratory Hot Fuel Examination Facility/North (HFEF/N) at the Idaho National Engineering Laboratory include laser and contact profilometry (element diameter measurements), electrical eddy-current testing for cladding and thermal bond defects, bow and length measurements, neutron radiography, gamma scanning, remote visual exam, and photography. Profilometry was previously restricted to spiral profilometry of the element to prevent interference with the element spacer wire wrapped in a helix about the Experimental Breeder Reactor II (EBR-II)-type fuel element from end to end. By removing the spacer wire prior to conducting profilometry examination, axial profilometry techniques may be used, which are considerably faster than spiral techniques and often result in data acquisition more important to experiment sponsors. Because the element must often be reinserted into the nuclear reactor (EBR-II) for additional irradiation, however, the spacer wire must be reinstalled on the highly irradiated fuel element by remote means after profilometry of the wireless elements. The element spacer wire-wrap machine developed at HFEF is capable of helically wrapping fuel elements with diameters up to 1.68 cm (0.660 in.) and 2.44-m (96-in.) lengths. The machine can accommodate almost any desired wire pitch length by simply inserting a new wrapper gear module

Review of direct electrical heating experiments on irradiated mixed-oxide fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Bandyopadhyay, G.

1982-01-01

Results of approximately 50 out-of-reactor experiments that simulated various stages of a loss-of-flow event with irradiated fuel are presented. The tests, which utilized the direct electrical heating technique to simulate nuclear heating, were performed either on fuel segments with their original cladding intact or on fuel segments that were extruded into quartz tubes. The test results demonstrated that the macro- and microscopic fuel behavior was dependent on a number of variables including fuel heating rate, thermal history prior to a transient, the number of heating cycles, type of cladding (quartz vs stainless steel), and fuel burnup

Generation of heat on fuel rod in cosine pattern by using induction heating

International Nuclear Information System (INIS)

Keettikkal, Felix; Sajeesh, Divya; Rao, Poornima; Hande, Shashank; Dakave, Ganesh; Kute, Tushar; Mahajan, Akshay; Kulkarni, R.D.

2017-01-01

Fuel rods are used in a nuclear reactor for fission process. When these rods are cooled by water during the heat transfer, the temperature stress causes undesirable defects in the fuel rod. Studying these defects occurring in the fuel rod in the nuclear cluster during nuclear reaction is a difficult task because fission reaction makes it difficult to analyse the changes in the rod. Hence there is a need to use a replica of the rod with similar thermal stress to study and analyse the rod for the defects. Normally the heat generated on the fuel rod follows a cosine pattern which is an inherent characteristic inside a nuclear reactor. In view of this, in this paper induction heating method is used on a rod to create an exact replica of the cosine pattern of heat by varying the pitch of the coil. First, a MATLAB simulation is done using simulink. Then a prototype of the model has been developed comprising of carbon steel pipe, with length and outside diameter of 1 meter and 48.2 mm, respectively. Instead of using water as coolant, rod is simulated in air. Therefore, the heat generated is lost by normal convection and radiation. Non-nuclear testing can be a valuable tool in the development or in some kind of experiment using nuclear reactor. Induction heating becomes an alternative to classical heating technologies because of its advantages such as efficiency, quickness, safety, clean heating and accurate power control. (author)

Fresh fuel pre-heating device in reactor facility

International Nuclear Information System (INIS)

Samejima, Asakuni.

1988-01-01

Purpose: To simplify the structure of a fresh nuclear fuel pre-heating device and improve the reliability to gas supply. Constitution: Fresh fuels taken out from a fresh fuel stredge rack and contained in a fuel strage pipe of a fuel transportation cask are pre-heated at the pre-stage of transfer by sending heating gases from the outside. Gas outlet pipes of the device are led out from the lower portion of the strage pipe, disposed side by side at the top of the strage pipe and opened upwardly. Further, gas supply pipes are connected to the inside of a movable guiding cylinder on the side of the floor surface and the opening end of return pipes are opposed to the exit opening end of the strage pipe. In such a constitution, a gas recycling loop can be formed between the strage pipe and the gas heating device by way of the movable guiding cylinder only by the operation of combining the fuel strage pipe of the transportation cask and the movable guiding pipe disposed on the side of the floor surface. Thus, the coupling structure is facilitated, the connection operation can surely be conducted to improve the reliability as compared with the conventional case. (Horiuchi, T.)

Nuclear fuel safety studies by laser pulse heating

International Nuclear Information System (INIS)

Viswanadham, C.S.; Kumar, Santosh; Dey, G.K.; Kutty, T.R.G.; Khan, K.B.; Kumar, Arun; Jathar, V.P.; Sahoo, K.C.

2009-01-01

The behaviour of nuclear fuels under transient heating conditions is vital to nuclear safety. A laser pulse based heating system to simulate the transient heating conditions experienced by the fuel during reactor accidents like LOCA and RIA is under development at BARC, Mumbai. Some of the concepts used in this system are under testing in pilot studies. This paper describes the results of some pilot studies carried out on unirradiated UO 2 specimens by laser pulse heating, followed by metallography and X-ray diffraction measurements. (author)

Methodology for substantiation of the fast reactor fuel element serviceability

International Nuclear Information System (INIS)

Tsykanov, V.A.; Maershin, A.A.

1988-01-01

Methodological aspects of fast reactor fuel element serviceability substantiation are presented. The choice of the experimental program and strategies of its realization to solve the problem set in short time, taking into account available experimental means, are substantiated. Factors determining fuel element serviceability depending on parameters and operational conditions are considered. The methodological approach recommending separate studing of the factors, which points to the possibility of data acquisition, required for the development of calculational models and substantiation of fuel element serviceability in pilot and experimental reactors, is described. It is shown that the special-purpose data are more useful for the substantiation of fuel element serviceability and analytical method development than unsubstantial and expensive complex tests of fuel elements and fuel assemblies, which should be conducted only at final stages for the improvement of the structure on the whole

Program for responsible and safe disposal of spent fuel elements and radioactive wastes (National disposal program)

International Nuclear Information System (INIS)

2015-01-01

The contribution covers the following topics: fundamentals of the disposal policy; amount of radioactive wastes and prognosis; disposal of radioactive wastes - spent fuel elements and wastes from waste processing, radioactive wastes with low heat production; legal framework of the nuclear waste disposal in Germany; public participation, cost and financing.

Nuclear fuel preheating system

International Nuclear Information System (INIS)

Andrea, C.

1975-01-01

A nuclear reactor new fuel handling system which conveys new fuel from a fuel preparation room into the reactor containment boundary is described. The handling system is provided with a fuel preheating station which is adaptd to heat the new fuel to reactor refueling temperatures in such a way that the fuel is heated from the top down so that fuel element cladding failure due to thermal expansions is avoided. (U.S.)

Gap's dimensions in fuel elements from neutron radiography

International Nuclear Information System (INIS)

Notea, A.; Segal, Y.; Trichter, F.

1985-01-01

Quantitative Nondestructive evaluation (QNDE) is of upmost importance in the design and manufacture of nuclear fuel elements. Accurate non-destructive measurements of gaps, cracks, displacements, etc. supply vital information for optimizing fuel manufacturing. Neutron radiography is a powerful NDT method for examining spent fuel elements. However, it turned out that the extraction of dimensions, especially in the submillimetric range is questionable. In this paper neutron radiography of pellet-to-pellet gaps in fuel elements is modelled and two procedures for dimension extraction are presented. It is shown that for a wide gap the dimension is preferable, extracted from the width of the film profile, while for narrow gaps it is preferable to extract it from the maximum of the density profile

POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

Directory of Open Access Journals (Sweden)

Vojtěch Caha

2016-12-01

Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

Near-field heat transfer at the spent fuel test-climax: a comparison of measurements and calculations

International Nuclear Information System (INIS)

Patrick, W.C.; Montan, D.N.; Ballou, L.B.

1981-01-01

The Spent Fuel Test in the Climax granitic stock at the DOE Nevada Test Site is a test of the feasibility of storage and retrieval of spent nuclear reactor fuel in a deep geologic environment. Eleven spent fuel elements, together with six thermally identical electrical resistance heaters and 20 peripheral guard heaters, are emplaced 420 m below surface in a three-drift test array. This array was designed to simulate the near-field effects of thousands of canisters of nuclear waste and to evaluate the effects of heat alone, and heat plus ionizing radiation on the rock. Thermal calculations and measurements are conducted to determine thermal transport from the spent fuel and electrical resistance heaters. Calculations associated with the as-built Spent Fuel Test geometry and thermal source histories are presented and compared with thermocouple measurements made throughout the test array. Comparisons in space begin at the spent fuel canister and include the first few metres outside the test array. Comparisons in time begin at emplacement and progress through the first year of thermal loading in this multi-year test

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

International Nuclear Information System (INIS)

Gurin, V.A.; Konotop, Yu.F.; Odejchuk, N.P.; Shirochenkov, S.D.; Yakovlev, V.K.; Aksenov, N.A.; Kuprienko, V.A.; Lebedev, I.G.; Samsonov, B.V.

1990-01-01

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

Nuclear fuel element, and method of producing same

International Nuclear Information System (INIS)

Armijo, J.S.; Esch, E.L.

1986-01-01

This invention relates to an improvement in nuclear fuel elements having a composite container comprising a cladding sheath provided with a protective barrier of zirconium metal covering the inner surface of the sheath, rendering such fuel elements more resistant to hydrogen accumulation in service. The invention specifically comprises removing substantially all zirconium metal of the barrier layer from the part of the sheath surrounding and defining the plenum region. Thus the protective barrier of zirconium metal covers only the inner surface of the fuel container in the area immediately embracing the fissionable fuel material

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.

Energy efficiency of electrical infrared heating elements

International Nuclear Information System (INIS)

Brown, K.J.; Farrelly, R.; O’Shaughnessy, S.M.; Robinson, A.J.

2016-01-01

Highlights: • Characterization of the radiant energy efficiency of infrared heating elements. • Performed for a commercially available ceramic heater element for two cases. • Total radiant power and net radiant efficiency is computed. • Radiant efficiencies are strongly dependant on the input power to the element. • In-plane efficiencies depend on the distance from the heater. - Abstract: A measurement system has been designed to characterize the radiant energy efficiency of infrared heating elements. The system also allows for measurement of the radiant heat flux distribution emitted from radiant heater assemblies. To facilitate these, a 6-axis robotic arm is fitted with a Schmidt–Boelter radiant heat flux gauge. A LabVIEW interface operates the robot and positions the sensor in the desired location and subsequently acquires the desired radiant heat flux measurement. To illustrate the functionality of the measurement system and methodology, radiant heat flux distributions and efficiency calculations are performed for a commercially available ceramic heater element for two cases. In the first, a spherical surface is traced around the entire heater assembly and the total radiant power and net radiant efficiency is computed. In the second, 50 cm × 50 cm vertical planes are traced parallel to the front face of the heater assembly at distances between 10 cm and 50 cm and the in-plane power and efficiencies are computed. The results indicate that the radiant efficiencies are strongly dependant on the input power to the element and, for the in-plane efficiencies, depend on the distance from the heater.

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

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

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

Behavior of mixed-oxide fuel elements during an overpower transient

International Nuclear Information System (INIS)

Tsai, H.; Shikakura, S.

1993-01-01

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.)

Fuel Cycle Concept with Advanced METMET and Composite Fuel in LWRs

International Nuclear Information System (INIS)

Savchenko, A.; Skupov, M.; Vatulin, A.; Glushenkov, A.; Kulakov, G.; Lipkina, K.

2014-01-01

The basic factor that limits the serviceability of fuel elements developing in the framework of RERTR Program (transition from HEU to LEU fuel of research reactors) is interaction between U10Mo fuel and aluminium matrix . Interaction results in extra swelling of fuels, disappearance of a heat conducting matrix, a temperature rise in the fuel centre, penetration porosity, etc. Several methods exist to prevent fuel-matrix interaction. In terms of simplifying fuel element fabrication technology and reducing interaction, doping of fuel is the most optimal version

Irradiation of MEU and LEU test fuel elements in DR 3

International Nuclear Information System (INIS)

Haack, K.

1984-01-01

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)

Hydraulic modelling of the CARA Fuel element

International Nuclear Information System (INIS)

Brasnarof, Daniel O.; Juanico, Luis; Giorgi, M.; Ghiselli, Alberto M.; Zampach, Ruben; Fiori, Jose M.; Yedros, Pablo A.

2004-01-01

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 5x10 4 and 1,5x10 5 ) 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) [es

Improved nuclear fuel element

International Nuclear Information System (INIS)

1974-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a metal liner disposed between the cladding and the nuclear fuel material and a high lubricity material in the form of a coating disposed between the liner and the cladding. The liner preferably has a thickness greater than the longest fission product recoil distance and is composed of a low neutron capture cross-section material. The liner is preferably composed of zirconium, an alloy of zirconium, niobium or an alloy of niobium. The liner serves as a preferential reaction site for volatile impurities and fission products and protects the cladding from contact and reaction with such impurities and fission products. The high lubricity material acts as an interface between the liner and the cladding and reduces localized stresses on the cladding due to fuel expansion and cracking of the fuel

SAFE: A computer code for the steady-state and transient thermal analysis of LMR fuel elements

International Nuclear Information System (INIS)

Hayes, S.L.

1993-12-01

SAFE is a computer code developed for both the steady-state and transient thermal analysis of single LMR fuel elements. The code employs a two-dimensional control-volume based finite difference methodology with fully implicit time marching to calculate the temperatures throughout a fuel element and its associated coolant channel for both the steady-state and transient events. The code makes no structural calculations or predictions whatsoever. It does, however, accept as input structural parameters within the fuel such as the distributions of porosity and fuel composition, as well as heat generation, to allow a thermal analysis to be performed on a user-specified fuel structure. The code was developed with ease of use in mind. An interactive input file generator and material property correlations internal to the code are available to expedite analyses using SAFE. This report serves as a complete design description of the code as well as a user's manual. A sample calculation made with SAFE is included to highlight some of the code's features. Complete input and output files for the sample problem are provided

Temperature Analysis and Failure Probability of the Fuel Element in HTR-PM

International Nuclear Information System (INIS)

Yang Lin; Liu Bing; Tang Chunhe

2014-01-01

Spherical fuel element is applied in the 200-MW High Temperature Reactor-Pebble-bed Modular (HTR-PM). Each spherical fuel element contains approximately 12,000 coated fuel particles in the inner graphite matrix with a diameter of 50mm to form the fuel zone, while the outer shell with a thickness of 5mm is a fuel-free zone made up of the same graphite material. Under high burnup irradiation, the temperature of fuel element rises and the stress will result in the damage of fuel element. The purpose of this study is to analyze the temperature of fuel element and to discuss the stress and failure probability. (author)

Fire and blast safety manual for fuel element manufacture

International Nuclear Information System (INIS)

Ensinger, U.; Koehler, B.; Mester, W.; Riotte, H.G.; Sehrbrock, H.W.

1988-01-01

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 UO 2 , MOX and HTR fuel elements. (orig./DG) [de

Development of Passive Fuel Cell Thermal Management Heat Exchanger

Science.gov (United States)

Burke, Kenneth A.; Jakupca, Ian J.; Colozza, Anthony J.

2010-01-01

The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA Exploration program. The passive thermal management system relies on heat conduction within highly thermally conductive cooling plates to move the heat from the central portion of the cell stack out to the edges of the fuel cell stack. Using the passive approach eliminates the need for a coolant pump and other cooling loop components within the fuel cell system which reduces mass and improves overall system reliability. Previous development demonstrated the performance of suitable highly thermally conductive cooling plates that could conduct the heat, provide a sufficiently uniform temperature heat sink for each cell of the fuel cell stack, and be substantially lighter than the conventional thermal management approach. Tests were run with different materials to evaluate the design approach to a heat exchanger that could interface with the edges of the passive cooling plates. Measurements were made during fuel cell operation to determine the temperature of individual cooling plates and also to determine the temperature uniformity from one cooling plate to another.

The Calculation Of Total Radioactivity Of Kartini Reactor Fuel Element

International Nuclear Information System (INIS)

Budisantoso, Edi Trijono; Sardjono, Y.

1996-01-01

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

Ignition of an organic water-coal fuel droplet floating in a heated-air flow

Science.gov (United States)

Valiullin, T. R.; Strizhak, P. A.; Shevyrev, S. A.; Bogomolov, A. R.

2017-01-01

Ignition of an organic water-coal fuel (CWSP) droplet floating in a heated-air flow has been studied experimentally. Rank B2 brown-coal particles with a size of 100 μm, used crankcase Total oil, water, and a plasticizer were used as the main CWSP components. A dedicated quartz-glass chamber has been designed with inlet and outlet elements made as truncated cones connected via a cylindrical ring. The cones were used to shape an oxidizer flow with a temperature of 500-830 K and a flow velocity of 0.5-5.0 m/s. A technique that uses a coordinate-positioning gear, a nichrome thread, and a cutter element has been developed for discharging CWSP droplets into the working zone of the chamber. Droplets with an initial size of 0.4 to 2.0 mm were used. Conditions have been determined for a droplet to float in the oxidizer flow long enough for the sustainable droplet burning to be initiated. Typical stages and integral ignition characteristics have been established. The integral parameters (ignition-delay times) of the examined processes have been compared to the results of experiments with CWSP droplets suspended on the junction of a quick-response thermocouple. It has been shown that floating fuel droplets ignite much quicker than the ones that sit still on the thermocouple due to rotation of an CWSP droplet in the oxidizer flow, more uniform heating of the droplet, and lack of heat drainage towards the droplet center. High-speed video recording of the peculiarities of floatation of a burning fuel droplet makes it possible to complement the existing models of water-coal fuel burning. The results can be used for a more substantiated modeling of furnace CWSP burning with the ANSYS, Fluent, and Sigma-Flow software packages.

In-core fuel element temperature and flow measurment of HFETR

International Nuclear Information System (INIS)

Chen Daolong; Jiang Pei

1988-02-01

The HFETR in-core fuel element temperature-flow measurement facility and its measurement system are expounded. The applications of the instrumented fuel element to stationary and transient states measurements during the lift of power, the operation test of all lifetime at first load, and the deepening burn-up test at second load are described. The method of determination of the hot point temperature under the fin is discussed. The error analysis is made. The fuel element out-of-pile water deprivation test is described. The development of this measurement facility and succesful application have made important contribution to high power and deep burn-up safe operation at two load, in-core fuel element irradiation, and varied investigation of HFETR. After operation at two loads, the integrated power of this instrumented fuel element arrives at 90.88 MWd, its maximum point burn-up is about 64.9%, so that the economy of fuel use of HFETR is raised very much

Analysis of the ATR fuel element swaging process

International Nuclear Information System (INIS)

Richins, W.D.; Miller, G.K.

1995-12-01

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

Inspection of fuel elements in the cooling pond of a research reactor

International Nuclear Information System (INIS)

Pavlov, S.V.; Mestnikov, A.V.

1992-01-01

Nondestructive testing methods for fuel bundles and fuel elements in the cooling ponds of atomic power plants, using special inspection stands, have come into widespread use during the past decade. This paper describes a methodological stand that was built for the laboratory development of methods and individual units of inspection stands for fuel bundles of RBMK and VVER-1000 reactors. A complex of equipment was developed for the study of irradiated fuel elements, thus creating a methodological base for developing techniques for nondestructive testing of irradiated fuel elements and equipment to obtain information about the state of the fuel elements in a reactor expeditiously. The time required to inspect a fuel element can be shortened using some techniques simultaneously. The length of a fuel element can be measured simultaneously with visual inspection, eddy-current flaw detection can be preformed at the same time as the tranverse size of the fuel element is being determined. 6 refs., 5 figs

HTGR fuel element size reduction system

International Nuclear Information System (INIS)

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

The calculation - experimental investigations of the HTGR fuel element construction

International Nuclear Information System (INIS)

Eremeev, V.S.; Kolesov, V.S.; Chernikov, A.S.

1985-01-01

One of the most important problems in the HTGR development is the creation of the fuel element gas-tight for the fission products. This problem is being solved by using fuel elements of dispersion type representing an ensemble of coated fuel particles dispersed in the graphite matrix. Gas-tightness of such fuel elements is reached at the expense of deposing a protective coating on the fuel particles. It is composed of some layers serving as diffusion barriers for fission products. It is apparent that the rate of fission products diffusion from coated fuel particles is determined by the strength and temperature of the protective coating

Modeling of PHWR fuel elements using FUDA code

International Nuclear Information System (INIS)

Tripathi, Rahul Mani; Soni, Rakesh; Prasad, P.N.; Pandarinathan, P.R.

2008-01-01

The computer code FUDA (Fuel Design Analysis) is used for modeling PHWR fuel bundle operation history and carry out fuel element thermo-mechanical analysis. The radial temperature profile across fuel and sheath, fission gas release, internal gas pressure, sheath stress and strains during the life of fuel bundle are estimated

Space reactor fuel element testing in upgraded TREAT

International Nuclear Information System (INIS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ∼60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ∼100 MW/L may be achievable

Space reactor fuel element testing in upgraded TREAT

Science.gov (United States)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

Ultimate storage of spent fuel elements of the AVR test power station in the Asse salt mine

International Nuclear Information System (INIS)

Wolf, J.

1975-02-01

With regard to the ultimate storage of irradiated AVR pebble-bed reactor carbide fuel elements in the saline of Asse, a number of tests and calculations are presented to demonstrate that there is no credible possibility of the MCA (maximum credible accident) defined for the saline. The safety of persons is not threatened during the operation of spent fuel storage nor at any later time (extrapolation up to approx. 1,000 years after storage). 1,000 fuel elements at a time are packed up in gas-tight containers which are stacked in boreholes. The boreholes are then sealed with concrete. Lay-out and functions of the special airlock and transportation systems - from the packing of the containers in a hot cell to the final storage in the borehole - are described with special reference to aspects of the safety of the overall procedure. The possible accidents in the mine are discussed in detail. 85 Kr and T release rates are determined in laboratory tests by heating of the spherical fuel elements. Tests with fuel elements embedded in salt or stagnant brine were carried out at varies temperatures to investigate their behaviour in final storage. Kr and T release, extraction of fission products, mechanical resistance and corrosion were examined in these tests. Finally, the permeability of salt and salt concrete to radioactive gases were investigated in a special experimental arrangement. The diffusion and permeability coefficients obtained for 85 Kr, HT and HTO allow an estimation of the gas discharge of the stored fuel element. (RB/AK) [de

Configuring a fuel cell based residential combined heat and power system

Science.gov (United States)

Ahmed, Shabbir; Papadias, Dionissios D.; Ahluwalia, Rajesh K.

2013-11-01

The design and performance of a fuel cell based residential combined heat and power (CHP) system operating on natural gas has been analyzed. The natural gas is first converted to a hydrogen-rich reformate in a steam reformer based fuel processor, and the hydrogen is then electrochemically oxidized in a low temperature polymer electrolyte fuel cell to generate electric power. The heat generated in the fuel cell and the available heat in the exhaust gas is recovered to meet residential needs for hot water and space heating. Two fuel processor configurations have been studied. One of the configurations was explored to quantify the effects of design and operating parameters, which include pressure, temperature, and steam-to-carbon ratio in the fuel processor, and fuel utilization in the fuel cell. The second configuration applied the lessons from the study of the first configuration to increase the CHP efficiency. Results from the two configurations allow a quantitative comparison of the design alternatives. The analyses showed that these systems can operate at electrical efficiencies of ∼46% and combined heat and power efficiencies of ∼90%.

Mechanisms of the initial stage of fuel elements degradation of BN reactor fuel assemblies

International Nuclear Information System (INIS)

Zagorul'ko, Yu.I.; Kashcheev, M.V.; Ganichev, N.S.

2015-01-01

On the base of developed calculational technique numerical evaluation is carried out to the time of fuel element fracture in conditions of loss of sodium flow through fuel element jacket. Data on mechanical properties of steel EhK-164 is used in calculations. Calculations are carried out for different conditions of jacket outer surface cooling: by sodium of 1073 K temperature, by boiling sodium and by sodium in condition of film boiling. It is shown that time to jacket fracture under considered rupture mechanisms essentially depends on fuel element cooling conditions [ru

Process for changing fuel elements of a water-cooled nuclear reactor

International Nuclear Information System (INIS)

Fleischmann, R.; Rau, P.

1986-01-01

In order to change fuel elements, a water-filled duct can be installed between the rector pressure vessel and a space for accommodating the fuel elements. The fuel elements are transported there under water by a fuelling machine. The duct is installed as watertight connection closed on all sides between the reactor pressure vessel and a fuel element transport container brought close to it. The fuelling machine works in this duct. (orig./HP) [de

Laser assisted decontamination of nuclear fuel elements

International Nuclear Information System (INIS)

Padma Nilaya, J.; Biswas, Dhruba J.; Kumar, Aniruddha

2010-04-01

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)

End plug welding of nuclear fuel elements-AFFF experience

International Nuclear Information System (INIS)

Bhatt, R.B.; Singh, S.; Aniruddha Kumar; Amit; Arun Kumar; Panakkal, J.P.; Kamath, H.S.

2004-01-01

Advanced Fuel Fabrication Facility is engaged in the fabrication of mixed oxide (U,Pu)O 2 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

Spacer grid for fuel elements

International Nuclear Information System (INIS)

Hensolt, T.; Huenner, M.; Rau, P.; Veca, A.

1978-01-01

The spacer grid for fuel elements of a gas-cooled fast breeder reactor (but also for PWRs and BWRs) consists of a lattice field with dodecagonal meshes. These meshes are formed by three each adjacent hexagons grouped arround a central axis. The pairs of legs extending into the dodecagon and being staggered by 120 0 are designed as knubs with inclined abutting surfaces for the fuel rods. By this means there is formed a three-point bearing for centering the fuel rods. The spacer grid mentioned above is rough-worked from a single disc- resp. plate-shaped body (unfinished piece). (DG) [de

Spacer grid for fuel elements

International Nuclear Information System (INIS)

Hensolt, T.; Huenner, M.; Rau, P.; Veca, A.

1980-01-01

The spacer grid for fuel elements of a gas-cooled fast breeder reactor (but also for PWRs and BWRs) consists of a lattice field with dodecagonal meshes. These meshes are formed by three each adjacent hexagons grouped arround a central axis. The pairs of legs extending into the dodecagon and being staggered by 120 are designed as knubs with inclined abutting surfaces for the fuel rods. By this means there is formed a three-point bearing for centering the fuel rods. The spacer grid mentioned above is rough-worked from a single disc- resp. plate-shaped body (unfinished piece). (orig.)

RA-3 core with uranium silicide fuel elements

International Nuclear Information System (INIS)

Abbate, Maximo J.; Sbaffoni, Maria M.

2000-01-01

Following on with studies on uranium silicide fuel elements, this paper reports some comparisons between the use of standard ECN [U 3 O 8 ] fuel elements and type P-06 [from U 3 Si 2 ] 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)

Experimental and numerical investigation of water flow through spacer grids of nuclear fuel elements using the Open FOAM code

International Nuclear Information System (INIS)

Vidal, Guilherme A.M.; Vieira, Tiago A.S.; Castro, Higor F.P.

2017-01-01

With the advancement and development of computational tools, the studies of thermofluidodynamic behavior in nuclear fuel elements have been developed in recent years. Of the devices present in these elements, the spacing grids received more attention. They have kept the fuel rods equally spaced and have fins that aim to improve the heat transfer process between the water and the fuel element. Therefore, the grids present an important structural and thermal function. This work was carried out with the purpose of verifying and validating simulations of spacer grids using OpenFOAM (2017) software of Computational Fluid Dynamics (CFD). The simulations were validated using results obtained through the commercial CFD program, Ansys CFX, and experiments available in the literature and obtained in test sections assembled on the Water-Air Circuit (CCA) of the CDTN thermo-hydraulic laboratory

Evaluation of thermal conductivity of multi-component and multi-phase nuclear fuels by the finite element method

International Nuclear Information System (INIS)

Kurosaki, K.

2015-01-01

The effects of the shape and distribution state of voids on the thermal conductivity of UO 2 , and the temperature distribution and heat flow within the irradiated MOX fuel were evaluated by finite element analysis. Although the work is still in progress, some preliminary results are presented. (author)

Control in fabrication of PWR and BWR type reactor fuel elements

International Nuclear Information System (INIS)

Gorskij, V.V.

1981-01-01

Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

Natural uranium metallic fuel elements: fabrication and operating experience

International Nuclear Information System (INIS)

Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

1980-01-01

The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

Design of JMTR high-performance fuel element

International Nuclear Information System (INIS)

Sakurai, Fumio; Shimakawa, Satoshi; Komori, Yoshihiro; Tsuchihashi, Keiichiro; Kaminaga, Fumito

1999-01-01

For test and research reactors, the core conversion to low-enriched uranium fuel is required from the viewpoint of non-proliferation of nuclear weapon material. Improvements of core performance are also required in order to respond to recent advanced utilization needs. In order to meet both requirements, a high-performance fuel element of high uranium density with Cd wires as burnable absorbers was adopted for JMTR core conversion to low-enriched uranium fuel. From the result of examination of an adaptability of a few group constants generated by a conventional transport-theory calculation with an isotropic scattering approximation to a few group diffusion-theory core calculation for design of the JMTR high-performance fuel element, it was clear that the depletion of Cd wires was not able to be predicted accurately using group constants generated by the conventional method. Therefore, a new generation method of a few group constants in consideration of an incident neutron spectrum at Cd wire was developed. As the result, the most suitable high-performance fuel element for JMTR was designed successfully, and that allowed extension of operation duration without refueling to almost twice as long and offer of irradiation field with constant neutron flux. (author)

Fuel elements for pressurised-gas reactors; Elements combustibles des piles a gaz sous pression

Energy Technology Data Exchange (ETDEWEB)

Stohr, J A; Englander, M; Gauthron, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The design and fabrication of fuel elements for the first CO{sub 2} pressurized reactors have induced to investigate: various cladding materials, natural uranium base fuels, canning processes. The main analogical tests used in connection with the fuel element study are described. These various tests have enabled, among others, the fabrication of the fuel element for the EL2 reactor. Lastly, future solutions for electrical power producing reactors are foreseen. (author)Fren. [French] L'etude et la realisation d'elements combustibles pour les premieres piles a CO{sub 2} sous pression ont conduit a examiner: les divers materiaux de gaine, les combustibles a base d'uranium naturel, les modes de gainage. Les principaux essais analogiques ayant servi au cours de l'etude de la cartouche sont decrits. Ces divers essais ont notamment permis la realisation de la cartouche de la pile EL2. Enfin sont envisagees les solutions futures pour les piles productrices d'energie electrique. (auteur)

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E. D.

1984-01-01

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

Nuclear reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Hindle, E. D.

1984-10-16

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

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

International Nuclear Information System (INIS)

Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

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

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

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1999-01-01

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

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

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1999-01-01

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

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

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-08-01

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

Assessment of core characteristics during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

International Nuclear Information System (INIS)

Jeong, Chang Joon; Suk, Ho Chun

2002-01-01

A transition from 37-element natural uranium fuel to CANFLEX-NU fuel has been modeled in a 1200-day time-dependent fuel management simulation for a CANDU 6 reactor. The simulation was divided into three parts. The pre-transition period extended from 0 to 300 FPD, in which the reactor was fuelled only with standard 37-element fuel bundles. In the transition period, refueling took place only with the CANFLEX-NU fuel bundle. The transition stage lasted from 300 to 920 FPD, at which point all of the 37-element fuel in the core had been replaced by CANFLEX-NU fuel bundle. In the post-transition phase, refueling continued with CANFLEX-NU fuel until 1200 FPD, to arrive at estimate of the equilibrium core characteristics with CANFLEX-NU fuel. Simulation results show that the CANFLEX-NU fuel bundle has a operational compatibility with the CANDU 6 reactor during the transition core, and also show that the transition core from 37-element natural uranium fuel to CANFLEX-NU can be operated without violating any license limit of the CANDU 6 reactor