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

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

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

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

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.

Study on the performance of fuel elements with carbide and carbide-nitride fuel

International Nuclear Information System (INIS)

Golovchenko, Yu.M.; Davydov, E.F.; Maershin, A.A.

1985-01-01

Characteristics, test conditions and basic results of material testing of fuel elements with carbide and carbonitride fuel irradiated in the BOR-60 reactor up to 3-10% burn-up at specific power rate of 55-70 kW/m and temperatures of the cladding up to 720 deg C are described. Increase of cladding diameter is stated mainly to result from pressure of swelling fuel. The influence of initial efficient porosity of the fuel on cladding deformation and fuel stoichiometry on steel carbonization is considered. Utilization of carbide and carbonitride fuel at efficient porosity of 20% at the given test modes is shown to ensure their operability up to 10% burn-up

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

Design and performance of sodium-bonded uranium--plutonium carbide fuel elements

International Nuclear Information System (INIS)

Kerrisk, J.F.; DeMuth, N.S.; Petty, R.L.; Latimer, T.W.; Vitti, J.A.; Jones, L.J.

1979-01-01

Recent results from irradiation tests indicate that sodium-bonded elements provide a practical advanced fuel element design for use in LMFBRs. Shroud tubes have effectively controlled fuel-cladding mechanical interaction; thicker and stronger claddings have also been effective in this respect. Burnups to 11 at.% have been achieved under typical operating conditions. A hetrogeneous core with a breeding ratio of 1.55 and a compound system doubling time of less than 13 years has been designed using these element designs

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

The influence of fast reactor emergency conditions upon fuel element performance

International Nuclear Information System (INIS)

Bagdasarov, Yu.E.; Buksha, Yu.K.; Zabudko, L.M.; Likhachev, Yu.I.

1985-01-01

Fuel-pin cladding is one of the most important protective barriers preventing the release and propagation of radioactive contamination. By now the calculated determination of fast-reactor fuel-element performance under stationary conditions has been considered in detail but the investigation of the influence of emergency conditions has been given less attention. Under emergency conditions of the fast reactor operation there arise short-duration excesses of rated parameters (temperature, energy release, etc.) which are confined within tolerable limits with the use of the safety system. Some features of the sodium-cooled fast reactors (small mean prompt-neutron lifetime, relatively weak reactivity feedback, etc.) complicate the work of safety systems. Therefore, the tolerable deviations of parameters should be carefully validated

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.)

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.)

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

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.)

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.

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)

Fuel performance, design and development

International Nuclear Information System (INIS)

Prasad, P.N.; Tripathi, Rahul Mani; Soni, Rakesh; Ravi, M.; Vijay Kumar, S.; Dwivedi, K.P.; Pandarinathan, P.R.; Neema, L.K.

2006-01-01

The normal fuel configurations for operating 220 MWe and 540 MWe PHWRs are natural uranium dioxide 19-element and 37- element fuel bundle types respectively. The fuel configuration for BWRs is 6 x 6 fuel. So far, about 330 thousand PHWR fuel bundles and 3500 number of BWR bundles have been irradiated in the 14 PHWRs and 2 BWRs. Improvements in fuel design, fabrication, quality control and operating practices are continuously carried out towards improving fuel utilization as well as reducing fuel failure rate. Efforts have been put to improve the fuel bundle utilization by increasing the fuel discharge burnup of the natural uranium bundles The overall fuel failure rate currently is less than 0.1 % . Presently the core discharge burnups in different reactors are around 7500 MWD/TeU. The paper gives the fuel performance experience over the years in the different power reactors and actions taken to improve fuel performance over the years. (author)

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

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

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

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. 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

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

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

Preliminary Nuclear Analysis for the HANARO Fuel Element with Burnable Absorber

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

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)

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

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)

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

International Nuclear Information System (INIS)

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

2001-05-01

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

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

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

International Nuclear Information System (INIS)

Soba, Alejandro; Denis, Alicia C.

1999-01-01

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

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)

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

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.)

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)

A CAREM type fuel element dynamic analysis

International Nuclear Information System (INIS)

Magoia, J.E.

1990-01-01

A first analysis on the dynamic behaviour of a fuel element designed for the CAREM nuclear reactor (Central Argentina de Elementos Modulares) was performed. The model used to represent this dynamic behaviour was satisfactorily evaluated. Using primary estimations for some of its numerical parameters, a first approximation to its natural vibrational modes was obtained. Results obtained from fuel elements frequently used in nuclear power plants of the PWR (Pressurized Water Reactors) type, are compared with values resulting from similar analysis. (Author) [es

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.

AKORT-1 on-line system for technological control of the mixed fuel distribution in fuel elements

International Nuclear Information System (INIS)

Baklanov, V.S.; Besednov, G.Yu.; Gadzhiev, G.I.

1982-01-01

An on-line system for technological control of experimental fuel elements with vibrocompacted UO 2 -PuO 2 fuel rods fabricated for the BOR-60 reactor is described. Equipment and performance specific features of the system mechanical part and electronic circuits are considered. The results of the system performance testing are given. The fuel element quality sorting is made on the base of the analysis of Pu and fuel density distributions in the rod length. Gamma-absorption method for density measuring and the method of Pu content determination by its own gamma radiation are used in the system simultaneously. The system has the following main characteristics: tested fuel element diameter is 6 mm; the range of fuel rod mean densities is 7-10 g/cm 3 ; Pu content in the fuel is more than 20%; gamma detectors are the NaI(Tl) detectors with dimensions 40x40 and 25x25 mm; energy resolution is 137 Cs gamma line. Electronic circuits of the system operating on-line with the D3-28 microcomputer are made using the VECTOR standard. The system testing has shown that the error in the fuel density determination is less than 1%, that for Pu content measuring is 4%, the system capacity is 6 fuel elements per hour

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

Properties of U3Si2-Al dispersion fuel element and its application

International Nuclear Information System (INIS)

Yin Changgeng

2001-01-01

The properties of U 3 Si 2 fuel and U 3 Si 2 -Al dispersion fuel element are introduced, which include U-loading; the banding quality, U-homogeneity and 'dog-bone' phenomenon, the minimum thickness of cladding and the corrosion performances. The fabrication technique of fuel elements, NDT for fuel plates, assemble technique of fuel elements and the application of U 3 Si 2 -Al dispersion fuel elements in the world are introduced

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)

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)

Fuel elements and safety engineering goals

International Nuclear Information System (INIS)

Schulten, R.; Bonnenberg, H.

1990-01-01

There are good prospects for silicon carbide anti-corrosion coatings on fuel elements to be realised, which opens up the chance to reduce the safety engineering requirements to the suitable design and safe performance of the ceramic fuel element. Another possibility offered is combined-cycle operation with high efficiencies, and thus good economic prospects, as with this design concept combining gas and steam turbines, air ingress due to turbine malfunction is an incident that can be managed by the system. This development will allow economically efficient operation also of nuclear power reactors with relatively small output, and hence contribute to reducing CO 2 emissions. (orig./DG) [de

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)

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 criticality assessment of Oak Ridge research fuel element storage

International Nuclear Information System (INIS)

Thomas, J.T.

1978-06-01

Spent and partially spent Oak Ridge Research Reactor (ORR) fuel elements are retained in the storage section of the ORR pool facility. Determination of a maximum expected neutron multiplication factor for the storage area is accomplished by a validated calculational method. The KENO Monte Carlo code and the Hansen-Roach 16-group neutron cross section sets were validated by calculations of critical experiments performed with early ORR fuel elements and with SPERT-D fuel elements. Calculations of various fuel element arrangements are presented which confirm the subcriticality previously inferred from critical experiments and indicate the k/sub eff/ would not exceed 0.85, were the storage area to be filled to capacity with storage racks containing elements with the fissionable material loading increased to 350 g of 235 U

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

Sipping test on a failed MTR fuel element

International Nuclear Information System (INIS)

Terremoto, Luis Antonio Albiac; Zeituni, Carlos Alberto; Silva, Antonio Teixeira e; Perrotta, Jose Augusto; Silva, Jose Eduardo Rosa da

2002-01-01

This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes 131 I and 133 I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U 3 O 8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs. (author)

TACO: fuel pin performance analysis

International Nuclear Information System (INIS)

Stoudt, R.H.; Buchanan, D.T.; Buescher, B.J.; Losh, L.L.; Wilson, H.W.; Henningson, P.J.

1977-08-01

The thermal performance of fuel in an LWR during its operational lifetime must be described for LOCA analysis as well as for other safety analyses. The determination of stored energy in the LOCA analysis, for example, requires a conservative fuel pin thermal performance model that is capable of calculating fuel and cladding behavior, including the gap conductance between the fuel and cladding, as a function of burnup. The determination of parameters that affect the fuel and cladding performance, such as fuel densification, fission gas release, cladding dimensional changes, fuel relocation, and thermal expansion, should be accounted for in the model. Babcock and Wilcox (B and W) has submitted a topical report, BAW-10087P, December 1975, which describes their thermal performance model TACO. A summary of the elements that comprise the TACO model and an evaluation are presented

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.)

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

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

International Nuclear Information System (INIS)

Warinner, D.K.

1980-01-01

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

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

Fuel Retrieval Sub-Project (FRS) Stuck Fuel Station Performance Test Data Report

International Nuclear Information System (INIS)

THIELGES, J.R.

2000-01-01

This document provides the test data report for Stuck Fuel Station Performance Testing in support of the Fuel Retrieval Sub-Project. The stuck fuel station was designed to provide a means of cutting open a canister barrel to release fuel elements, etc

Atomic pile Directorate, Department of Metallurgy, Departments of Technology, Department of Fuel Elements and Structures, Division of Study of Fuel Elements - Semi annual report on the 1968-10-1

International Nuclear Information System (INIS)

Arnaud, M.; Tortel, J.; Viallet, H.; Marinot, R.; Rulleau, A.; Lestiboudois, G.; Rousseau, G.; Faussat, A.; Ollier, H.; Truffert, J.; Ferrier, C.; Courcon, P.; Rendu, M.; Dieumegard, M.; Bret, A.

1968-01-01

This document gathers a set of reports of studies performed on nuclear fuel elements. The addressed topics are: creep behaviour of UMo and UMoAl tubes and pellets under the action of an external pressure (creep strength of tubes under external pressure, creep strength of pellets under external pressure, uncertainties on irradiation parameters in Pegase), problems related to centring devices (measurements and tests), irradiations of ring elements in power reactors, uranium/sheath metallurgical relationship for Bugey and influence of irradiation (cartridge behaviour in Pegase, long duration irradiation in power reactors, extrapolation in Bugey of results obtained in G2), theoretical study of kinetic oxidation phenomena in metal fuels, tests of leaking cartridges in EdF2, evolution of pressure in EL4 type irradiated fuel rods with ZrCu liners with respect to the conductivity integral, a focus on irradiations of Z0 type fuel elements in Pegase, cluster safety tests with uranium carbide in pile and out of pile, a review of studies performed on fuel elements with blowhole, and application of neutrography to fuel elements

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.)

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

International Nuclear Information System (INIS)

Hofman, G.L.; Rest, J.; Snelgrove, J.L.

1995-01-01

Calculations of fuel swelling of U 3 SiAl-Al and U 3 Si 2 were performed for various dispersion fuel element designs. Breakaway swelling criteria in the form of critical fuel volume fractions were derived with data obtained from U 3 SiAl-Al plate irradiations. The results of the analysis show that rod-type elements remain well below the pillowing threshold. However, tubular fuel elements, which behave essentially like plates, will likely develop pillows or blisters at around 90% 235 U burnup. The U 3 Si 2 -Al compounds demonstrate stable swelling behavior throughout the entire burnup range for all fuel element designs

Recent status of development and irradiation performance for plate type fuel elements with reduced 235U enrichment at NUKEM

International Nuclear Information System (INIS)

Hrovat, M.F.; Hassel, H.W.

1984-01-01

According to the present state of development full size test fuel elements with the maximum uranium densities of 2,2 g U/cm 3 meat for UAlsub(x), 3,2 g U/cm 3 meat for U 3 O 8 and 4,8 g U/cm 3 meat for U 3 Si 2 can be fabricated at NUKEM in production scale. Special chemical procedures for the uranium recovery were developed ensuring an economic fuel fabrication process. The post irradiation examinations (PIE) of 12 UAlsub(x) (U density 2,2 g U/cm 3 meat) and U 3 O 8 (up to 3,1 g U/cm 3 meat) test plates irradiated in the ORR, Oak Ridge research reactor, were terminated. All 12 test plates show unobjectionable irradiation behavior. Extensive irradiation tests on full size fuel elements were performed. All inserted elements show perfect irradiation behavior. The PIE of the first HFR Petten U 3 O 8 fuel elements are in progress. The full size ORR U 3 Si 2 fuel elements with so far highest uranium density of 4,76 g U/cm 3 meat achieved a burnup of 50 % loss of 235 U up to May 1983. One element was withdrawn from the reactor for PIE, the second will be irradiated to a burnup of 75 % loss of 235 U. The further development is concentrated on Usub(x)Sisub(y) fuel with highest uranium density. U 3 Si miniplates with up to 6,1 g U/cm 3 meat are supplied meeting the required specification, U 3 Si miniplates with 6,7 g U/cm 3 are in fabrication. (author)

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

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

PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements

International Nuclear Information System (INIS)

Ganzmann, I.; Hille, D.; Staude, U.

2009-01-01

The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

International Nuclear Information System (INIS)

Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

1999-01-01

The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses

Fuel element for a nuclear reactor

International Nuclear Information System (INIS)

Tanihiro, Yasunori; Sumita, Isao.

1970-01-01

An improved fuel element of the heat pipe type is disclosed in which the fuel element itself is given a heat pipe structure and filled with a coated particle fuel at the section thereof having a capillary tube construction, whereby the particular advantages of heat pipes and coated fuels are combined and utilized to enhance thermal control and reactor efficiency. In an embodiment, the fuel element of the present invention is filled at its lower capillary tube section with coated fuel and at its upper section with a granurated neutron absorber. Both sections are partitioned from the central shaft by a cylindrically shaped wire mesh defining a channel through which the working liquid is vaporized from below and condensed by the coolant external to the fuel element. If the wire mesh is chosen to have a melting point lower than that of the fuel but higher than that of the operating temperature of the heat pipe, the mesh will melt and release the neutron absorbing particles should hot spots develop, thus terminating fission. (Owens, K. J.)

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.)

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.

Weld Joint Design for SFR Metallic Fuel Element Closures

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Kim, Ki Hwan; Yoon, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The sodium-cooled fast reactor (SFR) system is among the six systems selected for Gen-IV promising systems and expected to become available for commercial introduction around 2030. In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the joint designs for endplug welding were investigated. For the irradiation test of SFR metallic fuel element, the TIG welding technique was adopted and the welding joint design was developed based on the welding conditions and parameters established. In order to make SFR metallic fuel elements, the weld joint design was developed based on the TIG welding technique.

Endplug Welding Techniques developed for SFR Metallic Fuel Elements

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established.

Endplug Welding Techniques developed for SFR Metallic Fuel Elements

International Nuclear Information System (INIS)

Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan

2013-01-01

In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established

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)

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.)

Pellet-clad interaction observations in boiling water reactor fuel elements

International Nuclear Information System (INIS)

Sahoo, K.C.; Bahl, J.K.; Sivaramakrishnan, K.S.; Roy, P.R.

1981-01-01

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

Remote real time x-ray examination of fuel elements in a hot cell environment

International Nuclear Information System (INIS)

Yapuncich, F.L.

1993-01-01

This report discusses the Remote Real Time X-ray System which will allow for detailed examination of fuel elements. This task will be accomplished in a highly radioactive hot cell environment. Two remote handling systems win be utilized at the examination station. One handling system will transfer the fuel element to and from the shielded x-ray system. A second handling system will allow for vertical and rotational inspection of the fuel elements. The process win include removing a single nuclear fuel element from a element fabrication magazine(EFM), positioning the fuel element within the shielding envelope of the x-ray system and transferring the fuel element from the station manipulator to the x-ray system manipulator, performing the x-ray inspection, and then transferring the fuel element to either the element storage magazine(ESM) or a reject bin

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

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.

Study of candu fuel elements irradiated in a nuclear power plant

International Nuclear Information System (INIS)

Ionescu, S.; Uta, O.; Mincu, M.; Anghel, D.; Prisecaru, I.

2015-01-01

The object of this work is the behaviour of CANDU fuel elements after service in nuclear power plant. The results are analysed and compared with previous result obtained on unirradiated samples and with the results obtained on samples irradiated in the TRIGA reactor of INR Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor, the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) of INR Pitesti on samples from fuel elements after they were removed out of the nuclear power plant: - dimensional and macrostructural characterization; - microstructural characterization by metallographic analyses; - determination of mechanical properties; - fracture surface analysis by scanning electron microscopy (SEM). A full set of non-destructive and destructive examinations concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the cladding was performed. The obtained results are typical for CANDU 6-type fuel. The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for the improvement of the CANDU fuel. (authors)

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

Science.gov (United States)

Stewart, Mark E.; Schnitzler, Bruce G.

2015-01-01

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

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

In-reactor cladding breach of EBR-II driver-fuel elements

International Nuclear Information System (INIS)

Seidel, B.R.; Einziger, R.E.

1977-01-01

Knowledge of performance and minimum useful element lifetime of Mark-II driver-fuel elements is required to maintain a high plant operating capacity factor with maximum fuel utilization. To obtain such knowledge, intentional cladding breach has been obtained in four run-to-cladding-breach Mark-II experimental driver-fuel subassemblies operating under normal conditions in EBR-II. Breach and subsequent fission-product release proved benign to reactor operations. The breaches originated on the outer surface of the cladding in the root of the restrainer dimples and were intergranular. The Weibull distribution of lifetime accurately predicts the observed minimum useful element lifetime of 10 at.% burnup, with breach ensuing shortly thereafter

Nondestructive examination techniques on Candu fuel elements

International Nuclear Information System (INIS)

Gheorghe, G.; Man, I.

2013-01-01

During irradiation in nuclear reactor, fuel elements undergo dimensional and structural changes, and changes of surface conditions sheath as well, which can lead to damages and even loss of integrity. Visual examination and photography of Candu fuel elements are among the non-destructive examination techniques, next to dimensional measurements that include profiling (diameter, bending, camber) and length, sheath integrity control with eddy currents, measurement of the oxide layer thickness by eddy current techniques. Unirradiated Zircaloy-4 tubes were used for calibration purposes, whereas irradiated Zircaloy-4 tubes were actually subjected to visual inspection and dimensional measurements. We present results of measurements done by eddy current techniques on Zircaloy- 4 tubes, unirradiated, but oxidized in an autoclave prior to examinations. The purpose of these nondestructive examination techniques is to determine those parameters that characterize the behavior and performance of nuclear fuel operation. (authors)

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

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

Status and aspects of fuel element development for advanced high-temperature reactors in the FRG

International Nuclear Information System (INIS)

Nickel, H.; Balthesen, E.

1975-01-01

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

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

Study on thermal performance and margins of BWR fuel elements

International Nuclear Information System (INIS)

Stosic, Zoran

1999-01-01

This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

Analysis of neutron flux depression across the pellet radius in CANDU fuel elements

International Nuclear Information System (INIS)

Sim, K.S.; Suk, H.C.

1998-08-01

The TUBRNP model, originally developed to perform the analysis of the flux depression across the pellet radius in LWR fuel elements, was improved for the application to CANDU fuel elements. The improved model was verified through comparison with existing CANDU model named FLUXDEP in prediction for various fuel conditions. A sensitivity study was also performed to investigate the effects on the flux depression of fuel initial enrichment and burnup, the contents of isotopes U-234 and U-236 and pellet diameter. (author). 9 refs., 8 figs

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-19

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

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

International Nuclear Information System (INIS)

Montierth, Leland M.

2016-01-01

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

Failure analysis for WWER-fuel elements

International Nuclear Information System (INIS)

Boehmert, J.; Huettig, W.

1986-10-01

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

High performance nuclear fuel element

International Nuclear Information System (INIS)

Mordarski, W.J.; Zegler, S.T.

1980-01-01

A fuel-pellet composition is disclosed for use in fast breeder reactors. Uranium carbide particles are mixed with a powder of uraniumplutonium carbides having a stable microstructure. The resulting mixture is formed into fuel pellets. The pellets thus produced exhibit a relatively low propensity to swell while maintaining a high density

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

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

International Nuclear Information System (INIS)

Agueda, Horacio C.; Estevez, Esteban; Gerding, Jose R.; Markiewicz, Mario E.

2003-01-01

Some features of the mechanical design of the low enrichment fuel element for the RA-3 reactor are described, with emphasis in those aspects of the original design that have been modified considering the experience acquired in the design of other fuel elements. The proposed modification is based fundamentally on the replacement of all welded joints by screwed joints, which facilitates the manufacture of the fuel element, avoiding the distortions produced by the welds used at present and contributing to the fulfillment of the foreseen tolerances. A basic characteristic of this design is a careful manufacture of the fuel element's structural components in order to assure an assembling of the fuel element that fulfills the tolerances intrinsically required. The fuel is designed for the RA-3 reactor and uses U 3 O 8 or U 3 Si 2 as carrying phase of the fissile material with an enrichment of 19.70% of 235 U. The design verification was performed by analytical and numerical methods, and is supported by testing of materials in laboratory, hydrodynamics tests and performance evaluations of the fuel elements in the RA-3 reactor. (author)

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)

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

International Nuclear Information System (INIS)

Sudjatmi KA

2013-01-01

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

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)

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

International Nuclear Information System (INIS)

Brasnarof D; Marino, A. C; Florido, P. C; Daverio, H

2006-01-01

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

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

Current status of the transient integral fuel element performance code URANUS

International Nuclear Information System (INIS)

Preusser, T.; Lassmann, K.

1983-01-01

To investigate the behavior of fuel pins during normal and off-normal operation, the integral fuel rod code URANUS has been extended to include a transient version. The paper describes the current status of the program system including a presentation of newly developed models for hypothetical accident investigation. The main objective of current development work is to improve the modelling of fuel and clad material behavior during fast transients. URANUS allows detailed analysis of experiments until the onset of strong material transport phenomena. Transient fission gas analysis is carried out due to the coupling with a special version of the LANGZEIT-KURZZEIT-code (KfK). Fuel restructuring and grain growth kinetics models have been improved recently to better characterize pre-experimental steady-state operation; transient models are under development. Extensive verification of the new version has been carried out by comparison with analytical solutions, experimental evidence, and code-to-code evaluation studies. URANUS, with all these improvements, has been successfully applied to difficult fast breeder fuel rod analysis including TOP, LOF, TUCOP, local coolant blockage and specific carbide fuel experiments. Objective of further studies is the description of transient PCMI. It is expected that the results of these developments will contribute significantly to the understanding of fuel element structural behavior during severe transients. (orig.)

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

Fuel element burnup determination in HEU-LEU mixed TRIGA research reactor core

International Nuclear Information System (INIS)

Zagar, Tomaz; Ravnik, Matjaz

2000-01-01

This paper presents the results of a burnup calculations and burnup measurements for TRIGA FLIP HEU fuel elements and standard TRIGA LEU fuel elements used simultaneously in small TRIGA Mark II research reactor in Ljubljana, Slovenija. The fuel element burnup for approximately 15 years of operation was calculated with two different in house computer codes TRIGAP and TRIGLAV (both codes are available at OECD NEA Data Bank). The calculation is performed in one-dimensional radial geometry in TRIGAP and in two-dimensional (r,φ) geometry in TRIGLAV. Inter-comparison of results shows important influence of in-core water gaps, irradiation channels and mixed rings on burnup calculation accuracy. Burnup of 5 HEU and 27 LEU fuel elements was also measured with reactivity method. Measured and calculated burnup values are inter-compared for these elements (author)

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.

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.)

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.)

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)

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

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

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

Science.gov (United States)

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

2013-01-01

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

Optimization of FBR fuel element for high burnup

International Nuclear Information System (INIS)

Marbach, G.; Millet, P.

1985-03-01

After a brief historical background showing evolution of the French fast reactor fuel element from RAPSODIE to PHENIX and SUPER PHENIX we have examined the main points which have permitted to increase irradiation performance of the subassembly

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

Fuel performance analysis code 'FAIR'

International Nuclear Information System (INIS)

Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1994-01-01

For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs

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)

Automatic inspection for remotely manufactured fuel elements

International Nuclear Information System (INIS)

Reifman, J.; Vitela, J.E.; Gibbs, K.S.; Benedict, R.W.

1995-01-01

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

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

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

International Nuclear Information System (INIS)

1973-01-01

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

Performance of candu-6 fuel bundles manufactured in romania nuclear fuel plant

International Nuclear Information System (INIS)

Bailescu, A.; Barbu, A.; Din, F.; Dinuta, G.; Dumitru, I.; Musetoiu, A.; Serban, G.; Tomescu, A.

2013-01-01

The purpose of this article is to present the performance of nuclear fuel produced by Nuclear Fuel Plant (N.F.P.) - Pitesti during 1995 - 2012 and irradiated in units U1 and U2 from Nuclear Power Plant (N.P.P.) Cernavoda and also present the Nuclear Fuel Plant (N.F.P.) - Pitesti concern for providing technology to prevent the failure causes of fuel bundles in the reactor. This article presents Nuclear Fuel Plant (N.F.P.) - Pitesti experience on tracking performance of nuclear fuel in reactor and strategy investigation of fuel bundles notified as suspicious and / or defectives both as fuel element and fuel bundle, it analyzes the possible defects that can occur at fuel bundle or fuel element and can lead to their failure in the reactor. Implementation of modern technologies has enabled optimization of manufacturing processes and hence better quality stability of achieving components (end caps, chamfered sheath), better verification of end cap - sheath welding. These technologies were qualified by Nuclear Fuel Plant (N.F.P.) - Pitesti on automatic and Computer Numerical Control (C.N.C.) programming machines. A post-irradiation conclusive analysis which will take place later this year (2013) in Institute for Nuclear Research Pitesti (the action was initiated earlier this year by bringing a fuel bundle which has been reported defective by pool visual inspection) will provide additional information concerning potential damage causes of fuel bundles due to manufacturing processes. (authors)

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.)

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.)

Vibration test and endurance test for HANARO 36-element fuel assembly

International Nuclear Information System (INIS)

Ryu, Jeong Soo; Kim, Heon ll; Chung, Heung June

1998-06-01

Vibration test and endurance test for HANARO DU (depleted uranium) 36-element fuel assembly which was fabricated by KAERI were carried out based on the HANARO operation conditions. The endurance test of 22 days was added to the previous 18 days test. The vibration test was performed at various flow rates. Vibration frequency for the 36-element fuel assembly is between 11 to 14.5 Hz. And the maximum vibration displacement is less than 100 μm. From the endurance test result, it can be concluded that the appreciable fretting wear for the 36-element fuel assembly and the hexagonal flow tube was not observed. (author). 4 refs., 5 tabs., 29 figs

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)

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

3D finite element analysis of a nuclear fuel rod with gap elements between the pellet and the cladding

International Nuclear Information System (INIS)

Kang, Chang-Hak; Lee, Sung-Uk; Yang, Dong-Yol; Kim, Hyo-Chan; Yang, Yong-Sik

2016-01-01

Nuclear fuel rods which comprises an important component of a nuclear power plant are composed of nuclear fuel and cladding. Simulating the nuclear fuel rod using a computer program is the universal method to verify its safety. The computer program used for this is called the fuel performance code. The main objective of this study is to simulate the nuclear fuel rod behavior considering the gap conductance using three-dimensional gap elements. Gap elements are used because, unlike other methods, this approach does not require special methods or other variables such as the Lagrange multiplier. In this work, a nuclear fuel rod has been simulated and the results are compared with the experimental results. (author)

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

SGHWR fuel performance, safety and reliability

International Nuclear Information System (INIS)

Pickman, D.O.; Inglis, G.H.

1977-01-01

The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of performance, safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith S.G.H.W.R. that the necessary bias toward safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications and has to be paid for. By a detailed evaluation of S.G.H.W.R. fuel defects it is shown that very few defects can be shown to be related to design, rating or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that must be met to achieve the desirable reliability level. It is possible that large-scale experience with S.G.H.W.R. fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to S.G.H.W.R. fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. The philosophy and supporting experimental work programme are outlined and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to new 60 pin fuel element to be used in the commercial S.G.H.W.R.'s and how it compares in design and performance aspects with the 36 pin element that has been used to date in the Winfrith S.G.H.W.R

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

International Nuclear Information System (INIS)

1983-03-01

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

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)

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

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.

1994-10-01

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

DUPIC fuel performance from reactor physics viewpoint

International Nuclear Information System (INIS)

Choi, H.; Rhee, B.W.; Park, H.

1995-01-01

A preliminary study was performed for the evaluation of Stress Corrosion Cracking (SCC) parameters of nominal DUPIC fuel in CANDU reactor. For the reference 2-bundle shift refueling scheme, the predicted ramped power and power increase of the 43-element DUPIC fuel in the equilibrium core are below the SCC thresholds of CANDU natural uranium fuel. For 4-bundle shift refueling scheme, the envelope of element ramped power and power increase upon refueling are 8% and 44% higher than those of 2-bundle shift refueling scheme on the average, respectively, and both schemes are not expected to cause SCC failures. (author)

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

International Nuclear Information System (INIS)

Kulichevsky, R.; Martin Ghiselli, A.; Fiori, J.; Yedros, P.

2002-01-01

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

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

International Nuclear Information System (INIS)

Lassmann, K.

1991-01-01

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

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)

Irradiation effects on mechanical properties of fuel element cladding from thermal reactors

International Nuclear Information System (INIS)

Chatterjee, S.

2005-01-01

During reactor operation, UO 2 expands more than the cladding tube (Zirconium alloys for thermal reactors), is hotter, cracks and swells. The fuel therefore will interact with the cladding, resulting in straining of the later. To minimize the possibility of rupture of the cladding, ideally it should have good ductility as well as high strength. However, the ductility reduces with increase in fuel element burn-up. Increased burn-up also increases swelling of the fuel, leading to increased contact pressure between the fuel and the cladding tube. This would cause strains to be concentrated over localized regions of the cladding. For fuel elements burnup exceeding 40 GWd/T, the contribution of embrittlement due to hydriding, and the increased possibility of embrittlement due to stress corrosion cracking, also need to be considered. In addition to the tensile properties, the other mechanical properties of interest to the performance of cladding tube in an operating fuel element are creep rate and fatigue endurance. Irradiation is reported to have insignificant effect on high cycle endurance limit, and fatigue from fuel element vibration is most unlikely, to be life limiting. Even though creep rates due to irradiation are reported to increase by an order of magnitude, the cladding creep ductility would be so high that creep type failures in fuel element would be most improbable. Thus, the most important limiting aspect of mechanical performance of fuel element cladding has been recognized as the tensile ductility resulting from the stress conditions experienced by the cladding. Some specific fission products of threshold amount (if) deposited on the cladding, and hydride morphology (e.g. hydride lenses). The presentation will brief about irradiation damage in cladding materials and its significance, background of search for better Zirconium alloys as cladding materials, and elaborate on the types of mechanical tests need to be conducted for the evaluation of claddings

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.)

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)

Transient performance of EBR-II driver fuel

International Nuclear Information System (INIS)

Buzzell, J.A.; Hudman, G.D.; Porter, D.L.

1981-01-01

The first phases of qualification of the EBR-II driver fuel for repeated transient overpower operation have recently been completed. The accomplishments include prediction of the transient fuel and cladding performance through ex-core testing and fuel-element modeling studies, localized in-core power testing during steady-state operation, and whole-core multiple transient testing. The metallic driver fuel successfully survived 56 transients, spaced over a 45-day period, with power increases of approx. 160% at rates of approx. 1%/s with a 720-second hold at full power. The performance results obtained from both ex-core and n-core tests indicate that the fuel is capable of repeated transient operation

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.)

Tests on CANDU fuel elements sheath samples

International Nuclear Information System (INIS)

Ionescu, S.; Uta, O.; Mincu, M.; Prisecaru, I.

2016-01-01

This work is a study of the behavior of CANDU fuel elements after irradiation. The tests are made on ring samples taken from fuel cladding in INR Pitesti. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory. By metallographic and ceramographic examination we determinate that the hydride precipitates are orientated parallel to the cladding surface. A content of hydrogen of about 120 ppm was estimated. After the preliminary tests, ring samples were cut from the fuel rod, and were subject of tensile test on an INSTRON 5569 model machine in order to evaluate the changes of their mechanical properties as consequence of irradiation. Scanning electron microscopy was performed on a microscope model TESCAN MIRA II LMU CS with Schottky FE emitter and variable pressure. The analysis shows that the central zone has deeper dimples, whereas on the outer zone, the dimples are tilted and smaller. (authors)

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)

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

International Nuclear Information System (INIS)

Bahl, J.K.; Sah, D.N.; Chatterjee, S.; Sivaramkrishnan, K.S.

1979-01-01

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

Distribution of fission products in Peach Bottom HTGR fuel element E01-01

International Nuclear Information System (INIS)

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

1978-10-01

The fifth in a projected series of six postirradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements is described. The element analyzed received an equivalent of 897 full-power days of irradiation prior to the scheduled termination of Core 2 operation. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a 137 Cs inventory of 20.3 Ci in the graphite sleeve and 8.1 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides 134 Cs, /sup 110 m/Ag, 60 Co, and 154 Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the beta-emitters 3 H, 14 C, and 90 Sr were obtained at four axial locations of the fueled region of the element sleeve and two axial locations of the element spine. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. In addition to fission product distributions, the appearance of the component parts of the element was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed

Crossflow characteristics of flange type fuel element for very high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Takizuka, Takakazu; Kaburaki, Hideo; Suzuki, Kunihiko; Nakamura, Masahide.

1987-01-01

Fuel element design incorporating mating flanges at block end faces has the potential to improve thermal hydraulic performance of a VHTR (very high temperature gas-cooled reactor) core. As part of research and development efforts to establish flange type fuel element design, experiments and analyses were carried out on crossflow through interface gap between elements. Air at atmospheric pressure and ambient temperature was used as a fluid. Crossflow loss coefficient factors were obtained with three test models, having different flange mating clearances, for various interface gap configurations, gap widths and block misalignments. It was found that crossflow loss coefficient factors for flange type fuel element were much larger than those for conventional flat-faced element. Numerical analyses were also made using a simple model devised to represent the crossflow path at the fuel element interface. The close agreement between numerical results and experimental data indicated that this model could predict well the crossflow characteristics of the flange type fuel element. (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

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

International Nuclear Information System (INIS)

Vatulin, A. V.; Stetskiy, Y.A.; Mishunin, V.A.; Suprun, V.B.; Dobrikova, I.V.

2002-01-01

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

Experimental study of some mounting brackets to support fuel elements

International Nuclear Information System (INIS)

Aubert, M.; Poglia, S.; Roche, R.

1958-09-01

In an atomic pile with vertical channels, fuel elements are stacked on one another. According to a possible assembly, fuel element can be contained by a graphite sleeve and be supported by a mounting bracket in this sleeve. Sleeves are then stacked on one another. The authors report the investigation of different designs for these mounting brackets. They describe their mechanical role and their mechanical, aerodynamic, neutronic and test conditions. They report tests performed on brackets made in graphite and on brackets made in stainless steel and graphite, and discuss the obtained results

Examination of U3Si2-Al fuel elements from the Oak Ridge Research Reactor

International Nuclear Information System (INIS)

Copeland, G.L.; Snelgrove, J.L.; Hofman, G.L.

1986-01-01

The results of postirradiation examination of low-enriched U 3 Si 2 fuel elements from the Oak Ridge Research Reactor are presented. The elements replaced standard high-enriched elements and were handled routinely except that the burnup of half the elements was extended beyond normal limits up to about 98% peak. The elements were manufactured by commercial fuel suppliers. The performance was completely satisfactory for all the elements

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

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

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

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

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

Pyrochemical head-end treatment for fast reactor fuel elements

International Nuclear Information System (INIS)

Avogadro, A.

1978-01-01

The paper presents the R and D work performed at Ispra and Mol during the period 1965-1975 in order to find a way to overcome technical and economical difficulties arising when the conventional reprocessing is applied to fast reactor fuel elements. The work had been directed towards 3 specific topics: a) liquid-metal decladding of spent stainless steel - clad fuels (solinox process). b) oxidative pulverisation by fused salts and extraction of volatile fission products (satex process). c) Pyrochemical separation of plutonium from the bulk of the fuel

Finite element modelling of different CANDU fuel bundle types in various refuelling conditions

International Nuclear Information System (INIS)

Roman, M. R.; Ionescu, D. V.; Olteanu, G.; Florea, S.; Radut, A. C.

2016-01-01

The objective of this paper is to develop a finite element model for static strength analysis of the CANDU standard with 37 elements fuel bundle and the SEU43 with 43 elements fuel bundle design for various refuelling conditions. The computer code, ANSYS7.1, is used to simulate the axial compression in CANDU type fuel bundles subject to hydraulic drag loads, deflection of fuel elements, stresses and displacements in the end plates. Two possible situations for the fuelling machine side stops are considered in our analyses, as follows: the last fuel bundle is supported by the two side stops and a side stop can be blocked therefore, the last fuel bundle is supported by only one side stop. The results of the analyses performed are briefly presented and also illustrated in a graphical form. The finite element model developed in present study is verified against test results for endplate displacement and element bowing obtained from strength tests with fuel bundle string and fuelling machine side-stop simulators. Comparison of ANSYS model predictions with these experimental results led to a very good agreement. Despite the difference in hydraulic load between SEU43 and CANDU standard fuel bundles strings, the maximum stress in the SEU43 endplate is about the same with the maximum stress in the CANDU standard endplate. The comparative assessment reveals that SEU43 fuel bundle is able to withstand high flow rate without showing a significant geometric instability. (authors)

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

International Nuclear Information System (INIS)

Lassmann, K.

1979-01-01

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

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

International Nuclear Information System (INIS)

Kim, S. S.; Kim, W. K.; Park, C. H.; Ko, J. H.; Lee, J. W.; Yang, M. S.

2001-01-01

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

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)

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.)

SGHWR fuel performance, safety and reliability

International Nuclear Information System (INIS)

Pickman, D.O.; Inglis, G.H.

1977-05-01

The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith SGHWR that the necessary bias towards safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications. By a detailed evaluation of SGHWR fuel defects it is shown that very few defects can be shown to be related to design, rating, or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that most be met to achieve the desirable reliability level. It is possible that large scale experience on SGHWR fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to SGHWR fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. the philosophy and supporting experimental work programme are outlines and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to the new 60-pin fuel element to be used in the commercial SGHWRs and to its comparison in design and performance aspects with the 36-pin element that has been used to date in the Winfrith SGHWR. (author)

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

Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

International Nuclear Information System (INIS)

Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

2005-10-01

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor

Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

2005-10-15

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

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

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.)

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)

Effect of power variations across a fuel bundle and within a fuel element on fuel centerline temperature in PHWR bundles in uncrept and crept pressure tubes

International Nuclear Information System (INIS)

Onder, E.N.; Roubtsov, D.; Rao, Y.F.; Wilhelm, B.

2017-01-01

Highlights: • Pressure tube creep effect on fuel pin power and temperatures was investigated. • Noticeable effects were observed for 5.1% crept pressure tube. • Bundle eccentricity effect on power variations was insignificant for uncrept channels. • Difference of 112 °C was observed between top & bottom elements in 5.1% crept channel. • Not discernible fission gas release was expected with temperature difference of 112 °C. - Abstract: The neutron flux and fission power profiles through a fuel bundle and across a fuel element are important aspects of nuclear fuel analysis in multi-scale/multi-physics modelling of Pressurized Heavy Water Reactors (PHWRs) with advanced fuel bundles. Fuel channels in many existing PHWRs are horizontal. With ageing, pressure tubes creep and fuel bundles in these pressure tubes are eccentrically located, which results in an asymmetric coolant flow distribution between the top and bottom of the fuel bundles. The diametral change of the pressure tube due to creep is not constant along the fuel channel; it reaches a maximum in the vicinity of the maximum neutron flux location. The cross-sectional asymmetric positioning of fuel bundles in a crept pressure tube contributes to an asymmetric power distribution within a ring of fuel elements. Modern reactor physics lattice codes (such as WIMS-AECL) are capable of predicting the details of power distribution from basic principles. Thermalhydraulics subchannel codes (such as ASSERT-PV) use models to describe inhomogeneous power distribution within and across fuel elements (e.g., flux tilt model, different powers in different ring elements, or radial power profiles). In this work, physics and thermalhydraulics codes are applied to quantify the effect of eccentricity of a fuel bundle on power variations across it and within a fuel element, and ultimately on the fuel temperature distribution and fuel centerline temperature, which is one of the indicators of fuel performance under normal

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

Research on Elemental Technology of Advanced Nuclear Fuel Performance Verification

International Nuclear Information System (INIS)

Kim, Yong Soo; Lee, Dong Uk; Jean, Sang Hwan; Koo, Min

2003-04-01

Most of current properties models and fuel performance models used in the performance evaluation codes are based on the in-pile data up to 33,000 MWd/MtU. Therefore, international experts are investigating the properties changes and developing advanced prediction models for high burn-up application. Current research is to develop high burn-up fission gas release model for the code and to support the code development activities by collecting data and models, reviewing/assessing the data and models together, and benchmarking the selected models against the appropriate in-pile data. For high burn-up applications, two stage two step fission gas release model is developed based on the real two diffusion process in the grain lattice and grain boundaries of the fission gases and the observation of accelerated release rate in the high burn-up. It is found that the prediction of this model is in excellent agreement with the in-pile measurement results, not only in the low burn-up but also in the high burn-up. This research is found that the importance of thermal conductivity of oxide fuel, especially in the high burn-up, is focused again. It is found that even the temperature dependent models differ from one to another and most of them overestimate the conductivity in the high burn-up. An in-pile data benchmarking of high LHGR fuel rod shows that the difference can reach 30%∼40%, which predicts 400 .deg. C lower than the real fuel centerline temperature. Recent models on the thermal expansion and heat capacity of oxide fuel are found to be well-defined. Irradiation swelling of the oxide fuel are now well-understood that in most cases in LWRs solid fission product swelling is dominant. Thus, the accumulation of in-pile data can enhance the accuracy of the model prediction, rather than theoretical modeling works. Thermo-physical properties of Zircaloy cladding are also well-defined and well-understood except the thermal expansion. However, it turns out that even the

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

International Nuclear Information System (INIS)

Ravnik, M.

1988-11-01

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

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

Finite element analysis of advanced neutron source fuel plates

International Nuclear Information System (INIS)

Luttrell, C.R.

1995-08-01

The proposed design for the Advanced Neutron Source reactor core consists of closely spaced involute fuel plates. Coolant flows between the plates at high velocities. It is vital that adjacent plates do not come in contact and that the coolant channels between the plates remain open. Several scenarios that could result in problems with the fuel plates are studied. Finite element analyses are performed on fuel plates under pressure from the coolant flowing between the plates at a high velocity, under pressure because of a partial flow blockage in one of the channels, and with different temperature profiles

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-12-15

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

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

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

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

ETR fuel element shipping container addendum to PR-T-79-011 (TR-466). Internal technical report

International Nuclear Information System (INIS)

Smith, M.C.

1979-01-01

In July, 1979, EG and G Idaho, Inc. was requested to evaluate the ETR Fuel Element Shipping Container for compliance with existing transport regulations, in order to ship GETR fuel elements from Vallecitos, California to the INEL. Technical report PR-T-79-011 (TR-466), ATR Fuel Element Shipping Container Safety Analysis, was used as a basis for this evaluation. The safety analysis contained in technical report PR-T-79-011 (TR-466) was performed utilizing the ATR, ETR, MTR, and SPERT shipping containers. The report determined the ETR Fuel Element Shipping Container does comply with the existing transport regulations for a Type A quantity, Fissile Class I shipping container. The ETR and GETR fuel elements are essentially identical in physical size, construction, and fissile material content, the analysis documented in this report has determined the shipment of GETR fuel elements in the ETR shipping container to be safe and pose no threat to the public health and safety

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.

Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

Energy Technology Data Exchange (ETDEWEB)

Wang Qiming; Yan Xiaoqing [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Ding Shurong, E-mail: dsr1971@163.co [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)

2010-04-01

The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

Fuel performance of licensed nuclear power plants through 1974

International Nuclear Information System (INIS)

Bobe, P.E.

1976-01-01

General aspects of fuel element design and specific design data for typical Pressurized and Boiling Water Reactors are presented. Based on a literature search, failure modes and specific failures incurred through December 31, 1974 are described, together with a discussion of those problems which have had a significant impact upon plant operation. The relationship between fuel element failures and the resultant coolant activity/radioactive gaseous effluents upon radiation exposure, plant availability and capacity factors, economic impact, and waste management, are discussed. An assessment was made regarding the generation, availability, and use of fuel performance data

Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

Science.gov (United States)

Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

2009-08-01

Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

One-fifth-scale and full-scale fuel element rocking tests

International Nuclear Information System (INIS)

Nau, P.V.; Olsen, B.E.

1978-06-01

Using 1 / 5 -scale and 1 / 1 -scale (prototype H451) fuel elements, one, two, or three stacked elements on a clamped base element were rocked from an initial release position. Relative displacement, rock-down loads, and dowel pin shear forces were measured. A scaled comparison between 1 / 5 -scale and 1 / 1 -scale results was made to evaluate the model scaling laws, and an error analysis was performed to assess the accuracy and usefulness of the test data

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)

FEMAXI-III, a computer code for fuel rod performance analysis

International Nuclear Information System (INIS)

Ito, K.; Iwano, Y.; Ichikawa, M.; Okubo, T.

1983-01-01

This paper presents a method of fuel rod thermal-mechanical performance analysis used in the FEMAXI-III code. The code incorporates the models describing thermal-mechanical processes such as pellet-cladding thermal expansion, pellet irradiation swelling, densification, relocation and fission gas release as they affect pellet-cladding gap thermal conductance. The code performs the thermal behavior analysis of a full-length fuel rod within the framework of one-dimensional multi-zone modeling. The mechanical effects including ridge deformation is rigorously analyzed by applying the axisymmetric finite element method. The finite element geometrical model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The 8-node quadratic isoparametric ring elements are adopted for obtaining accurate finite element solutions. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behaviors accurately and stably. The pellet-cladding interaction mechanism is exactly treated using the nodal continuity conditions. The code is applicable to the thermal-mechanical analysis of water reactor fuel rods experiencing variable power histories. (orig.)

Irradiated fuel performance evaluation technology development

International Nuclear Information System (INIS)

Koo, Yang Hyun; Bang, J. G.; Kim, D. H.

2012-01-01

Alpha version performance code for dual-cooled annular fuel under steady state operation, so called 'DUOS', has been developed applying performance models and proposed methodology. Furthermore, nonlinear finite element module which could be integrated into transient/accident fuel performance code was also developed and evaluated using commercial FE code. The first/second irradiation and PIE test of annular pellet for dual-cooled annular fuel in the world have been completed. In-pile irradiation test DB of annular pellet up to burnup of 10,000 MWd/MTU through the 1st test was established and cracking behavior of annular pellet and swelling rate at low temperature were studied. To do irradiation test of dual-cooled annular fuel under PWR's simulating steady-state conditions, irradiation test rig/rod design/manufacture of mock-up/performance test have been completed through international collaboration program with Halden reactor project. The irradiation test of large grain pellets has been continued from 2002 to 2011 and completed successfully. Burnup of 70,000 MWd/MTU which is the highest burnup among irradiation test pellets in domestic was achieved

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

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

Energy Technology Data Exchange (ETDEWEB)

Oka, S; Becirspahic, S [Institute of Nuclear Sciences Boris Kidric, Heat Transfer Department, Vinca, Beograd (Serbia and Montenegro)

1964-10-15

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

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)

Neutron induced activity in fuel element components

International Nuclear Information System (INIS)

Kjellbert, N.

1978-03-01

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

Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

2007-07-01

Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

2007-01-01

Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

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

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

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, 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

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

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.)

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)

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

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

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

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

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

Behavior of UO2-Zy fuel elements of nuclear power plants up to 40000 MWj/t U

International Nuclear Information System (INIS)

Atabek, R.; Contenson, G. de; Houdaille, B.; Lestiboudois, G.; Vignesoult, N.

1979-01-01

The two principal types of fuel elements studied are unstable oxide elements in 15x15 geometry and stable oxide elements in 17x17. Semi-statistical processing of the fission gas amounts released was performed on different fuel elements at specific burn-up varying between 2000 and 40,000 MWd/t U and linear powers between 250 and 600 W/cm. This study enabled the following essential points to be stated at this burn-up level: the swelling of the oxide appears to be less than predicted by the linear law (S=0.75 %/10,000 MWd/t U); the migration of volatile fission products is relatively low and without effect on the behavior of the fuel element; strong zircaloy 4 claddings exhibit little creep and their hydriding is insignificant. On a more general level, the analyses of the fission gases performed in the fuel elements after irradiation show an increase of the fraction released with specific burn-up at a given linear power or central temperature [fr

Fuel element services

International Nuclear Information System (INIS)

Marta, H.; Alvarez, P.; Jimenez, J.

2006-01-01

Refuelling outages comprise a number of maintenance tasks scheduled long in advance to assure a reliable operation throughout the next cycle and, in the long run, a safer and more efficient plant. Most of these tasks are routine service of mechanical and electrical system and likewise fuel an be considered a critical component as to handling, inspection, cleaning and repair. ENUSA-ENWESA AIE has been working in this area since 1995 growing from fuel repair to a more integrated service that includes new and spent fuel handling, inserts, failed fuel rod detection systems, ultrasonic fuel cleaning, fuel repair and a comprehensive array of inspection and tests related to the reliability of the mechanical components in the fuel assembly, all this, performed in compliance with quality, safety, health physics and any other nuclear standard. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

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

International Nuclear Information System (INIS)

Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro

2011-01-01

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

Cooling flow measurement in fuel elements of the RA-6

International Nuclear Information System (INIS)

Brollo, F; Silin, N

2009-01-01

Under the UBERA6 project for the core change and power increase of the RA-6 reactor, the total coolant flow was increased to meet the requirements imposed by the new operating conditions. The flow through the fuel elements is an important parameter and is difficult to determine due to the geometric complexity of the core. To ensure safe operation of the reactor, adequate safety margins must be kept for all operating conditions. In the present work we performed the direct measurement of the cooling flow rate of a fuel in the reactor core, for which we used a turbine flowmeter built specifically for this use. This helped to confirm previous results obtained during the launch, made by an indirect method based on measuring the pressure difference of the core. The turbine flowmeter was chosen due to its robustness, ease of operation and low disturbance of the input stream to the fuel. We describe the calibration of this instrument and the results of flow measurements made on some of the RA6 reactor fuel elements under conditions of zero power. [es

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

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)

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

Postirradiation examination and evaluation of Peach Bottom fuel test element FTE-6

International Nuclear Information System (INIS)

Wallroth, C.F.; Holzgraf, J.F.; Jensen, D.D.

1977-09-01

Fuel test element FTE-6 was irradiated in the Peach Bottom high-temperature gas-cooled reactor (HTGR) for 645 equivalent full power days. Four fuel varieties, contained in H-327 graphite bodies, were tested. A primary result of this test has been to demonstrate acceptable performance even with calculated high stresses in the graphite bodies. Heterogeneous fuel loadings in the element caused local power peaking and azimuthal power variations, deforming the graphite fuel bodies and thereby causing bowing nearly five times as large as the diametral clearance within the sleeve. The axial stresses resulting from interference between the fuel bodies and sleeve were estimated to have reached 45% of the ultimate material strength at the end of the irradiation. Residual stresses from differential contraction within the fuel body resulted in probable in-plane stress levels of 130% of the material strength at the end-of-life shutdown and of up to 150% of the strength at shutdown during the irradiation cycle. The high in-plane stresses are local peaks at the corners of a sharp notch in the element, which may account for the stresses failing to cause damage. The lack of observable damage, however, indicates that the methods and data used for stress analysis give results that are either fairly accurate or conservative

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

Poolside inspection, repair and reconstitution of LWR fuel elements. Proceedings of a Technical Committee meeting

International Nuclear Information System (INIS)

1998-11-01

The Technical Committee Meeting on Poolside Inspection, repair and reconstruction of LWR Fuel Elements was organize by IAEA upon the recommendations of the International Working Group on Fuel performance Technology and held in Switzerland in October 1997. The purpose of the Meeting was to review the state of art in the area of poolside inspection, repair and reconstruction of light water fuel elements and to evaluate the progress achieved in this area since previous IAEA Meetings on the same topic in 1981 and 1984. The Meeting provided a forum on exchange of information between utilities, fuel designers and other authorities and specialists on a topic of current interest and real concern to industries in many Member States. The respective technologies are widely used or planned to be used in order to identify elementary major causes of fuel failure and to improve fuel utilization by repair and subsequent reuse of fuel elements. The Proceedings includes papers presented at the Meeting each described by a separate abstract

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-01-01

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

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

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.)

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

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

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)

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

International Nuclear Information System (INIS)

Paraschiv, Marius; Turnbull, J.A.

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Makmal, T. [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel); Nuclear Physics and Engineering Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Aviv, O. [Radiation Safety Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Gilad, E., E-mail: gilade@bgu.ac.il [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel)

2016-10-21

A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections. - Highlights: • Simple, inexpensive, safe and flexible experimental setup that can be quickly deployed. • Experimental results are thoroughly corroborated against ORIGEN2 burnup code. • Experimental uncertainty of 9% and 5% deviation between measurements and simulations. • Very high burnup MTR fuel element is examined, with 60% depletion of {sup 235}U. • Impact of highly irregular irradiation regime on burnup evaluation is studied.

Operational limitations of light water reactors relating to fuel performance

International Nuclear Information System (INIS)

Cheng, H.S.

1976-07-01

General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed

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.

Reactor physics assessment of modified 37-element CANDU fuel bundles

International Nuclear Information System (INIS)

Pristavu, R.; Rizoiu, A.

2016-01-01

Reducing the central element diameter in order to improve the total flow area of CANDU fuel bundle and redistribute the power density of all remaining elements was studied in Canada and Korea when considering the effect of aging pressure tube diametral creep. The aim of this paper is to study the modified bundle behavior using the transport codes WIMS and DRAGON. In calculations, a WIMS nuclear data library on 172 energy groups was used. 2-D transport calculations were performed with WIMS and DRAGON, leading to similar results in estimated cell parameters. Additionally, 3-D DRAGON calculations were carried on in order to evaluate the local flux distribution shift, as well as the incremental cross sections for supercells containing modified CANDU bundles and reactivity devices. The overall effect of using modified fuel bundles was meaningless for both cell and supercell parameters, thus ensuring this possibility of fuel improvement for thermal-hydraulic purposes only. (authors)

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.

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

Research reactors. Problems of fuel element enrichment reduction. Deliberations and comments

International Nuclear Information System (INIS)

1978-10-01

This paper summarises the main data from the major research reactors in the Federal Republic of Germany utilising highly enriched uranium (HEU) and presently available fuel technology for their fuel elements. The required modification for an adaption of the fabrication to lower enriched fuel are considered as well as the consequences on reactor performance operation and licensing. On the basis of past experience with reactor modifications a rough estimate of 82 months is given for the conversion of a reactor to a modified type of fuel and of 70 months for a fuel test program. The conclusions reflect the own calculations and data from other papers submitted to INFCE-WG 8C

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

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

Examination of the surface coatings removed from K-East Basin fuel elements

International Nuclear Information System (INIS)

Abrefah, J.; Marschman, S.C.; Jenson, E.D.

1998-05-01

This report provides the results of studies conducted on coatings discovered on the surfaces of some N-Reactor spent nuclear fuel (SNF) elements stored at the Hanford K-East Basin. These elements had been removed from the canisters and visually examined in-basin during FY 1996 as part of a series of characterization tests. The characterization tests are being performed to support the Integrated Process Strategy developed to package, dry, transport, and store the SNF in an interim storage facility on the Hanford site. Samples of coating materials were removed from K-East canister elements 2350E and 2540E, which had been sent, along with nine other elements, to the Postirradiation Testing Laboratory (327 Building) for further characterization following the in-basin examinations. These coating samples were evaluated by Pacific Northwest National Laboratory using various analytical methods. This report is part of the overall studies to determine the drying behavior of corrosion products associated with the K-Basin fuel elements. Altogether, five samples of coating materials were analyzed. These analyses suggest that hydration of the coating materials could be an additional source of moisture in the Multi-Canister Overpacks being used to contain the fuel for storage

Examination of the surface coating removed from K-East Basin fuel elements

Energy Technology Data Exchange (ETDEWEB)

Abrefah, J.; Marschman, S.C.; Jenson, E.D.

1998-05-01

This report provides the results of studies conducted on coatings discovered on the surfaces of some N-Reactor spent nuclear fuel (SNF) elements stored at the Hanford K-East Basin. These elements had been removed from the canisters and visually examined in-basin during FY 1996 as part of a series of characterization tests. The characterization tests are being performed to support the Integrated Process Strategy developed to package, dry, transport, and store the SNF in an interim storage facility on the Hanford site. Samples of coating materials were removed from K-East canister elements 2350E and 2540E, which had been sent, along with nine other elements, to the Postirradiation Testing Laboratory (327 Building) for further characterization following the in-basin examinations. These coating samples were evaluated by Pacific Northwest National Laboratory using various analytical methods. This report is part of the overall studies to determine the drying behavior of corrosion products associated with the K-Basin fuel elements. Altogether, five samples of coating materials were analyzed. These analyses suggest that hydration of the coating materials could be an additional source of moisture in the Multi-Canister Overpacks being used to contain the fuel for storage.

Design and main characteristics of HTGR fuel elements

International Nuclear Information System (INIS)

Chernikov, A.S.; Kolesov, V.S.; Permyakov, L.N.; Koshelev, Yu.V.; Mikhajlichenko, L.I.

1983-01-01

Two types of spherical fuel elements and coated particles were investigated under the operating conditions of the high temperature reactors in the Soviet Union (VGR-50 and VG-400). This paper gives the main characteristics of spherical fuel elements (thermal conductivity, static and dynamic strength, wear resistance, release of gaseous fission products, etc.) as determined in test facilities. (author)

Transfer flask for hot active fuel elements

International Nuclear Information System (INIS)

Aubert, Roger; Moutard, Daniel.

1980-01-01

This invention concerns a flask for transporting active fuel elements removed from a nuclear reactor vessel, after only a few days storage and hence cooling, either within a nuclear power station itself or between such a station and a near-by storage area. This containment system is not a flask for conveyance over long and medium distances. Specifically, the invention concerns a transport flask that enables hot fuel elements to be cooled, even in the event of accidents [fr

The industrial production of fuel elements; La fabrication en france des elements combustibles

Energy Technology Data Exchange (ETDEWEB)

Boussard, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Nadal, J [Societe Industrielle de Combustible Nucleaire (SICN), 75 - Paris (France); Pellen, A [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques (CERCA), 75 - Paris (France)

1964-07-01

The authors deal successively with the industrial production of fuel elements for power reactors of the natural uranium-graphite-gas type, and more particularly for the EDF power stations, and with the industrial production of fuel elements containing enriched uranium designed for swimming-pool type reactors. 1. part: advanced fuel elements for the EDF reactors. After recalling the characteristics of the fuel elements now being produced industrially for the Marcoule and Chinon reactors, the authors give the various steps leading to the industrial production of a new type of fuel element both as concerns the can, and in certain cases the graphite sleeve, and the fuel itself. As for as the production of the fuel is concerned, they describe the various operations, stressing the original aspects of the production and of the equipment such as: - casting in hot moulds, - thermal treatments, of Uranium containing 1% in weight molybdenum, - welding of the pellets for closing the tubes of uranium, - canning, - controls in the various steps. As far as can production is concerned they show why the extruded can was replaced by a machined can and give a few characteristics of the equipment used as well as the controls effected. They give also some details concerning the production and machining of the sleeves. After recalling the state of the nuclear fuel industry in France in mid-1964 the authors stress the economic aspects of the production of fuel elements. They show the relative importance of capital costs on the cost price of the fuel itself and examine the various items involved. They analyse the cost price of a completed fuel element using present date knowledge. In conclusion the authors show the particular points which should be the subject of future efforts in order to decrease the cost of a production which is perhaps delicate but now will define, and review the development of this new industrial branch. 2. part: industrial production of fuel elements for swimming

Preparation of spherical fuel elements for HTR-PM in INET

International Nuclear Information System (INIS)

Xiangwen, Zhou; Zhenming, Lu; Jie, Zhang; Bing, Liu; Yanwen, Zou; Chunhe, Tang; Yaping, Tang

2013-01-01

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

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

Design verification testing for fuel element type CAREM

International Nuclear Information System (INIS)

Martin Ghiselli, A.; Bonifacio Pulido, K.; Villabrille, G.; Rozembaum, I.

2013-01-01

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

International experience and status of fuel element performance and modelling for water reactors

International Nuclear Information System (INIS)

Gittus, J.H.

1982-04-01

Current knowledge concerning water reactor fuel performance and technology is reviewed (212 references). The emphasis is on aspects of in-reactor performance including behaviour in accidents. Computer models for predicting fuel behaviour during the ordinary running of the reactor and during accidents are described. These codes include COMETHE, HOTROD, SLEUTH-SEER and FRAPCON. Their agreement with experimental data is examined. (U.K.)

Legal questions concerning the termination of spent fuel element reprocessing

International Nuclear Information System (INIS)

John, Michele

2005-01-01

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

Fission product release from defected nuclear reactor fuel elements

International Nuclear Information System (INIS)

Lewis, B.J.

1983-01-01

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

Criticality analysis of the CAREM-25 reactor irradiated fuel elements storage pool

International Nuclear Information System (INIS)

Albornoz, A.F.; Jatuff, F.E.; Gho, C.J.

1993-01-01

A criticality safety analysis of the irradiated fuel element pool storage of the CAREM-25 reactor was performed. The CAREM project is property of the Comision Nacional de Energia Atomica (CNEA) of Argentine, and it is being executed by INVAP S.E. difficult evaluation of the CAREM core (relatively high -3,4%- enriched U O 2 , Gd 2 O 3 burnable absorber in different densities, or criticality achievement with as few as 7 fuel elements is inherited by the pool storage. The lattice code CONDOR 1.1 was used for investigating the problem scene, and some results compared on the Monte Carlo codes MONK 5.0 and MONK 6.3. Circular and square tubes of 304-L stainless steel, borated steel and boral B 4 C in Al) were tested as suitable channels for fuel element containment, in square and hexagonal arrays; in addition, burnup, burnable absorber concentration, Sm and leakage credits were determined. It was found that the critical is strongly dependent on the separation of the fuel elements in the pool. Out-of-nominal conditions were investigated too, showing that the loss of coolant and the change in temperature and density conditions in the storage lead to an increase in reactivity, but the system's reactivity remains near the safety limits. (author)

Development of LWR fuel performance code FEMAXI-6

International Nuclear Information System (INIS)

Suzuki, Motoe

2006-01-01

LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

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

International Nuclear Information System (INIS)

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

1998-06-01

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 fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0

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

Improved techniques for appendage attachment to PHWR fuel elements

International Nuclear Information System (INIS)

Raj, R.N.J.; Laxminarayana, B.; Narayanan, P.S.A.; Gupta, U.C.; Varma, B.P.; Sinha, K.K.

1995-01-01

Nuclear Fuel Complex, India switched-over to split-wart type PHWR fuel bundles in mid-80s. Since then over 60,000 bundles of this type have been fabricated for Indian PHWRs. After considering various technical aspects, resistance welding was chosen for appendage attachment to the fuel elements. The paper describes experiences in scaling up of the technique to industrial production of PHWR fuel bundles, design and development of special-purpose equipment for this purpose, and the QA procedures employed for regular production. It also deals with appendage welding of 37 Element fuel bundles and improvements planned in the appendage welding process. (author)

Fuel element for a nuclear reactor

International Nuclear Information System (INIS)

Rau, P.

1981-01-01

Fuel elements which consist of parallel longitudinal fuel rods of circular crossection, can be provided with spiral distance pieces, by which the fuel rods support one another, if they are collected together by an outer enclosure. According to the invention, the enclosure includes several strips extending over a small fraction of the rod length, which are connected together by a skeleton rod instead of a fuel rod. The strips can be composed of flat parts which are connected together by the skeleton rod acting as a hinge. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.; Papastergiou, C.

1992-01-01

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed

CARA CVN: inherently safe fuel element for PHWR power plants

International Nuclear Information System (INIS)

Brasnarof, Daniel O.; Lestani, Hector A.; Agueda, Horacio C.; Marino, Armando C.; Florido, Pablo C.; Daverio, Hernando

2007-01-01

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

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

Fuel element performance computer modelling

International Nuclear Information System (INIS)

Locke, D.H.

1978-01-01

The meeting was attended by 88 participants from 17 countries. Altogether 47 papers were presented. The majority of the presentations contained a description of the equations and solutions used to describe and evaluate some of the physical processes taking place in water reactor fuel pins under irradiation. At the same time, particular attention was paid to the ''bench marking'' of the codes wherein solutions arrived at for particular experiments are compared with the results at the experiments

Results of tests under normal and abnormal operating conditions concerning LMFBR fuel element behaviour

International Nuclear Information System (INIS)

Languille, A.; Bergeonneau, P.; Essig, C.; Guerin, Y.

1985-04-01

The objective of this paper is to improve the knowledge on LMFBR fuel element behaviour during protected and unprotected transients in RAPSODIE and PHENIX reactors in order to evaluate its reliability. The range of the tests performed in these reactors is sufficiently large to cover normal and also extreme off normal conditions such as fuel melting. Results of such tests allow to better establish transient design limits for reactor structural components in particular for fuel pin cladding which play a lead role in controlling the accident sequence. Three main topics are emphasized in this paper: fuel melting during slow over-power excursions; influence of the fuel element geometrical evolution on reactivity feedback effects and reactor dynamic behaviour; clad damage evaluation during a transient (essentially very severe loss of flow)

Development and testing of the EDF-2 reactor fuel element

International Nuclear Information System (INIS)

Delpeyroux, P.

1964-01-01

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

Improved nuclear fuel element

International Nuclear Information System (INIS)

1980-01-01

The invention is of a nuclear fuel element which comprises a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium and mixtures thereof, and an elongated composite cladding container comprising a zirconium alloy tube containing constituents other than zirconium in an amount greater than about 5000 parts per million by weight and an undeformed metal barrier of moderate purity zirconium bonded to the inside surface of the alloy tube. The container encloses the core so as to leave a gap between the container and the core during use in a nuclear reactor. The metal barrier is of moderate purity zirconium with an impurity level on a weight basis of at least 1000ppm and less than 5000ppm. Impurity levels of specific elements are given. Variations of the invention are also specified. The composite cladding reduces chemical interaction, minimizes localized stress and strain corrosion and reduces the likelihood of a splitting failure in the zirconium alloy tube. Other benefits are claimed. (U.K.)

Fuel element concept for long life high power nuclear reactors

Science.gov (United States)

Mcdonald, G. E.; Rom, F. E.

1969-01-01

Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

Finite element analysis of BWR fuel channel buckling during a seismic event

International Nuclear Information System (INIS)

Kinoshita, Mika; Iwamoto, Yuji; Ledford, Kevin; Cantonwine, Paul

2014-01-01

This paper documents the predicted response of three BWR fuel channel designs in bending using a typical moment profile for GNF fuel designs. The bending performance of the fuel channel is predicted using ANSYS, a finite element modeling tool. Specifically, linear and non-linear buckling analyses were performed to determine the onset of elastic buckling, which causes a wavy structure on the compression face in bending that might also increase channel – control blade friction, and to determine to onset of channel collapse, which causes permanent deformation and would inhibit control rod insertion. The three channel designs considered in this paper are the 0.080 inch uniform channel, the 0.100 inch uniform channel and the 0.120 inch uniform channel at the beginning of fuel life (BOL) and at the end of fuel life (EOL). (author)

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

International Nuclear Information System (INIS)

Masson, M.

1986-12-01

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

Dimensional measurements and eddy currents control of the sheath integrity for a set of irradiated candu fuel elements

International Nuclear Information System (INIS)

Gheorghe, G.; Man, I.

2015-01-01

During irradiation in the nuclear reactor, fuel elements undergo dimensional and structural changes, and changes of sheath surface condition as well, which can lead to damages and even loss of integrity. This paper presents the results of dimensional measurements and of examination technique with eddy currents for three fuel elements of an irradiated CANDU fuel bundle. One of the fuel elements (FE), which is studied in detail, presented a crack about 40 mm long. The purpose of these nondestructive examination techniques is to determine those parameters that characterize the behavior and performance of nuclear fuel operation. This paper contains images of defects and interpretations of the causes of their occurrence. (authors)

Compilation of papers presented to the KTG conference on 'Advanced LWR fuel elements: Design, performance and reprocessing', 17-18 November 1988, Karlsruhe Nuclear Research Center

International Nuclear Information System (INIS)

Bahm, W.

1989-05-01

The two expert groups of the Nuclear Society (KTG), 'chemistry and waste disposal' and 'fuel elements' discussed interdisciplinary problems concerning the development and reprocessing of advanced fuel elements. The 10 lectures deal with waste disposal, mechanical layout, operating behaviour, operating experiences and new developments of fuel elements for water moderated reactors as well as operational experiences of the Karlsruhe reprocessing plant (WAK) with reprocessing of high burnup LWR and MOX fuel elements, the distribution of fission products, the condition of the fission products during dissolution and with the effects of the higher burnup of fuel elements on the PUREX process. (DG) [de

Repurposing an irradiated instrumented TRIGA fuel element for regular use

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Determining axial perturbation of the reactor cell by introducing construction material into reactor fuel element

International Nuclear Information System (INIS)

Dimitrijevic, V.

1975-01-01

Axial distribution of thermal neutrons in the center and on the surface of a fuel element in the presence of aluminium was measured by reactor cell perturbation method. Experiments were performed by Dy activation foils using 20 mm thick Al disc placed between two fuel elements. Measured values of thermal neutron flux distribution in the reactor cell were compared to calculated values obtained by one-group neutron diffusion method

Reliability analysis of dispersion nuclear fuel elements

Science.gov (United States)

Ding, Shurong; Jiang, Xin; Huo, Yongzhong; Li, Lin an

2008-03-01

Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

Reliability analysis of dispersion nuclear fuel elements

Energy Technology Data Exchange (ETDEWEB)

Ding Shurong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: dsr1971@163.com; Jiang Xin [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: yzhuo@fudan.edu.cn; Li Linan [Department of Mechanics, Tianjin University, Tianjin 300072 (China)

2008-03-15

Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

Examination of irradiated fuel elements using gamma scanning technique

International Nuclear Information System (INIS)

Ichim, O.; Mincu, M.; Man, I.; Stanica, M.

2016-01-01

The purpose of this paper is to validate the gamma scanning technique used to calculate the activity of gamma fission products from CANDU/TRIGA irradiated fuel elements. After a short presentation of the equipments used and their characteristics, the paper describes the calibration technique for the devices and how computed tomography reconstruction is done. Following the previously mentioned steps is possible to obtain the axial and radial profiles and the computed tomography reconstruction for calibration sources and for the irradiated fuel elements. The results are used to validate the gamma scanning techniques as a non-destructive examination method. The gamma scanning techniques will be used to: identify the fission products in the irradiated CANDU/TRIGA fuel elements, construct the axial and radial distributions of fission products, get the distribution in cross section through computed tomography reconstruction, and determine the nuclei number and the fission products activity of the irradiated CANDU/TRIGA fuel elements. (authors)

Fuel element replacement and cooling water activity at the musashi reactor

International Nuclear Information System (INIS)

Nozaki, Tetsuya; Honda, Teruyuki; Horiuchi, Norikazu; Aizawa, Otohiko; Sato, Tadashi

1989-01-01

The Musashi Institute of Technology Research Reactor (TRIGA 11, 100 kW) has been operated without serious problems since 1963. However, because there is no more spare fuel element, it was necessary to decide how to solve the problem. In the end, it was decided to obtain many stainless steel-clad fuel elements and operate with those fuel elements only, under the auspices of the Ministry of Education, Science and Culture. The bulk shielding experimental pool was remodeled as the storage for spent fuel elements, where the neutrons from the thermalizing column were shielded with cadmium and boron polyethylene plates. The equipment for transferring spent fuel elements was built and temporarily set up between the core tank and the new storage. These works were started in 1983, and finished in 1985. After the reactor was restarted, the count rate of the conventional cooling water monitor which was set in the cooling system using a GM counter drastically decreased. The spent fuel storage, the equipment and the works for fuel transfer, and the radioactivity of cooling water are reported. (K.I.)

Cna 1 spent fuel element interim dry storage system thermal analysis

International Nuclear Information System (INIS)

Hilal, R. E; Garcia, J. C; Delmastro, D. F

2006-01-01

At the moment, the Atucha I Nuclear Power Plant (Cnea-I) located in the city of Lima, has enough room to store its spent fuel (Sf) in their two pools spent fuel until about 2015.In case of life extension a spend fuel element interim dry storage system is needed.Nucleolectrica Argentina S.A. (N A-S A) and the Comision Nacional de Energia Atomica (Cnea), have proposed different interim dry storage systems.These systems have to be evaluated in order to choose one of them.The present work's objective is the thermal analysis of one dry storage alternative for the Sf element of Cna 1.In this work a simple model was developed and used to perform the thermal calculations corresponding to the system proposed by Cnea.This system considers the store of sealed containers with 37 spent fuels in concrete modules.Each one of the containers is filled in the pool houses and transported to the module in a transference cask with lead walls.Fulfill the maximum cladding temperature requirement ( [es

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Performance of PARR-1 with LEU Fuel

International Nuclear Information System (INIS)

Pervez, S.; Latif, M.; Bokhari, I.H.; Bakhtyar, S.

2005-01-01

Pakistan Research Reactor (PARR-1) went critical in 1965 with HEU fuel. The reactor core was converted to LEU fuel with power upgradation from 5 MW to 10 MW in 1992. The reactor has been operated with LEU fuel for about 10,000 hours and has produced about 66,000 MWh energy up to now. Average burn up of the irradiated fuel is about 42 %. The fuel performance during the last 12 years has been excellent. Post irradiation visual inspection of the fuel has revealed no abnormality. During operation there have been no signs of releases in the pool water establishing the full integrity of this fuel. The reactor has been mainly utilized for radioisotope production, beam tube experiments including neutron diffraction studies, neutron radiography etc. Studies have been completed to operate the reactor with a mixed core (HEU + LEU) to utilize the less burned HEU fuel elements. A major project of production of fission Moly using PARR-1 is in the final stages. (author)

Neutronic analysis of a fuel element with variations in fuel enrichment and burnable poison

Energy Technology Data Exchange (ETDEWEB)

Faria, Rochkhudson B. de; Martins, Felipe; Velasquez, Carlos E.; Cardoso, Fabiano; Fortini, Angela; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

In this work, the goal was to evaluate the neutronic behavior during the fuel burnup changing the amount of burnable poison and fuel enrichment. For these analyses, it was used a 17 x 17 PWR fuel element, simulated using the 238 groups library cross-section collapsed from ENDF/BVII.0 and TRITON module of SCALE 6.0 code system. The results confirmed the effective action of the burnable poison in the criticality control, especially at Beginning Of Cycle (BOC) and in the burnup kinetics, because at the end of the fuel cycle there was a minimal residual amount of neutron absorbers ({sup 155}Gd and {sup 157}Gd), as expected. At the end of the cycle, the fuel element was still critical in all simulated situations, indicating the possibility of extending the fuel burn. (author)

Nuclear reactor fuel element sub-assemblies

International Nuclear Information System (INIS)

Hill, G.D.; Trevalion, P.A.

1977-01-01

A fuel element sub-assembly for a liquid metal cooled fast reactor is described. It comprises a bundle of fuel pins enclosed by a tubular wrapper having a lower end journal for plugging into an upper aperture in a core supporting structure and a spike bar with an articulated bush for engaging a lower aperture in the core supporting structure. The articulated bush is retained on a spherical end portion of the spike bar by a pair of parallel retaining pins arranged transversely and disposed one each side of the spike bar. The pins are tubular and collapsible at a predetermined loading to enable the spherical end portion to pass between them. The articulated bush has an internal groove for engagement by a lifting grab, this groove being formed in a bore for receiving the spherical end portion of the spike bar. The construction lessens liability to rattling of the fuel element sub-assemblies and aids removal for replacement. (U.K.)

Postirradiation results and evaluation of helium-bonded uranium--plutonium carbide fuel elements irradiated in EBR-II. Interim report

International Nuclear Information System (INIS)

Latimer, T.W.; Barner, J.O.; Kerrisk, J.F.; Green, J.L.

1976-02-01

An evaluation was made of the performance of 74 helium-bonded uranium-plutonium carbide fuel elements that were irradiated in EBR-II at 38-96 kW/m to 2-12 at. percent burnup. Only 38 of these elements have completed postirradiation examination. The higher failure rate found in fuel elements which contained high-density (greater than 95 percent theoretical density) fuel than those which contained low-density (77-91 percent theoretical density) fuel was attributed to the limited ability of the high-density fuel to swell into the void space provided in the fuel element. Increasing cladding thickness and original fuel-cladding gap size were both found to influence the failure rates for elements containing low-density fuel. Lower cladding strain and higher fission-gas release were found in high-burnup fuel elements having smear densities of less than 81 percent. Fission-gas release was usually less than 5 percent for high-density fuel, but increased with burnup to a maximum of 37 percent in low-density fuel. Maximum carburization in elements attaining 5-10 at. percent burnup and clad in Types 304 or 316 stainless steel and Incoloy 800 ranged from 36-80 μm and 38-52 μm, respectively. Strontium and barium were the fission products most frequently found in contact with the cladding but no penetration of the cladding by uranium, plutonium, or fission products was observed

Design of experiments for test of fuel element reliability

International Nuclear Information System (INIS)

Boehmert, J.; Juettner, C.; Linek, J.

1989-01-01

Changes of fuel element design and modifications of the operational conditions have to be tested in experiments and pilot projects for nuclear safety. Experimental design is an useful statistical method minimizing costs and risks for this procedure. The main problem of our work was to investigate the connection between failure rate of fuel elements, sample size, confidence interval, and error probability. Using the statistic model of the binomial distribution appropriate relations were derived and discussed. A stepwise procedure based on a modified sequential analysis according to Wald was developed as a strategy of introduction for modifications of the fuel element design and of the operational conditions. (author)

Method of removing crud deposited on fuel element clusters

International Nuclear Information System (INIS)

Yokota, Tokunobu; Yashima, Akira; Tajima, Jun-ichiro.

1982-01-01

Purpose: To enable easy elimination of claddings deposited on the surface of fuel element. Method: An operator manipulates a pole from above a platform, engages the longitudinal flange of the cover to the opening at the upper end of a channel box and starts up a suction pump. The suction amount of the pump is set such that water flow becomes within the channel box at greater flow rate than the operational flow rate in the channel box of the fuel element clusters during reactor operation. This enables to remove crud deposited on the surface of individual fuel elements with ease and rapidly without detaching the channel box. (Moriyama, K.)

Transportation of irradiated fuel elements

International Nuclear Information System (INIS)

Preece, A.H.

1980-01-01

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

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

BWR fuel performance

International Nuclear Information System (INIS)

Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

1979-01-01

The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

Dynamic characterization of the CAREM fuel element prototype

International Nuclear Information System (INIS)

Ghiselli, Alberto M.; Fiori, Jose M.; Ibanez, Luis A.

2004-01-01

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

Probability analysis of WWER-1000 fuel elements behavior under steady-state, transient and accident conditions of reactor operation

International Nuclear Information System (INIS)

Tutnov, A.; Alexeev, E.

2001-01-01

'PULSAR-2' and 'PULSAR+' codes make it possible to simulate thermo-mechanical and thermo-physical parameters of WWER fuel elements. The probabilistic approach is used instead of traditional deterministic one to carry out a sensitive study of fuel element behavior under steady-state operation mode. Fuel elements initial parameters are given as a density of the probability distributions. Calculations are provided for all possible combinations of initial data as fuel-cladding gap, fuel density and gas pressure. Dividing values of these parameters to intervals final variants for calculations are obtained . Intervals of permissible fuel-cladding gap size have been divided to 10 equal parts, fuel density and gas pressure - to 5 parts. Probability of each variant realization is determined by multiplying the probabilities of separate parameters, because the tolerances of these parameters are distributed independently. Simulation results are turn out in the probabilistic bar charts. The charts present probability distribution of the changes in fuel outer diameter, hoop stress kinetics and fuel temperature versus irradiation time. A normative safety factor is introduced for control of any criterion realization and for determination of a reserve to the criteria failure. A probabilistic analysis of fuel element behavior under Reactivity Initiating Accident (RIA) is also performed and probability fuel element depressurization under hypothetical RIA is presented

Economical benefits for the use of slightly enriched fuel elements at the Atucha-I nuclear power plant

International Nuclear Information System (INIS)

Sidelnik, J.I.; Sosa, M.A.

1987-01-01

The fuel represents a very important factor in the operative cost of the Atucha I nuclear power plant. This cost is drastically reduced with the use of fuel elements of slightly enriched uranium. The annual saving is analyzed with actual values for fuel elements with an enrichment of 0.85% by weight of U-235. With the reactor core in equilibrium state the annual saving achieved is approximately 7.5-10 u$s. According to the present irradiation plan, the benefit for the transition period is studied. An analysis of the sensitivity to differential increments in factors determining the cost of fuel elements or to changes in manufacturing losses is also performed, calculating its effect on the waste, the storage of irradiated elements and the amount of UO 2 required. (Author)

WELWING, Material Buckling for HWR with Annular Fuel Elements

International Nuclear Information System (INIS)

Grosskopf, O.G.P.

1973-01-01

1 - Nature of the physical problem solved: WELWING was developed to calculate the material buckling of reactor systems consisting of annular fuel elements in heavy water as moderator for various moderator to fuel ratios. The moderator to fuel ratio for the maximum material buckling for the particular system is selected automatically and the corresponding material buckling is calculated. 2 - Method of solution: The method used is an analytical solution of the one-group diffusion equations with various corrections and approximations. 3 - Restrictions on the complexity of the problem: Up to 32 different materials in the fuel element may be used

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

International Nuclear Information System (INIS)

Saurwein, J.J.

1982-10-01

Fifty-four fuel and reflector elements irradiated in core segment 2 of the Fort St. Vrain high-temperature gas-cooled reactor (HTGR) were nondestructively examined. The time- and volume-averaged graphite irradiation temperatures for the elements ranged from approx. 350 0 to 750 0 C. The element-averaged fast neutron fluences ranged from approx. 0.2 to 1.6 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/. The elements, except for two fuel elements in which single localizeed cracks developed during irradiation, were in excellent condition. No evidence was observed of significant graphite oxidation or mechanical interaction beween elements. The cracks in the two elements did not affect their performance or handling. These elements were, otherwise, in excellent condition. Nearly all elements shrank in both the axial and radial directions, but the dimensional changes were relatively small

performance calculations of gadolinium oxide and boron nitride coated fuel

International Nuclear Information System (INIS)

Tanker, E.; Uslu, I.; Disbudak, H.; Guenduez, G.

1997-01-01

A comparative study was performed on the behaviour of natural uranium dioxide-gadolinium oxide mixture fuel and boron nitride coated low enriched fuel in a pressurized water reactor. A fuel element containing one burnable poison fuel pins was modeled with the computer code WIMS, and burn-up dependent critically, fissile isotope inventory and two dimensional power distribution were obtained. Calculations were performed for burnable poison fuels containing 5% and 10% gadolinium oxide and for those coated with 1μ,5μ and 10μ of boron nitride. Boron nitride coating was found superior to gadolinium oxide on account of its smoother criticality curve, lower power peaks and insignificant change in fissile isotope content

Calculation of plate temperatures in a Mk 4 LEU fuel element

International Nuclear Information System (INIS)

Haack, K.

1988-09-01

A calculation method for estimating the axial temperature distributions of each tube in each of the 26 fuel elements of the DR 3 core is described and demonstrated. With input data for fuel element power, D2O outlet temperature and main D2O circulator combination, a computer code calculates all important temperatures in the fuel element. 11 tabs., 32 ills. 8 refs. (author)

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

Chemical aspects of pellet-cladding interaction in light water reactor fuel elements

International Nuclear Information System (INIS)

Olander, D.R.

1982-01-01

In contrast to the extensive literature on the mechanical aspects of pellet-cladding interaction (PCI) in light water reactor fuel elements, the chemical features of this phenomenon are so poorly understood that there is still disagreement concerning the chemical agent responsible. Since the earliest work by Rosenbaum, Davies and Pon, laboratory and in-reactor experiments designed to elucidate the mechanism of PCI fuel rod failures have concentrated almost exclusively on iodine. The assumption that this is the reponsible chemical agent is contained in models of PCI which have been constructed for incorporation into fuel performance codes. The evidence implicating iodine is circumstantial, being based primarily upon the volatility and significant fission yield of this element and on the microstructural similarity of the failed Zircaloy specimens exposed to iodine in laboratory stress corrosion cracking (SCC) tests to cladding failures by PCI

Nuclear fuel element