Considerations about the utilization of electrically heated rods used for simulation of nuclear fuel pins

International Nuclear Information System (INIS)

Lima, R. de C.F. de; Carajilescov, P.

1987-01-01

The dinamic behavior of electrically heated rods used for simulation of nuclear fuel pins in nuclear power transients, is analysed by the application of the lumped parameter and the finite difference methods. Deviations of the rods surface conditions, for extreme accidental transient conditions are presented and discussed. (author) [pt

Performance of refractory alloy-clad fuel pins

International Nuclear Information System (INIS)

Dutt, D.S.; Cox, C.M.; Millhollen, M.K.

1984-12-01

This paper discusses objectives and basic design of two fuel-cladding tests being conducted in support of SP-100 technology development. Two of the current space nuclear power concepts use conventional pin type designs, where a coolant removes the heat from the core and transports it to an out-of-core energy conversion system. An extensive irradiation testing program was conducted in the 1950's and 1960's to develop fuel pins for space nuclear reactors. The program emphasized refractory metal clad uranium nitride (UN), uranium carbide (UC), uranium oxide (UO 2 ), and metal matrix fuels (UCZr and BeO-UO 2 ). Based on this earlier work, studies presented here show that UN and UO 2 fuels in conjunction with several refractory metal cladding materials demonstrated high potential for meeting space reactor requirements and that UC could serve as an alternative but higher risk fuel

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

System for measuring spacer pin pitch in a nuclear fuel assembly

International Nuclear Information System (INIS)

Isono, Kenji; Tateishi, Yoshinori; Mano, Tadashi.

1975-01-01

Object: To reduce the period for discriminating whether or not spacer pin pitch is satisfactory by simultaneously inserting a number of reference rods into a nuclear fuel assembly spacer ring element of a reactor and arranging them such that they can be simultaneously withdrawn to simplify the withdrawing operation. Structure: A spacer provided with a ring element which clamps a nuclear fuel element is supported on a spacer support with a rod secured to the support as a guide and is secured to the support by securing means. A vertically movable structure with a reference rod provided upright and thru-holes formed in two support plates provided in the same row as the spacer ring element is operated by a fluid pressure mechanism to simultaneously insert the reference rod into the spacer ring element. The reference rod is mounted in support plates via ball bearings such that it is slightly movable in the horizontal direction, and it is aligned with respect to the core of the ring element. The intercore distance of the reference rod is measured with the reference rod inserted in the ring element, thereby measuring the space pin pitch. From the results of measurement, discrimination as to whether the spacer is satisfactory or not is made. (Kamimura, M.)

Means for supporting nuclear fuel

International Nuclear Information System (INIS)

Cocker, P.; Price, M.A.

1975-01-01

Reference is made to means for supporting nuclear fuel pins in a reactor coolant channel and the problems that arise in this connection. For reasons of nuclear reactivity and neutron economy 'parasitic' material in a reactor core must be kept to a minimum, whilst for heat transfer reasons the use of fuel pins of large cross-sectional areas should be avoided. Fuel pins tend to be long thin objects having a can of minimum thickness and typically a pin may have a length/diameter ratio of about 500/1 and for fast reactor fuel pins, the outside diameter may be about 0.2 inch. The long slender pins must also be spaced very close together. A fast reactor fuel assembly may involve 200 to 300 fuel pins, each a few tenths of an inch in diameter, supported end on to coolant flowing up a channel of about 22 square inches in total area. The pins have a heavy metal oxide filling and require support. Details are given of a suitable method of support. Such support also allows withdrawal of pins from a fuel channel without the risk of breach of the can, after irradiation. (U.K.)

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

Science.gov (United States)

Bowles, K. J.

1974-01-01

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

State and parameter estimation in a nuclear fuel pin using the extended Kalman filter

International Nuclear Information System (INIS)

Feeley, J.J.

1979-03-01

The Kalman filter is a powerful tool for the design and analysis of stochastic systems. The general nature of the method permits such diverse applications as on-line state estimation in optimal control systems, as well as state and parameter estimation applications in data analysis and system identification. However, while there have been a large number of Kalman filter applications in the aerospace industry, there have been relatively few in the nuclear industry. The report describes some initial efforts made at the Idaho National Engineering Laboratory to gain experience with the methods of Kalman filtering and to test their applicability to nuclear engineering problems. Two specific cases were considered: first, a real-time state estimation problem using a hybrid computer where the process was simulated on the analog portion of the computer, and the Kalman filter was programmed on the digital portion; second, a system identification problem where a digital extended Kalman filter program was used to estimate states and parameters in a nuclear fuel pin using data generated both by actual experiments and computer simulations. The report contains a derivation of the Kalman filter equations, a development of the mathematical model of the nuclear fuel pin, a description of the computer programs used in the analysis, and a discussion of the results obtained

Device for supporting a fuel pin cluster within a nuclear reactor fuel assembly wrapper

International Nuclear Information System (INIS)

Marmonier, P.; Mesnage, B.; Teulon, J.; Vayra, J.; Venobre, H.

1976-01-01

A supporting member for an array of parallel rails each carrying one row of slidably mounted pins of a fuel cluster is placed coaxially at the lower end of a vertical fuel assembly wrapper. Each parallel rail is provided at each end with a downward extension and terminal lug which engages in a lateral groove formed in the periphery of the supporting member in order to lock and maintain the rails and the fuel pins in uniformly spaced relation within the fuel assembly wrapper. 10 claims, 8 figures

Cesium chemistry in GCFR fuel pins

International Nuclear Information System (INIS)

Fee, D.C.; Johnson, C.E.

1979-01-01

The fuel rod design for the Gas Cooled Fast-Breeder Reactor (GCFR) is similar to that employed for the Liquid Metal Fast Breeder Reactor (LMFBR) with the exception of the unique features inherent to the use of helium as the coolant. These unique design features include the use of (1) vented and pressure-equalized fuel rods, and (2) ribbed cladding along 75% of the fuel section. The former design feature enables reduction in cladding thickness and prevention of possible creep collapse of the cladding due to the high coolant pressure (8.5 MPa). The latter design feature brings about improved heat transfer characteristics. Each GCFR fuel rod is vented to a manifold whereby gaseous fission products diffusing out of the fuel pin are retained on charcoal traps. As a result, the internal pressure of a GCFR fuel pin does not increase during irradiation. In addition, the venting system also maintains the pressure within the fuel pin slightly below (0.3 to 0.5 MPa) the coolant pressure outside the fuel pin. Consequently, should a breach occur in the cladding, helium flows into the breached fuel pin thereby minimizing fission product contamination of the coolant. These desirable aspects of a GCFR fuel pin can be maintained only as long as axial gas transport paths are available and operating within the fuel pin

Transient survivability of LMR oxide fuel pins

International Nuclear Information System (INIS)

Weber, E.T.; Pitner, A.L.; Bard, F.E.; Culley, G.E.; Hunter, C.W.

1986-01-01

Fuel pin integrity during transient events must be assessed for both the core design and safety analysis phases of a reactor project. A significant increase in the experience related to limits of integrity for oxide fuel pins in transient overpower events has been realized from testing of fuel pins irradiated in FFTF and PFR. Fourteen FFTF irradiated fuel pins were tested in TREAT, representing a range of burnups, overpower ramp rates and maximum overpower conditions. Results of these tests along with similar testing in the PFR/TREAT program, provide a demonstration of significant safety margins for oxide fuel pins. Useful information applied in analytical extrapolation of fuel pin test data have been developed from laboratory transient tests on irradiated fuel cladding (FCTT) and on unirradiated fuel pellet deformation. These refinements in oxide fuel transient performance are being applied in assessment of transient capabilities of long lifetime fuel designs using ferritic cladding

Radial power distribution shaping within a PWR fuel assembly utilizing asymmetrically loaded gadolinia-bearing fuel pins

International Nuclear Information System (INIS)

Stone, I.Z.

1992-01-01

As in-core fuel management designs evolve to meet the demands of increasing energy output, more innovative methods are developed to maintain power peaking within acceptable thermal margin limits. In-core fuel management staff must utilize various loading pattern strategies such as cross-core movement of fuel assemblies, multibatch enrichment schemes, and burnable absorbers as the primary means of controlling the radial power distribution. The utilization of fresh asymmetrically loaded gadolinia-bearing assemblies as a fuel management tool provides an additional means of controlling the radial power distribution. At Siemens Nuclear Power Corporation (SNP), fresh fuel assemblies fabricated with asymmetrically loaded gadolinia-bearing fuel rods have been used successfully for several cycles of reactor operation. Asymmetric assemblies are neutronically modeled using the same tools and models that SNP uses to model symmetrically loaded gadolinia-bearing fuel assemblies. The CASMO-2E code is used to produce the homogenized macroscopic assembly cross sections for the nodal core simulator. Optimum fuel pin locations within the asymmetrical assembly are determined using the pin-by-pin PDQ7 assembly core model for each new assembly design. The optimum pin location is determined by the rod loading that minimizes the peak-to-average pin power

Fuel pin transfer tool

International Nuclear Information System (INIS)

Patenaude, R. S.

1985-01-01

A fuel pin transfer tool has a latching device of the collet type attached to a first member movable vertically through a long work stroke enabling a fuel pin in an under water assembly to be engaged and withdrawn therefrom or placed therein and released. The latching device has a collet provided with a plurality of resilient fingers having cam portions normally spaced apart to receive the upper end of a fuel pin between them and a second member, movable vertically through a short stroke relative to the first member is provided with cam portions engageable with those of the fingers and is yieldably and resiliently held in a raised position in which its cam portions engage those of the fingers and force the fingers into their pin-gripping positions. When a predetermined force is applied to the second member, it is so moved that its cam portions are disengaged from the cam portions of the fingers permitting the latter to move into their normal relationship in which a gripped pin is released or another pin received but with their pin-gripping relationship positively re-established and maintained once the force on the tubular member is lessened. Movement of the first member in either direction and movement of the second member into its raised position is attended by forces inadequate to affect the integrity of fuel pin cladding. That force is applied in the preferred embodiment, by a power operated actuator which is within the upper portion of a housing and, in the preferred embodiment, carried by the long stroke member but always in the upper housing portion which is of a material sufficiently translucent to enable the actuator to be observed throughout the work stroke and is sufficiently light in weight to prevent the tool from being top heavy

Activities at the Institute of Materials and Solid State Research of the Karlsruhe Nuclear Research Centre in the field of fuel pin modelling

International Nuclear Information System (INIS)

Elbel, H.

1979-01-01

Fuel pin modelling has been pursued at the Institute of Materials and Solid State Research (IMF) of the Karlsruhe Nuclear Research Centre (KfK) with the main objective to provide a detailed quantitative analysis of the fuel pin behaviour in a LMFBR under normal and off-normal operation conditions. The computer programs and models developed at the IMF serve the purpose to aid effectively in the development of an optimized fuel pin concept for a LMFBR. What extent of clad deformation can be tolerated without running into clad failure? What is the influence of neutron dose, temperature, corrosion attack, arid cyclic forces on the state of the clad? What may be the reasons for clad failure? In answering these questions computer programs can play an important role. The activities at the IMF in the field of fuel pin modelling cover the following topics: development of computer programs and models; validation of these programs and models, application to the design of fuel pins for irradiation experiments; assistance in the evaluation of operation data and post- irradiation results, and parametric studies on the influence of design parameters, operation conditions and certain material phenomena on the in-pile behaviour of the fuel pin

Method of dismantling nuclear fuel elements

International Nuclear Information System (INIS)

Adams, G.J.

1983-01-01

Nuclear fuel assemblies of the kind comprising fuel pins in dimpled cellular grids are freed from the grids to aid dismantling of the assemblies by causing a rotary sleeve to pass concentrically over the pins to remove the dimples in the grids and thereby increase the freedom of the pins in the cells of the grids. (author)

Preliminary evaluation of pin power distribution for fuel assemblies of SMART by MCNP

International Nuclear Information System (INIS)

Kim, Kyo Youn

1998-08-01

Monte Carlo transport code MCNP can describe an object sophisticately by use of three-dimensional modelling and can adopt a continuous energy cross-section library. Therefore MCNP has been widely utilized in the field of radiation physics to estimate fluxes and dose rates for nuclear facilities and to review results from conventional methods such a as discrete ordinates method and point kernel method. The Monte Carlo method has recently been introduced to estimated the neutron multiplication factor and pin power distribution in the fuel assembly of a reactor core. The operating thermal power of SMART core is 330 MWt and there are 57 fuel assemblies in the core. In this study it was assumed that the core has 4 types of fuel assemblies. In this study, MCNP4a was used to perform to estimate criticality and normalized pin power distribution in a fuel assembly of SMART core. The results from MCNP4a calculations are able to be used review those from nuclear design/analysis code. It is very complicated to pick up interested data from MCNP output list and to normalize pin power distribution in a fuel assembly because MCNP is not only a nuclear design/analysis code. In this study a program FAPIN was developed to generated a generate a normalized pin power distribution from the MCNP output list. (author). 11 refs

Cesium migration in LMFBR fuel pins

International Nuclear Information System (INIS)

Karnesky, R.A.; Jost, J.W.; Stone, I.Z.

1978-10-01

The factors affecting the axial migration of cesium in mixed oxide fuel pins and the effects of cesium migration on fuel pin performance are examined. The development and application of a correlated model which will predict the occurrence of cesium migration in a mixed oxide (75 w/o UO 2 + 25 w/o PuO 2 ) fuel pins over a wide range of fabrication and irradiation conditions are described

Critical experiments supporting underwater storage of tightly packed configurations of spent fuel pins. Technical progress report, January 1-March 31, 1981

International Nuclear Information System (INIS)

Hoovler, G.S.; Baldwin, M.N.

1981-04-01

Critical experiments are in progress on arrays of 2 1/2% enriched UO 2 fuel pins simulating underwater pin storage of spent power reactor fuel. Pin storage refers to a spent fuel storage concept in which the fuel assemblies are dismantled and the fuel pins are tightly packed into specially designed canisters. These experiments are providing benchmark data with which to validate nuclear codes used to design spent fuel pin storage racks

Fuel-pin cladding transient failure strain criterion

International Nuclear Information System (INIS)

Bard, F.E.; Duncan, D.R.; Hunter, C.W.

1983-01-01

A criterion for cladding failure based on accumulated strain was developed for mixed uranium-plutonium oxide fuel pins and used to interpret the calculated strain results from failed transient fuel pin experiments conducted in the Transient Reactor Test (TREAT) facility. The new STRAIN criterion replaced a stress-based criterion that depends on the DORN parameter and that incorrectly predicted fuel pin failure for transient tested fuel pins. This paper describes the STRAIN criterion and compares its prediction with those of the stress-based criterion

Reactor physics analysis of the pin-cell Doppler effect in a thermal nuclear reactor

International Nuclear Information System (INIS)

Kruijf, W.J.M. de.

1995-01-01

This report has also been published as a PhD thesis. It deals with the Doppler effect in thermal nuclear reactors. Especially the behaviour of the reactor in transient conditions is an important issue. During such a transient the radial temperature profile in a fuel pin changes. In this PhD research effective fuel temperatures have been calculated for arbitrary temperature profiles in the fuel pin with the improved slowing-down code ROLAIDS-CPM. A general expression for the effective fuel temperature in a specific fuel pin is found by defining this effective fuel temperature as a weighted sum of the temperatures in different radial fuel zones. Also, the radial power profile in a fuel pin has been calculated by performing detailed burnup calculations, which agree very well with experimental data. (orig.)

Nuclear fuel assembly

International Nuclear Information System (INIS)

Delafosse, Jacques.

1977-01-01

This invention relates to a nuclear fuel assembly for a light or heavy water reactor, or for a fast reactor of the kind with a bundle of cladded pins, maintained parallel to each other in a regular network by an assembly of separate supporting grids, fitted with elastic bearing surfaces on these pins [fr

FFTF fuel pin design bases and performance

International Nuclear Information System (INIS)

Cox, C.M.; Hanson, J.E.; Roake, W.E.; Slember, R.J.; Weber, C.E.; Millunzi, A.C.

1975-04-01

The FFTF fuel pin was conservatively designed to meet thermal and structural performance requirements in the categories normal operation, upset events, emergency events, and hypothetical, faulted events. The fuel pin operating limits consistent with these requirements were developed from a strong fuel pin irradiation testing program scoped to define the performance capability under relevant steady state and transient conditions. Comparison of the results of the irradiation testing program with design requirements indicates that the FFTF fuel pin can exceed its goal burnup of 80,000 MWd/MTM. (U.S.)

Mode of failure of LMFBR fuel pins

International Nuclear Information System (INIS)

Washburn, D.F.

1975-01-01

The objectives of the irradiation test described were to evaluate mixed-oxide fuel performance and to confirm the design adequacy of the FFTF fuel pins. After attainment of the initial objectives the irradiation of several of the original fuel pins was continued until a cladding breach occurred. The consequences of a cladding breach were evaluated by reconstituting the original 37-pin subassembly into two 19-pin subassemblies after a burnup at 50,000 MWd/MTM (5.2 a/o). The original pins were supplemented with fresh pins as necessary. Irradiation of the subassemblies was continued until a cladding breach occurred. Results are presented and discussed

Assessment of pin-by-pin fission rate distribution within MOX/UO{sub 2} fuel assembly using MCNPX code

Energy Technology Data Exchange (ETDEWEB)

Louis, Heba Kareem; Amin, Esmat [Nuclear and Radiological Regulation Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

2016-03-15

The aim of the present paper is to assess the calculations of pin-by-pin group integrated fission rates within MOX/UO{sub 2} Fuel assemblies using the Monte Carlo code MCNP2.7c with two sets of the available latest nuclear data libraries used for calculating MOX-fueled systems. The data that are used in this paper are based on the benchmark by the NEA Nuclear Science Committee (NSC). The k{sub ∞} and absorption/fission reaction rates per isotope, k{sub eff} and pin-by-pin group integrated fission rates on 1/8 fraction of the geometry are determined. To assess the overall pin-by-pin fission rate distribution, the collective per cent error measures were investigated. The results of AVG, MRE and RMS error measures were less than 1 % error. The present results are compared with other participants using other Monte Carlo codes and with CEA results that were taken in the benchmark as reference. The results with ENDF/B-VI.6 are close to the results received by MVP (JENDL3.2) and SCALE 4.2 (JEF2.2). The results with ENDF/BVII.1 give higher values of k{sub ∞} reflecting the changes in the newer evaluations. In almost all results presented here, the MCNP calculated results with ENDF/B VII.1 should be considered more than those obtained by using other Monte Carlo codes and nuclear data libraries. The present calculations may be consider a reference for evaluating the numerical schemes in production code systems, as well as the global performance including cross-section data reduction methods as the calculations used continuous energy and no geometrical approximations.

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)

MONJU fuel pin performance analysis

International Nuclear Information System (INIS)

Kitagawa, H.; Yamanaka, T.; Hayashi, H.

1979-01-01

Monju fuel pin has almost the same properties as other LMFBR fuel pins, i.e. Phenix, PFR, CRBR, but would be irradiated under severe conditions: maximum linear heat rate of 381 watt/cm, hot spot cladding temperature of 675 deg C, peak burnup of 131,000 MWd/t, peak fluence (E greater than 0.1 MeV) of 2.3 10 23 n/cm 2 . In order to understand in-core performance of Monju fuel pin, its thermal and mechanical behaviour was predicted using the fast running performance code SIMPLE. The code takes into account pellet-cladding interaction due to thermal expansion and swelling, gap conductance, structural changes of fuel pellets, fission product gas release with burnup and temperature increase, swelling and creep of fuel pellets, corrosion of cladding due to sodium flow and chemical attack by fission products, and cumulative damage of the cladding due to thermal creep

Analytical method for reconstruction pin to pin of the nuclear power density distribution

International Nuclear Information System (INIS)

Pessoa, Paulo O.; Silva, Fernando C.; Martinez, Aquilino S.

2013-01-01

An accurate and efficient method for reconstructing pin to pin of the nuclear power density distribution, involving the analytical solution of the diffusion equation for two-dimensional neutron energy groups in homogeneous nodes, is presented. The boundary conditions used for analytic as solution are the four currents or fluxes on the surface of the node, which are obtained by Nodal Expansion Method (known as NEM) and four fluxes at the vertices of a node calculated using the finite difference method. The analytical solution found is the homogeneous distribution of neutron flux. Detailed distributions pin to pin inside a fuel assembly are estimated by the product of homogeneous flux distribution by local heterogeneous form function. Furthermore, the form functions of flux and power are used. The results obtained with this method have a good accuracy when compared with reference values. (author)

Analytical method for reconstruction pin to pin of the nuclear power density distribution

Energy Technology Data Exchange (ETDEWEB)

Pessoa, Paulo O.; Silva, Fernando C.; Martinez, Aquilino S., E-mail: ppessoa@con.ufrj.br, E-mail: fernando@con.ufrj.br, E-mail: aquilino@imp.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

2013-07-01

An accurate and efficient method for reconstructing pin to pin of the nuclear power density distribution, involving the analytical solution of the diffusion equation for two-dimensional neutron energy groups in homogeneous nodes, is presented. The boundary conditions used for analytic as solution are the four currents or fluxes on the surface of the node, which are obtained by Nodal Expansion Method (known as NEM) and four fluxes at the vertices of a node calculated using the finite difference method. The analytical solution found is the homogeneous distribution of neutron flux. Detailed distributions pin to pin inside a fuel assembly are estimated by the product of homogeneous flux distribution by local heterogeneous form function. Furthermore, the form functions of flux and power are used. The results obtained with this method have a good accuracy when compared with reference values. (author)

Neutron radiography of fuel pins

International Nuclear Information System (INIS)

Jackson, C.N. Jr.; Powers, H.G.; Burgess, C.A.

1975-01-01

Neutron radiography performed with a reactor source has been shown to be a superior radiographic method for the examination of unirradiated mixed oxide fuel pins at the Hanford Engineering Development Laboratory. Approximately 1,700 fuel pins were contained in a sample that demonstrated the capability of the method for detecting laminations, structural flaws, fissile density variation, hydrogenous inclusions and voids in assembled fuel pins. The nature, extent, and importance of the detected conditions are substantiated by gamma autoradiography and by destructive analysis employing alpha autoradiography, electron microprobe and visual inspection. Also, a series of radiographs illustrate the response of neutron radiography as compared to low voltage and high voltage x-ray and gamma source Iridium 192 radiography. (U.S.)

Electro-optical fuel pin identification system

International Nuclear Information System (INIS)

Kirchner, T.L.

1978-09-01

A prototype Electro-Optical Fuel Pin Identification System referred to as the Fuel Pin Identification System (FPIS) has been developed by the Hanford Engineering Development Laboratory (HEDL) in support of the Fast Flux Test Facility (FFTF) presently under construction at HEDL. The system is designed to remotely read an alpha-numeric identification number that is roll stamped on the top of the fuel pin end cap. The prototype FPIS consists of four major subassemblies: optical read head, digital compression electronics, video display, and line printer

Fuel pin bowing and related investigation of WWER-440 control rod influence on power release inside of neighbouring fuel pins

International Nuclear Information System (INIS)

Mikus, J.

2005-01-01

The purpose of this work consists in investigation of the WWER-440 control rod (CR) influence on space power distribution, especially from viewpoint of the values and gradient occurrence that could result in static and cyclic loads with some consequences, e.g. fuel pin bowing. As known, CR can cause power peaks in periphery fuel pins of adjacent operating assemblies because of the butt joint design of the absorbing adapter to the CR fuel part, that is, presence of the water cavity resulting in a flash up of thermal neutrons. As a consequence, beside well-known peaks in axial power distribution, above power gradients can occur inside of mentioned fuel pins. Because of complicated geometry and material composition of the CR, the detailed calculations concerning both above phenomena are complicated, too. Therefore it is useful to acquire appropriate experimental data to investigate mentioned influence and compare them with calculations. Since detailed power distributions cannot be obtained in the NPP, needed information is provided by means of experiments on research reactors. In case of measurements inside of fuel pins, special (e.g. track) detectors placed between fuel pellets are used. Such works are relatively complicated and time consuming, therefore an evaluation based on mathematical modelling and numerical approximation was proposed by means of that, and using measured power release in some selected fuel pins, information about power release inside of one of these fuel pins, can be obtained. For this purpose, an experiment on light water, zero-power research reactor LR-0 was realized and axial power distribution measurements were performed in a WWER-440 type core near to an authentic CR model. Application of the above evaluation method is demonstrated on one ''investigated'' fuel pin neighbouring CR by means of following results: 1. Axial power distribution inside of investigated fuel pin in two opposite positions on its pellets surface that are situated to

Fabrication drawings of fuel pins for FUJI project among PSI, JNC and NRG. Revised version

International Nuclear Information System (INIS)

Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Nagayama, Masahiro

2002-02-01

Irradiation tests and post-irradiation examinations in the framework of JNC-PSI-NRG collaboration project will be performed in 2003-2005. Irradiation fuel pins will be fabricated by the middle of 2003. The fabrication procedure for irradiation fuel pins has been started in 2001. Several fabrication tests and qualification tests in JNC and PSI (Paul Scherrer Institute, Switzerland) have been performed before the fuel pin fabrication. According to the design assignment between PSI and JNC in the frame of this project, PSI should make a specification document for the fuel pellet, the sphere-pac fuel particles, the vipac fuel particles, and the fuel pin. JNC should make a fabrication drawing for irradiation pins. JNC has been performed the fuel design in cooperation with PSI and NRG (Nuclear Research and Consultancy Group, Netherlands). In this project, the pelletized fuel, the sphere-pac fuel, and the vipac fuel will be simultaneously irradiated on HFR (High Flux Reactor, Netherlands). This fabrication drawing has been made under the design assignment with PSI, and consists of the drawing of MOX pellet, thermal insulator pellet, pin components, fuel segments, and the constructed pin. The fabrication drawings were approved in October 2001, but after that, the optimization of specifications has been discussed and agreed among all partners. In this report, the revised fabrication drawings will be shown. Based on the commission of Plutonium Fuel Technology Group, Advanced Fuel Recycle Technology Division, this design work has been performed in Fuel Design and Evaluation Group, Plutonium Fuel Fabrication Division, Plutonium Fuel Center. (author)

Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5.5 at. % burnup

International Nuclear Information System (INIS)

Strain, R.V.; Johnson, C.E.

1978-05-01

Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760 0 C. The maximum diametral change that occurred during irradiation was 0.2% ΔD/D 0 . The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred

Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5. 5 at. % burnup

Energy Technology Data Exchange (ETDEWEB)

Strain, R V; Johnson, C E

1978-05-01

Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760/sup 0/C. The maximum diametral change that occurred during irradiation was 0.2% ..delta..D/D/sub 0/. The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred.

Nuclear fuel assembly

International Nuclear Information System (INIS)

Anthony, A.J.

1980-01-01

A bimetallic spacer means is cooperatively associated with a nuclear fuel assembly and operative to resist the occurrence of in-reactor bowing of the nuclear fuel assembly. The bimetallic spacer means in one embodiment of the invention includes a space grid formed, at least principally, of zircaloy to the external surface of which are attached a plurality of stainless steel strips. In another embodiment the strips are attached to fuel pins. In each of the embodiments, the stainless steel strips during power production expand outwardly to a greater extent than do the members to which the stainless steel strips are attached, thereby forming stiff springs which abut against like bimetallic spacer means with which the other nuclear fuel assemblies are provided in a given nuclear reactor core to thus prevent the occurrence of in-reactor bowing of the nuclear fuel assemblies. (author)

Heat transfer in a fuel pin shipping container

International Nuclear Information System (INIS)

Ingham, J.G.

1980-01-01

Maximum cladding temperatures occur when the IDENT 1578 fuel pin shipping container is installed in the T-3 Cask. The maximum allowable cladding temperature of 800 0 F is reached when the rate of energy deposited in the 19-pin basket reaches 400 watts. Since 45% of the energy which is generated in the fuel escapes the 19-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 400/.55 = 727 watts. Similarly, the maximum allowable cladding temperature of 800 0 F is reached when the rate of energy deposited in the 40-pin basket reaches 465 watts. Since 33% of the energy which is generated in the fuel escapes the 40-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 465/.66 = 704 watts. The IDENT 1578 fuel pin shipping container therefore meets its thermal design criteria. IDENT 1578 can handle fuel pins with a decay heat load of 600 watts while maintaining the maximum fuel pin cladding temperature below 800 0 F. The emissivities which were determined from the test results for the basket tubes and container are relatively low and correspond to new, shiny conditions. As the IDENT 1578 container is exposed to high temperatures for extended periods of time during the transportation of fuel pins, the emissivities will probably increase. This will result in reduced temperatures

Integral Fast Reactor fuel pin processor

International Nuclear Information System (INIS)

Levinskas, D.

1993-01-01

This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves

Correlation of creep and swelling with fuel pin performance

International Nuclear Information System (INIS)

Jackson, R.J.; Washburn, D.F.; Garner, F.A.; Gilbert, E.R.

1975-09-01

The HEDL PNL-11 experiment described was one in a series of fueled subassemblies irradiated in EBR-II to demonstrate the adequacy of the FFTF fuel pin design. The cladding material, dimensions, and fuel density are prototypic of FFTF. Because neutron flux in EBR-II is lower than in FFTF, the uranium enrichment is higher in these experimental fuel pins, irradiated in EBR-II, than the FFTF enrichment for comparable linear heat rates. Some pertinent oprating conditions for the center fuel pin in this experiment are listed. This 37-pin subassembly represents, at 110,000 MWd/MTM, the highest burnup yet attained by a prototypic FFTF subassembly. Similarly, this is the highest fluence presently attained by prototypic fuel pins. A cladding breach occurred in one fuel pin which is presently being examined. Results are presented and discussed

Limits on the experimental simulation of nuclear fuel rod response

International Nuclear Information System (INIS)

Hagar, R.C.

1980-01-01

The steady-state and transient effects of intrinxic geometric and material property differences between typical nuclear fuel pins and electric fuel pin simulators (FPSs) are identified. The effectiveness of varying the transient power supplied to the FPS in reducing the differences between the transient responses of nuclear fuel pins and FPSs is investigated. This effectiveness is shown to be limited by the heat capacity of the FPS, the allowed range of the power program, and different FPS power requirements at different positions on a full-length FPS

RANS based CFD methodology for a real scale 217-pin wire-wrapped fuel assembly of KAERI PGSFR

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jae-Ho, E-mail: jhjeong@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseoung-gu, Daejeon (Korea, Republic of); Song, Min-Seop [Department of Nuclear Engineering, Seoul National University, 559 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of); Lee, Kwi-Lim [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseoung-gu, Daejeon (Korea, Republic of)

2017-03-15

Highlights: • This paper presents a suitable way for a practical RANS based CFD methodology which is applicable to real scale 217-pin wire-wrapped fuel assembly of KAERI PGSFR. • A key point of differentiation of the RANS based CFD methodology in this study is adapting an innovative grid generation method using a fortran based in-house code with a GGI function in a general-purpose commercial CFD code, CFX. • The RANS based CFD methodology is implemented with high resolution scheme and SST turbulence model in the 7-pin 37-pin, and 127-pin wire-wrapped fuel assembly of PNC and JNC. Furthermore, the RANS based CFD methodology can be successfully extended to the real scale 217-pin wire-wrapped fuel bundles of KAERI PGSFR. • Three-dimensional thermal-hydraulic characteristics have been also investigated briefly. - Abstract: This paper presents a suitable way for a practical RANS (Reynolds Averaged Navier-Stokes simulation) based CFD (Computational Fluid Dynamics) methodology which is applicable to real scale 217-pin wire-wrapped fuel assembly of KAERI (Korea Atomic Energy Research Institute) PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor). The main purpose of the current study is to support license issue for the KAERI PGSFR core safety and to elucidate thermal-hydraulic characteristics in a 217-pin wire-wrapped fuel assembly of KAERI PGSFR. A key point of differentiation of the RANS based CFD methodology in this study is adapting an innovative grid generation method using a fortran based in-house code with a GGI (General Grid Interface) function in a general-purpose commercial CFD code, CFX. The innovative grid generation method with GGI function can achieve to simulate a real wire shape with minimizing cell skewness. The RANS based CFD methodology is implemented with high resolution scheme in convection term and SST (Shear Stress Transport) turbulence model in the 7-pin 37-pin, and 127-pin wire-wrapped fuel assembly of PNC (Power reactor and Nuclear fuel

Fabrication of FFTF fuel pin wire wrap

International Nuclear Information System (INIS)

Epperson, E.M.

1980-06-01

Lateral spacing between FFTF fuel pins is required to provide a passageway for the sodium coolant to flow over each pin to remove heat generated by the fission process. This spacing is provided by wrapping each fuel pin with type 316 stainless steel wire. This wire has a 1.435mm (0.0565 in.) to 1.448mm (0.0570 in.) diameter, contains 17 +- 2% cold work and was fabricated and tested to exacting RDT Standards. About 500 kg (1100 lbs) or 39 Km (24 miles) of fuel pin wrap wire is used in each core loading. Fabrication procedures and quality assurance tests are described

Irradiation project of SiC/SiC fuel pin 'INSPIRE': Status and future plan

International Nuclear Information System (INIS)

Kohyama, Akira; Kishimoto, Hirotatsu

2015-01-01

After the March 11 Disaster in East-Japan, Research and Development towards Ensuring Nuclear Safety Enhancement for LWR becomes a top priority R and D in nuclear energy policy of Japan. The role of high temperature non-metallic materials, such as SiC/SiC, is becoming important for the advanced nuclear reactor systems. SiC fibre reinforced SiC composite has been recognised to be the most attractive option for the future, now, METI fund based project, INSPIRE, has been launched as 5-year termed project at OASIS in Muroran Institute of Technology aiming at early realisation of this system. INSPIRE is the irradiation project of SiC/SiC fuel pins aiming to accumulate material, thermal, irradiation effect data of NITE-SiC/SiC in BWR environment. Nuclear fuel inserted SiC/SiC fuel pins are planned to be installed in the Halden reactor. The project includes preparing the NITE-SiC/SiC tubes, joining of end caps, preparation of rigs to control the irradiation environment to BWR condition and the instruments to measure the condition of rigs and pins in operation. Also, basic neutron irradiation data will be accumulated by SiC/SiC coupon samples currently under irradiation in BR2. The output from this project may present the potentiality of NITE-SiC/SiC fuel cladding with the first stage fuel-cladding interaction. (authors)

FFTF/IEM cell fuel pin weighing system

International Nuclear Information System (INIS)

Gibbons, P.W.

1987-01-01

The Interim Examination and Maintenance (IEM) cell in the Fast Flux Test Facility (FFTF) is used for remote disassembly of irradiated fuel and materials experiments. For those fuel experiments where the FFTF tag-gas detection system has indicated a fuel pin cladding breach, a weighing system is used in identifying that fuel pin with a reduced weight due to the escape of gaseous and volatile fission products. A fuel pin weighing machine, originally purchased for use in the Fuels and Materials Examination Facility (FMEF), was the basis for the IEM cell system. Design modifications to the original equipment were centered around adapting the machine to the differences between the two facilities and correcting deficiencies discovered during functional testing in the IEM cell mock-up

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

Analysis of fuel cladding chemical interaction in mixed oxide fuel pins

International Nuclear Information System (INIS)

Weber, J.W.; Dutt, D.S.

1976-01-01

An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves

Fuel Coolant Interaction Results in the Fuel Pins Melting Facility (PMF)

International Nuclear Information System (INIS)

Urunashi, H.; Hirabayashi, T.; Mizuta, H.

1976-01-01

The experimental work related to FCI at PNC has been concentrated into the molten UO 2 dropping test. After the completion of molten UO 2 drop experiments, emphasis is directed toward the FCI phenomena of the initiating conditions of the accident under the more realistic geometry. The experiments are conducted within the Pin Melt Facility (PMF) in which UO 2 pellets clad in stainless steel are melted by direct electric heating under the stagnant or flowing sodium. The primary objectives of the PMF test are to: - obtain detail experimental results (heat-input, clad temperature, sodium temperature, etc.) on the FCI under TOP and LOF conditions; - observe the movement of the fuel before and after the pin failure by the X-ray cinematography; - observe the degree of coherence of the pin failures; - accumulate the experience of the FCI experiment which is applicable to the subassembly or more larger scale; - simulate the fuel behavior of the in-pile test (GETR, CABRI). The preliminary conclusions can be drawn from the foregoing observations are as follows: - Although the fuel motion and FCI of the closed test section appeared to be different from those of the open test section, the conclusion of the effect of the inside pressure on FCI needs more experimental data. - The best heating condition of the UO 2 pellet for the FCI study with PMF is established as 40 w/cm at the steady state and 1680 J/g of UO 2 during the additional transient state. The total energy deposition of the UO 2 pellet is thus estimated in the range of 2400 J/g of UO 2 -2600 J/g of UO 2 . The analytical model of the fuel pin failure and the subsequent FCI are suggested to count the following parameters: - The fuel pin failure due to the fuel vaporization due to the rapid energy deposition; - Molten fuel, clad and sodium interaction in the fuel pin after the pin failure; - The upward flow of molten fuel with molten clad or vapor sodium, as well as the slumping of molten fuel

A simple nondestructive technique for monitoring the bond gas in sealed fast reactor nuclear fuel pins

Energy Technology Data Exchange (ETDEWEB)

Shriwastwa, B B; Mehrotra, R S; Ghosh, J K [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

1994-12-31

A simple nondestructive testing technique has been developed to identify bond gas inside a welded fuel pin. The technique is based on the accurate surface temperature measurement of fuel pins heated in a constant temperature water bath. This technique can be applied in Fast Breeder Test Reactor (FBTR) fuel pin production line due to simplicity of the set up, simple operation and quick response time. An attempt was made to develop a non destructive test method for monitoring the bond gas composition. Preliminary development work carried out in this connection, the test method adopted and the test results are presented. 1 ref., 5 figs., 1 tab.

Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.; Asaga, T.; Shikakura, S.

1991-02-01

The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectives of the designated TOPI-1A through -1D tests were to establish the cladding breaching threshold and mechanisms, and investigate the thermal and mechanical effects of the transient on pin behavior. The tests were conducted in EBR-2, a normally steady-state reactor. The modes of transient operation in EBR-2 were described in a previous paper. Two ramp rates, 0.1%/s and 10%/s, were selected to provide a comparison of ramp-rate effects on fuel behavior. The test pins chosen for the series covered a range of design and pre-test irradiation parameters. In the first test (1A), all pins maintained their cladding integrity during the 0.1%/s ramp to 60% peak overpower. Fuel pins with aggressive designs, i.e., high fuel- smear density and/or thin cladding, were, therefore, included in the follow-up 1B and 1C tests to enhance the likelihood of achieving cladding breaching. In the meantime, a higher pin overpower capability, to greater than 100%, was established by increasing the reactor power limit from 62.5 to 75 MWt. In this paper, the significant results of the 1B and 1C tests are presented. 4 refs., 5 figs., 1 tab

Post irradiation examination on test fuel pins for PWR

International Nuclear Information System (INIS)

Fogaca Filho, N.; Ambrozio Filho, F.

1981-01-01

Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

Reference Neutron Radiographs of Nuclear Reactor Fuel

DEFF Research Database (Denmark)

Domanus, Joseph Czeslaw

1986-01-01

Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

The interpretation of fuel centre temperature measurements on a suspected leaking fuel pin

International Nuclear Information System (INIS)

Ainscough, J.B.; Lang, C.; Clough, D.J.

1983-01-01

In order to study fuel densification a series of single instrumented pin irradiations has been carried out in the High Pressure Water Loop of DIDO at Harwell. The behaviour of two of these pins was different from that expected. In the fifth test, where the fuel was 95% dense pellet UO 2 and expected to densify readily in-reactor, the fuel centre temperature increased from its starting value of approx. 1300 deg. C at a rate somewhat higher than expected on the basis of predicted densification rates. After about six days, the temperature increased rapidly and unexpectedly to 2100-2200 deg. C and remained steady at this level for a further eight days until a reactor trip occurred and the pin was unloaded. Predictions made using the HOTROD code imply a maximum fuel temperature of less than 1500 deg. C after densification. Post-irradiation examination confirmed that fission gas release had occurred, that the measured temperatures were consistent with the fuel microstructure and that the pin had a high internal gas pressure. The fourth pin in the series contained 97% dense UO 2 which was also expected to be dimensionally unstable. Qualitatively its behaviour was similar to that of the fifth pin though the temperatures throughout were lower. This pin experienced a number of major power cycles and failed after about 30 days in-reactor. It is probable that coolant ingress occurred in both pins via the thermocouple Hoke seal, degrading the filling gas conductivity and allowing the fuel to densify rapidly with consequent increase in the fuel/clad gap and hence in fuel temperature. These irradiations show that, for a short time at least, an apparently unfailed pin could operate undetected with temperatures significantly higher than those predicted for normal operation. (author)

Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

International Nuclear Information System (INIS)

Roake, W.E.

1977-01-01

Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals

Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

Energy Technology Data Exchange (ETDEWEB)

Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States)

1977-04-01

Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals.

Post-irradiation examination of a fuel pin using a microscopic X-ray system: Measurement of carbon deposition and pin metrology

International Nuclear Information System (INIS)

Gras, Ch.; Stanley, S.J.

2008-01-01

The paper presents some interesting aspects associated with X-ray imaging and its potential application in the nuclear industry. The feasibility of using X-ray technology for the post-irradiation examination of a fuel pin has been explored, more specifically pin metrology and carbon deposition measurement. The non-active sample was specially designed to mimic the structure of an AGR fuel pin whilst a carbon based material was applied to the mock up fuel rod in order to mimic carbon deposition. Short duration low energy (50 kV) 2D digital radiography was employed and provided encouraging results (with respect to carbon deposition thickness and structure measurements) for the mock up fuel pin with a spatial resolution of around 10 μm. Obtaining quantitative data from the resultant images is the principal added value associated with X-ray imaging. A higher intensity X-ray beam (≥90 kV) was also used in conjunction with the low energy set-up to produce a clear picture of the cladding as well as the interface between the lead (Pb mimics the uranium oxide) and stainless steel cladding. Spent fuel metrology and routine radiography are two additional tasks that X-ray imaging could perform for the post-irradiation examination programme. Therefore, when compared to other techniques developed to deliver information on one particular parameter, X-ray imaging offers the possibility to extract useful information on a range of parameters

Calculations on the effect of pellet filling on the rewetting of overheated nuclear reactor fuel pins

International Nuclear Information System (INIS)

Pearson, K.G.; Loveless, J.

1977-03-01

Numerical solutions of the rewetting equations are presented which show the effect of filler material and gas gap on the rate of rewetting of an overheated fuel pin. It is shown that taking the presence of the fuel into account can lead to a large reduction in the calculated rewetting speed compared with a calculation which neglects the presence of fuel. The effect is most marked in conditions where rewetting speeds tend to be already low, such as at high pin temperatures and low ambient pressure. A comparison is made between the predictions of the present method and experimental data obtained on zircaloy and stainless steel pins filled with magnesia and with boron nitride. In all cases filling the pins produced a large reduction in rewetting speed and the agreement between the calculated and measured effect was encouraging. It is concluded that the presence of the UO 2 pellet filling should be taken into account when calculating rewetting speeds in safety assessments. (author)

Progress in fuel pin modelling in the USA

Energy Technology Data Exchange (ETDEWEB)

Stephen, J D; Biancheria, A; Leibnitz, D; O' Reilly, B D; Liu, Y Y; Labar, M P; Gneiting, B C [General Electric Company, Sunnyvale, CA (United States)

1979-12-01

In the USA, the focus for theoretical fuel pin modeling is the LIFE system. This system of codes, algorithms, criteria and analysis guidelines is intended to provide a common basis for communication amongst the development groups, a reference set of analysis guidelines for design, and eventually a consensus on the state-of-the-art for licensing. The technical objective is to predict the effect of design options on fuel pin performance limits, which include fuel temperature, pin deformation and cladding breach during normal operation and design basis transients. The mechanistic approach to modeling is taken in LIFE to the extent possible. That is, the approach is to describe the key phenomena in sufficient detail to provide a fundamental understanding of their synergistic effect on the fuel pin performance limits.

SP-100 Fuel Pin Performance: Results from Irradiation Testing

Science.gov (United States)

Makenas, Bruce J.; Paxton, Dean M.; Vaidyanathan, Swaminathan; Marietta, Martin; Hoth, Carl W.

1994-07-01

A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pins are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

Correlations between fuel pins irradiated in fast and thermal fluxes using the frump fuel pin modelling program

International Nuclear Information System (INIS)

Hayns, M.R.; Adam, J.

1975-08-01

There is no experimental facilities in which a fuel pin can be irradiated in a fast environment under well defined conditions of over power or flow run down. Consequently most of the infor mation which is being accumulated on the behaviour of fuel pins under severe conditions is obtained from either capsule or loop rigs in thermal reactors. It is the purpose of this paper to highlight the differences between the behaviour of fuel pins irradiated in a thermal flux and a fast flux. A typical set of conditions is taken from an overpower experiment in a thermal flux and the behaviour of the system is analysed using the fuel modelling program FRUMP. A second numerical experiment is then performed in which the same conditions prevail, except that a fast flux is assumed, the criterion for comparison being that the total power input to the system is the same in both cases. From the many possible correlations which result from such an exercise the fuel tempreature has been selected to highlight various important features of the two irradiations. It is demonstrated that the flux depression can cause differences in the pin behaviour, even to altering the order of events in a transient. For example fuel melting will occur at different times and at different positions in the fuel in the two cases. It is concluded that the techniques of fuel modelling, as typified in the program FRUMP can provide a very useful tool indeed for the analysis of such experiments and for guiding the establishment of the appropriate correlations for the extrapolation to the fast flux case. (author)

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

International Nuclear Information System (INIS)

Mella, R.; Wenman, M.R.

2013-01-01

Thermo-mechanical contributions to pellet–clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS’s well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used

Accuracy of dimension measurements from neutron radiographs of nuclear fuel pins

International Nuclear Information System (INIS)

Domanus, J. C.

1976-03-01

A review of different methods used for dimension measurements from neutron radiographs. The results are presented of an investigation performed using unirradiated fuel pins with calibrated UO 2 pellet-diameters and fuel-to-clad gaps. A projection microscope, three types of travelling microdensitometers and an electronic image analyzer were used to measure diameters and gaps from neutron radiographs produced at Risoe and Studsvik (Sweden) using different brands of X-ray films and transfer technique with 0.1 mm Dy foil. (author)

Assessment of Radiographic Image Quality by Visual Examination of Neutron Radiographs of the Calibration Fuel Pin

DEFF Research Database (Denmark)

Domanus, Joseph Czeslaw

1986-01-01

Up till now no reliable radiographic image quality standards exist for neutron radiography of nuclear reactor fuel. Under the Euratoro Neutron Radiography Working Group (NRWG) Test Program neutron radiographs were produced at different neutron radiography facilities within the European Community...... of a calibration fuel pin. The radiographs were made by the direct, transfer and tracketch methods using different film recording materials. These neutron radiographs of the calibration fuel pin were used for the assessement of radiographic image quality. This was done by visual examination of the radiographs...

Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

International Nuclear Information System (INIS)

Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

1990-05-01

The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs

Breached fuel pin contamination from Run Beyond Cladding Breach (RBCB) tests in EBR-II

International Nuclear Information System (INIS)

Colburn, R.P.; Strain, R.V.; Lambert, J.D.B.; Ukai, S.; Shibahara, I.

1988-09-01

Studies indicate there may be a large economic incentive to permit some continued reactor operation with breached fuel pin cladding. A major concern for this type of operation is the potential spread of contamination in the primary coolant system and its impact on plant maintenance. A study of the release and transport of contamination from naturally breached mixed oxide Liquid Metal Reactor (LMR) fuel pins was performed as part of the US Department of Energy/Power Reactor and Nuclear Fuel Development Corporation (DOE/PNC) Run Beyond Cladding Breach (RBCB) Program at EBR-II. The measurements were made using the Breached Fuel Test Facility (BFTF) at EBR-II with replaceable deposition samplers located approximately 1.5 meters from the breached fuel test assemblies. The effluent from the test assemblies containing the breached fuel pins was routed up through the samplers and past dedicated instrumentation in the BFTF before mixing with the main coolant flow stream. This paper discusses the first three contamination tests in this program. 2 refs., 5 figs., 2 tabs

Fuel canister and blockage pin fabrication for SLSF Experiment P4

International Nuclear Information System (INIS)

Rhude, H.V.; Folkrod, J.R.; Noland, R.A.; Schaus, P.S.; Benecke, M.W.; Delucchi, T.A.

1983-01-01

As part of its fast breeder reactor safety research program, Argonne National Laboratory (ANL) has conducted an experiment (SLSF Experiment P4) to determine the extent of fuel-failure propagation resulting from the release of molten fuel from one or more heat-generating fuel canisters. The test conditions consisted of 37 full-length FTR fuel pins operating at FTR rated core nominal peak fuel/reduced coolant conditions. Thirty-four of the the fuel pins were prototypical FTR mixed-oxide fuel pins. The other three fuel pins were fabricated with a mid-core section having an enlarged canister containing fully enriched UO 2 . Two of the canisters were cylindrical and one was fluted. The cylindrical canisters were designed to fail and release molten fuel into the 37-pin fuel cluster at near full power

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

International Nuclear Information System (INIS)

Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

1977-01-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

Energy Technology Data Exchange (ETDEWEB)

Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

1977-04-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Development of vibropac MOX fuel pins serviceable up TP superhigh burnups

International Nuclear Information System (INIS)

Mayorshin, A.A.; Gadzhiev, G.I.; Kisly, V.A.; Skiba, O.V.; Tzykanov, V.A.

1998-01-01

The main results on investigations of fast reactor fuel pins with (UPu)O 2 vibropac fuel to substantiate their serviceability up to the super-high burnups are presented. The BOR-60 reactor fuel pins radiation behaviour in stationary, transient and designed emergency conditions has been determined from the fuel pins dimensional stability analysis having regard to the results of investigation fuel and cladding swelling as well as estimations of fuel and cladding thermal-mechanical and physico-chemical interactions. It is shown that the change of the outer diameter is minimum in fuel pins with VMOX fuel with a getter-metallic uranium powder and ferrito-martensite steel cladding, and the corrosion damage of the cladding inner surface is absent up to 26% h.a. The experiments with over-heating of the irradiated fuel pins cladding up to 850 deg. C did not lead to any changes in pins integrity. The availability of the periphery area of the vibropac fuel cure initial structure provides the minimum level of the thermal-mechanical stress at transient conditions of reactor operation. (author)

Report on fabrication of pin components for fuel fabrication in FUJI project (Co-operation in the research and development of advanced sphere-pac fuel among PSI, JNC, and NRG)

International Nuclear Information System (INIS)

Suzuki, Masahiro; Hinai, Hiroshi; Shigetome, Yoshiaki; Kono, Shusaku; Matsuzaki, Masaaki

2003-03-01

Japan Nuclear Cycle Development Institute (JNC) has conducted the co-operation concerning vibro-packed fuels with Paul Scherrer Institut (PSI) in Switzerland and Nuclear Research and consultancy Group (NRG) in the Netherlands. The project 'Research and Development of advanced Sphere-pac Fuel' is called FUJI (FUel irradiations for JNC and PSI) Project. In this project, three types of fuels that are sphere-pac fuels, vipac fuels, and pellet fuels will be irradiated in the High Flux Reactor (HFR) to compare their performance. Based on the drawing which has been agreed among three parties, fabrication of the pin components and welding of the upper and lower connection end plugs were performed in accordance with ISO9001 in JNC. This report describes data of the fabricated pin components, results of welding qualification tests, and quality assurance of the welded components. The fabrication of pin components was successfully completed and they were delivered to PSI in October 2002. (author)

Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

International Nuclear Information System (INIS)

Stefanova, S.; Vitkova, M.; Simeonova, V.; Passage, G.; Manolova, M.; Haralampieva, Z.; Scheglov, A.; Proselkov, V.

1997-01-01

The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab

Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

Energy Technology Data Exchange (ETDEWEB)

Stefanova, S; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Institute for Nuclear Research and Nuclear Energy, Sofia (Bulgaria); Haralampieva, Z [National Electric Company Ltd., Kozloduy (Bulgaria); Scheglov, A; Proselkov, V [Institute of Nuclear Reactors, RSC Kurchatov Inst., Moscow (Russian Federation)

1997-08-01

The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab.

Some aspects of continuum physics used in fuel pin modeling

International Nuclear Information System (INIS)

Bard, F.E.

1975-06-01

The mathematical formulation used in fuel pin modeling is described. Fuel pin modeling is not a simple extension of the experimental and interpretative methods used in classical mechanics. New concepts are needed to describe materials in a reactor environment. Some aspects of continuum physics used to develop these new constitutive equations for fuel pins are presented. (U.S.)

Probabilistic distributions of pin gaps within a wire-spaced fuel subassembly and sensitivities of the related uncertainties to pin gap

International Nuclear Information System (INIS)

Sakai, K.; Hishida, H.

1978-01-01

Probabilistic fuel pin gap distributions within a wire-spaced fuel subassembly and sensitivities of the related uncertainties to fuel pin gaps are discussed. The analyses consist mainly of expressing a local fuel pin gap in terms of sensitivity functions of the related uncertainties and calculating the corresponding probabilistic distribution through taking all the possible combinations of the distribution of uncertainties. The results of illustrative calculations show that with the reliability level of 0.9987, the maximum deviation of the pin gap at the cladding hot spot of a center fuel subassembly is 8.05% from its nominal value and the corresponding probabilistic pin gap distribution is shifted to the narrower side due to the external confinement of a pin bundle with a wrapper tube. (Auth.)

Report of the collaboration project for research and development of sphere-pac fuel among JNC-PSI-NRG (1). Planning, fuel design, pin fabrication

International Nuclear Information System (INIS)

Morihira, Masayuki; Ozawa, Takayuki; Tomita, Yutaka; Suzuki, Masahiro; Kihara, Yoshiyuki; Shigetome, Yoshiaki; Kohno, Shusaku

2004-07-01

The collaboration project concerning sphere-pac fuel among JNC, Swiss PSI (Paul Scherrer Institut) and Dutch NRG (Nuclear Research and Consultancy Group) is in progress. Final target of the project is comparative irradiation tests of sphere-pac fuel in the HFR (High Flux Reactor) in Petten in the Netherlands with pellet type fuel and vipack fuel. Total 16 fuel segments (8 pins) of these three types of fuel are planned to be irradiated. Two sphere-pac fuel segments contain 5%Np in addition to 20%Pu-MOX. Other segments contain no Np. The objective of the irradiation tests is to obtain the restructuring data in the early beginning of life for SPF as well as power-to-melt test data for the potential study of SPF. At the same time introduction of modeling technique for irradiation performance analysis, fuel design, fuel fabrication is also important objective for JNC. Fabrication of irradiation test pins was completed till May 2003 in PSI. After transportation of the fuel pins to Petten, two times of irradiation were performed in January to March in 2004 and now post irradiation tests are in progress. Later two irradiations will be done till the autumn in 2004. This report summarized the basic plan, fuel design, and fabrication of irradiation test pins concerning this collaboration project. (author)

Development of wire wrapping technology for FBR fuel pin

International Nuclear Information System (INIS)

Nogami, Tetsuya; Seki, Nobuo; Sawayama, Takeo; Ishibashi, Takashi

1991-01-01

For the FBR fuel assembly, the spacer wire is adopted to maintain the space between fuel pins. The developments have been carried out to achieve automatically wire wrapping with high precision. Based on the fundamental technology developed through the mock-up test operation, Joyo 'MK-I', fuel pin fabrication was started using partially mechanized wire wrapping machine in 1973. In 1978, an automated wire wrapping machine for Joyo 'MK-II' was developed by the adoption of some improvements for the wire inserting system to end plug hole and the precision of wire pitch. On the bases of these experiences, fully automated wire wrapping machine for 'Monju' fuel pin was installed at Plutonium Fuel Production Facility (PFPF) in 1987. (author)

Fabrication drawings of fuel pins for FUJI project among PSI, JNC and NRG. Revised version 2

International Nuclear Information System (INIS)

Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Nagayama, Masahiro

2002-10-01

Irradiation tests and post-irradiation examinations in the framework of JNC-PSI-NRG collaboration project will be performed in 2003-2005. Irradiation fuel pins will be fabricated by the middle of 2003. The fabrication procedure for irradiation fuel pins has been started in 2001. Several fabrication tests and qualification tests in JNC and PSI (Paul Scherrer Institut, Switzerland) have been performed before the fuel pin fabrication. According to the design assignment between PSI and JNC in the frame of this project, PSI should make specification documents for the fuel pellet, the sphere-pac fuel particles, the vipac fuel fragments, and the fuel segment fabrication. JNC should make the fabrication drawings for irradiation pins. JNC has been performed the fuel design in cooperation with PSI and NRG (Nuclear Research and Consultancy Group, Holland). In this project, the pelletized fuel, the sphere-pac fuel, and the vipac fuel will be simultaneously irradiated on HFR (High Flux Reactor, Holland). The fabrication drawings have been made under the design assignment with PSI, and consist of the drawings of MOX pellet, thermal insulator pellet, pin components, fuel segments, and the constructed pin. The fabrication drawings were approved in October 2001, but after that, the optimization of specifications was discussed and agreed among all partners. According to this agreement, the fabrication drawings were revised in January 2002. After the earlier revision, the shape of particle retainer to be made by PSI was modified from its drawing beforehand delivered. In this report, the fabrication drawings revised again will be shown, and the fabrication procedure (welding Qualification Tests) will be modified in accordance with the result of discussion on the 3rd technical meeting held in September 2002. These design works have been performed in Fuel Design and Evaluation Group, Plutonium Fuel Fabrication Division, Plutonium Fuel Center under the commission of Plutonium Fuel

Alternatives for water basin spent fuel storage using pin storage

International Nuclear Information System (INIS)

Viebrock, J.M.; Carlson, R.W.

1979-09-01

The densest tolerable form for storing spent nuclear fuel is storage of only the fuel rods. This eliminates the space between the fuel rods and frees the hardware to be treated as non-fuel waste. The storage density can be as much as 1.07 MTU/ft 2 when racks are used that just satisfy the criticality and thermal limitations. One of the major advantages of pin storage is that it is compatible with existing racks; however, this reduces the storage density to 0.69 MTU/ft 2 . Even this is a substantial increase over the 0.39 MTU/ft 2 that is achievable with current high capacity stainless steel racks which have been selected as the bases for comparison. Disassembly requires extensive operation on the fuel assembly to remove the upper end fitting and to extract the fuel rods from the assembly skeleton. These operations will be performed with the aid of an elevator to raise the assembly where each fuel rod is grappled. Lowering the elevator will free the fuel rod for transfer to the storage canister. A storage savings of $1510 per MTU can be realized if the pin storage concept is incorporated at a new away-from-reactor facility. The storage cost ranges from $3340 to $7820 per MTU of fuel stored with the lower cost applying to storage at an existing away-from-reactor storage facility and the higher cost applying to at-reactor storage

Fuel pin integrity assessment under large scale transients

International Nuclear Information System (INIS)

Dutta, B.K.

2006-01-01

The integrity of fuel rods under normal, abnormal and accident conditions is an important consideration during fuel design of advanced nuclear reactors. The fuel matrix and the sheath form the first barrier to prevent the release of radioactive materials into the primary coolant. An understanding of the fuel and clad behaviour under different reactor conditions, particularly under the beyond-design-basis accident scenario leading to large scale transients, is always desirable to assess the inherent safety margins in fuel pin design and to plan for the mitigation the consequences of accidents, if any. The severe accident conditions are typically characterized by the energy deposition rates far exceeding the heat removal capability of the reactor coolant system. This may lead to the clad failure due to fission gas pressure at high temperature, large- scale pellet-clad interaction and clad melting. The fuel rod performance is affected by many interdependent complex phenomena involving extremely complex material behaviour. The versatile experimental database available in this area has led to the development of powerful analytical tools to characterize fuel under extreme scenarios

Shield requirement estimation for pin storage room in fuel fabrication plant

International Nuclear Information System (INIS)

Shanthi, M.M.; Keshavamurthy, R.S.; Sivashankaran, G.

2012-01-01

Fast Reactor Fuel Cycle Facility (FRFCF) is an upcoming project in Kalpakkam. It has the facility to recycle the fuel from PFBR. It is an integrated facility, consists of fuel reprocessing plant, fuel fabrication plant (FFP), core subassembly plant, Reprocessed Uranium plant (RUP) and waste management plant. The spent fuel from PFBR would be reprocessed in fuel reprocessing plant. The reprocessed fuel material would be sent to fuel fabrication plant. The main activity of fuel fabrication plant is the production of MOX fuel pins. The fuel fabrication plant has a fuel pin storage room. The shield requirement for the pin storage room has been estimated by Monte Carlo method. (author)

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

Science.gov (United States)

Mcdonald, G. E.

1973-01-01

A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

Fuel pellet relocation behavior in fast reactor uranium-plutonium mixed oxide fuel pin at beginning-of-life

International Nuclear Information System (INIS)

Inoue, Masaki; Ukai, Shigeharu; Asaga, Takeo

1999-08-01

The effects of fabrication parameters, irradiation conditions and fuel microstructural feature on fuel pellet relocation behavior in fast reactor fuel pins were investigated. This work focused only on beginning-of-life conditions, when fuel centerline temperature depends largely on the behavior. Fuel pellet relocation behavior in Joyo Mk-II driver could not be characterized because of the lack of data. And the behavior in FFTF driver and its larger diameter type fuel pins could not be characterized because of the extensive lot-by-lot scatters. The behavior both in Monju type and in Joyo power-to-melt type fuel pins were similar to each other, and depends largely on the as-fabricated gap width while the effects of linear heat rate and the extent of microstructural evolution were negligible. And fuel pellet centerline melting seems to affect slightly the behavior. The correlation, which describes the extent of relocation both in Monju type and in Joyo power-to-melt type fuel pins, were newly formulated and extrapolated for Joyo Mk-II driver, FFTF driver and its larger diameter type fuel pins. And the behavior in Joyo Mk-II driver seemed to be similar. On the contrary, the similarity with JNC fuel pins was observed case-by-case in FFTF driver and its larger diameter type fuel pins. (author)

Fabrication of Fast Reactor Fuel Pins for Test Irradiations

Energy Technology Data Exchange (ETDEWEB)

Karsten, G. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Dippel, T. [Institute for Radiochemistry, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Laue, H. J. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

1967-09-15

An extended irradiation programme is being carried out for the fuel element development of the Karlsruhe fast breeder project. A very important task within the programme is the testing of plutonium-containing fuel pins in a fast-reactor environment. This paper deals with fabrication of such pins by our laboratories at Karlsruhe. For the fast reactor test positions at present envisaged a fuel with 15% plutonium and the uranium fully enriched is appropriate. Hie mixed oxide is both pelletized and vibro-compacted with smeared densities between 80 and 88% theoretical. The pin design is, for example, such that there are two gas plena at the top and bottom, and one blanket above the fuel with the fuel zone fitting to the test reactor core length. The specifications both for fuel and cladding have been adapted to the special purpose of a fast-breeder reactor - the outer dimensions, the choice of cladding and fuel types, the data used and the kind of tests outline the targets of the development. The fuel fabrication is described in detail, and also the powder line used for vibro-compaction. The source materials for the fuel are oxalate PuO{sub 2} and UO{sub 2} from the UF{sub 6} process. The special problems of mechanical mixing and of plutonium homogeneity have been studied. The development of the sintering technique and grain characteristics for vibratory compactive fuel had to overcome serious problems in order to reach 82-83% theoretical. The performance of the pin fabrication needed a major effort in welding, manufacturing of fits and decontamination of the pin surfaces. This was a stimulation for the development of some very subtle control techniques, for example taking clear X-ray photographs and the tube testing. In general the selection of tests was a special task of the production routine. In conclusion the fabrication of the pins resulted in valuable experiences for the further development of fast reactor fuel elements. (author)

X-ray cinematography on the nuclear fuel and cladding motion diagnostics

International Nuclear Information System (INIS)

Mizuta, Hiroshi; Uruwashi, Shinichi.

1979-01-01

X-ray cinematography has been used for monitoring fuel motion in the out-of-pile fuel pin joule melting experiments for nuclear, liquid metal cooled fast breeder reactor, safety studies related to fuel pin failure, initial fuel motion and thermal fuel-coolant interaction (FCI) of the hypothetical core distractive accident. In order to visually observe the nuclear fuel motion, the X-ray cinematography system consists of an X-ray source located about 5 cm from the test section and an image intensifier located at a corresponding position on the opposite side of the test section. The image from the image intensifier has been recorded both with a high speed camera and video recorder. (author)

Criticality safety studies for plutonium–uranium metal fuel pin fabrication facility

International Nuclear Information System (INIS)

Stephen, Neethu Hanna; Reddy, C.P.

2013-01-01

Highlights: ► Criticality safety limits for PUMP-F facility is identified. ► The fissile mass which can be handled safely during alloy preparation is 10.5 kg. ► The number of fuel slugs which can be handled safely during injection casting is 53. ► The number of fuel slugs which can be handled safely after fuel fabrication is 71. - Abstract: This study focuses on the criticality safety during the fabrication of fast reactor metal fuel pins comprising of the fuel type U–15Pu, U–19Pu and U–19Pu–6Zr in the Plutonium–Uranium Metal fuel Pin fabrication Facility (PUMP-F). Maximum amount of fissile mass which can be handled safely during master alloy preparation, Injection casting and fuel slug preparation following fuel pin fabrication were identified and fixed based on this study. In the induction melting furnace, the fissile mass can be limited to 10.5 kg. During fuel slug preparation and fuel pin fabrication, fuel slugs and pins were arranged in hexagonal and square lattices to identify the most reactive configuration. The number of fuel slugs which can be handled safely after injection casting can be fixed to be 53, whereas after fuel fabrication it is 71

Fabrication and characterization of MX-type fuels and fuel pins

International Nuclear Information System (INIS)

Richter, K.; Bartscher, W.; Benedict, U.; Gueugnon, J.F.; Kutter, H.; Sari, C.; Schmidt, H.E.

1978-01-01

This paper summarizes the most important fabrication parameters and characterization of fuel and fuel pins obtained during the investigation of uranium-plutonium carbides, oxicarbides, carbonitrides and nitrides in the past years at the European Institute for Transuranium Elements at Karlsruhe. All preparation methods discussed are based on carbothermic reduction of a mechanical blend of uranium-plutonium oxide and carbon powder. General data for carbothermic reduction processes are discussed (influence of starting material, homogeneity, control of degree of reaction, etc). A survey of different preparation methods investigated is given. Limitations with respect to temperature and atmosphere for both carbothermic reduction processes and sintering conditions for the different compounds are summarized. A special preparation process for mixed carbonitrides with low nitrogen content (U,Pu)sub(1-x)Nsub(x) in the range 0.1 0 C to 1400 0 C by means of a modulated electron beam technique. A scheme is proposed, which allows to predict the thermal properties of MX fuels on the basis of their chemical composition and porosity. Preparation, preirradiation characterization and final controls of fuel test pins for pellet and vibrocompacted type of pins are described and the most important data summarized for all advanced fuels irradiated at Dounreay (DN1) and Rapsodie Fast Reactor (DN2) within the TU irradiation programme

Measuring deformation of Fuel pin in a Nuclear Fuel Test Rig

Energy Technology Data Exchange (ETDEWEB)

Heo, S. H.; Yang, T. H.; Hong, J. T.; Joung, C. Y.; Ahn, S. H.; Jang, S. Y.; Kim, J. H. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, an LVDT core for measuring the longitudinal displacement of fuel pellets and clad was designed and produced. A signal processing method for the prepared core was investigated. The Nuclear Fuel Test Rig is used to observe changes in the characteristics of the fuel according to the neutron irradiation at HANARO (High-flux Advanced Neutron Application Reactor), which is a research reactor. Which are the strain and internal temperature of the irradiated nuclear fuel and the internal pressure of fuel due to fission gas, the characteristics of the fuel are measured using various sensors such as a thermocouple, SPND and LVDT. In this study, two shaped LVDT (Linear Variable Differential Transformer) cores for displacement measurements were designed and manufactured in order to measure the displacement of a fuel pellet and cladding tube using LVDT sensors for measuring electrical signals by converting the physical variation such as the force and displacement into a linear motion. In addition, signals from the manufactured LVDT sensor were collected and calibrated. Moreover, a method for obtaining the displacement in the core according to the sensing signal was planned. A derived equation can used to predict the change in the position of core. A following study should be conducted to test the output signal and real variation of out-pile system. For further work, a performance verification is required for an in-pile irradiation test.

Design fix for vibration-induced wear in fuel pin bundles

International Nuclear Information System (INIS)

Naas, D.F.; Heck, E.N.

1976-01-01

In summary, results at 45,000 MWd/MTM burnup from the FFTF mixed oxide fuel pin irradiation tests in EBR-II show that reduction of the initial fuel pin bundle clearance and use of 20 percent cold-worked stainless steel ducts virtually eliminate vibration and wear observed in an initial series of 61-pin tests

Modernization of RTC for fabrication of MOX fuel, Vibropac fuel pins and BN-600 FA with weapon grade plutonium

International Nuclear Information System (INIS)

Grachyov, A.F.; Kalygin, V.V.; Skiba, O.V.; Mayorshin, A. A.; Bychkov, A.V.; Kisly, V.A.; Ovsyannikov, Y.F.; Bobrov, D.A.; Mamontov, S.I.; Tsyganov, A.N.; Churutkin, E.I.; Davydov, P.I.; Samosenko, E.A; Shalak, A.R.; Ojima, Hisao

2004-01-01

Since mid 70's RIAR has been performing activities on plutonium involvement in fuel cycle. These activities are considered a stage within the framework of the closed fuel cycle development. Developed at RIAR fuel cycle is based on two technologies: 'dry' process of fuel reprocessing and vibro-packing method for fuel pin fabrication. Due to the available scientific capabilities and a gained experience in operating the technological facilities (ORYOL, SIC) for plutonium (various grade) blending into fuel for fast reactors, RIAR is a participant of the activities aimed at solving these tasks. Under international program RIAR with financial support of JNC (Japan) is modernizing the facility for granulated fuel production, vibro-pac fuel pins and FA fabrication to provide the BN-600 'hybrid' core. In order to provide 'hybrid' core it is necessary to produce (per year): - 1775 kg of granulated MOX-fuel, 6500 fuel pins, 50 fuel assemblies. Potential output of the facility under construction is as follows: - 1800 kg of granulated MOX-fuel per year, 40 fuel pins per shift, 200 FAs for the BN-600 reactor per year. Taking into account domestic and foreign experience in MOX-fuel production, different options were discussed of the equipment layouts in the available premises of chemical technological division of RIAR: - in the shielded manipulator boxes, in the existing hot cells. During construction of the facility in the building under operation the following requirements should be met: - facility must meet all standards and regulations set for nuclear facilities, installation work at the facility must not influence other production programs implemented in the building, engineering supply lines of the facility must be connected to the existing service lines of the building, cost of the activities must not exceed amount of JNC funding. The paper presents results of comparison between two options of the process equipment layout: in boxes and hot cells. This equipment is intended

Advanced disassembling technique of irradiated driver fuel assembly for continuous irradiation of fuel pins

International Nuclear Information System (INIS)

Ichikawa, Shoichi; Haga, Hiroyuki; Katsuyama, Kozo; Maeda, Koji; Nishinoiri, Kenji

2012-01-01

It was necessary to carry out continuous irradiation tests in order to obtain the irradiation data of high burn-up fuel and high neutron dose material for FaCT (Fast Reactor Cycle Technology Development) project. There, the disassembling technique of an irradiated fuel assembly was advanced in order to realize further continuous irradiation tests. Although the conventional disassembling technique had been cutting a lower end-plug of a fuel pin needed to fix fuel pins to an irradiation vehicle, the advanced disassembling technique did not need cutting a lower end-plug. As a result, it was possible to supply many irradiated fuel pins to various continuous irradiation tests for FaCT project. (author)

Performance of advanced oxide fuel pins in EBR-II

International Nuclear Information System (INIS)

Lawrence, L.A.; Jensen, S.M.; Hales, J.W.; Karnesky, R.A.; Makenas, B.J.

1986-05-01

The effects of design and operating parameters on mixed-oxide fuel pin irradiation performance were established for the Hanford Engineering Development Laboratory (HEDL) advanced oxide EBR-II test series. Fourteen fuel pins breached in-reactor with reference 316 SS cladding. Seven of the breaches are attributed to FCMI. Of the remaining seven breached pins, three are attributed to local cladding over-temperatures similar to the breach mechanism for the reference oxide pins irradiated in EBR-II. FCCI was found to be a contributing factor in two high burnup, i.e., 11.7 at. % breaches. The remaining two breaches were attributed to mechanical interaction of UO 2 fuel and fission products accumulated in the lower cladding insulator gap, and a loss of cladding ductility possibly due to liquid metal embrittlement. Fuel smear density appears to have the most significant impact on lifetime. Quantitative evaluations of cladding diameter increases attributed to FCMI, established fuel smear density, burnup, and cladding thickness-to-diameter ratio as the major parameters influencing the extent of cladding strain

Timing analysis of PWR fuel pin failures

International Nuclear Information System (INIS)

Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J.; Straka, M.

1992-09-01

Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B ampersand W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B ampersand W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report

Fuel assemblies for use in nuclear reactors

International Nuclear Information System (INIS)

Schluderberg, D.C.

1981-01-01

A fuel assembly for use in pressurized water cooled nuclear fast breeder reactors is described in which moderator to fuel ratios, conducive to a high Pu-U-D 2 O reactor breeding ratio, are obtained whilst at the same time ensuring accurate spacing of fuel pins without the parasitic losses associated with the use of spacer grids. (U.K.)

Fuel pin design algorithm for conceptual design studies

International Nuclear Information System (INIS)

Uselman, J.P.

1979-01-01

Two models are available which are currently verified by part of the requirements and which are adaptable as algorithms for the complete range. Fuel thermal performance is described by the HEDL SIEX model. Cladding damage and total deformation are determined by the GE GRO-II structural analysis code. A preliminary fuel pin performance model for analysis of (U, P/sub U/)O 2 pins in the COROPT core conceptual design system has been constructed by combining the key elements of SIEX and GRO-II. This memo describes the resulting pin performance model and its interfacing with COROPT system. Some exemplary results are presented

Los Alamos Hot-Cell-Facility modifications for examining FFTF fuel pins

International Nuclear Information System (INIS)

Campbell, B.M.; Ledbetter, J.M.

1982-01-01

Commissioned in 1960, the Wing 9 Hot Cell Facility at Los Alamos was recently modified to meet the needs of the 1980s. Because fuel pins from the Fast Flux Test Facility (FFTF) at the Hanford Engineering Development Laboratory (HEDL) are too long for examination in the original hot cells, we modified cells to accommodate longer fuel pins and to provide other capabilities as well. For instance, the T-3 shipping cask now can be opened in an inert atmosphere that can be maintained for all nondestructive and destructive examinations of the fuel pins. The full-length pins are visually examined and photographed, the wire wrap is removed, and fission gas is sampled. After the fuel pin is cropped, a cap is seal-welded on the section containing the fuel column. This section is then transferred to other cells for gamma-scanning, radiography, profilometry, sectioning for metallography, and chemical analysis

Fabrication of oxide dispersion strengthened ferritic clad fuel pins

International Nuclear Information System (INIS)

Zirker, L.R.; Bottcher, J.H.; Shikakura, S.; Tsai, C.L.

1991-01-01

A resistance butt welding procedure was developed and qualified for joining ferritic fuel pin cladding to end caps. The cladding are INCO MA957 and PNC ODS lots 63DSA and 1DK1, ferritic stainless steels strengthened by oxide dispersion, while the end caps are HT9 a martensitic stainless steel. With adequate parameter control the weld is formed without a residual melt phase and its strength approaches that of the cladding. This welding process required a new design for fuel pin end cap and weld joint. Summaries of the development, characterization, and fabrication processes are given for these fuel pins. 13 refs., 6 figs., 1 tab

Oxide fuel pin transient performance analysis and design with the TEMECH code

International Nuclear Information System (INIS)

Bard, F.E.; Dutt, S.P.; Hinman, C.A.; Hunter, C.W.; Pitner, A.L.

1986-01-01

The TEMECH code is a fast-running, thermal-mechanical-hydraulic, analytical program used to evaluate the transient performance of LMR oxide fuel pins. The code calculates pin deformation and failure probability due to fuel-cladding differential thermal expansion, expansion of fuel upon melting, and fission gas pressurization. The mechanistic fuel model in the code accounts for fuel cracking, crack closure, porosity decrease, and the temperature dependence of fuel creep through the course of the transient. Modeling emphasis has been placed on results obtained from Fuel Cladding Transient Test (FCTT) testing, Transient Fuel Deformation (TFD) tests and TREAT integral fuel pin experiments

Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system

International Nuclear Information System (INIS)

Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

1980-05-01

The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin

Cladding properties under simulated fuel pin transients

International Nuclear Information System (INIS)

Hunter, C.W.; Johnson, G.D.

1975-01-01

A description is given of the HEDL fuel pin testing program utilizing a recently developed Fuel Cladding Transient Tester (FCTT) to generate the requisite mechanical property information on irradiated and unirradiated fast reactor fuel cladding under temperature ramp conditions. The test procedure is described, and data are presented

Development of disassembly and pin chopping technology for FBR spent fuels

International Nuclear Information System (INIS)

Kobayashi, Tsuguyuki; Namba, Takashi; Kawabe, Yukinari; Washiya, Tadahiro

2008-01-01

Japan Atomic Power Company (JAPC) and Japan Atomic Energy Agency (JAEA) have been developing fuel disassembly and fuel pin chopping systems for a future Japanese commercial FBR. At first, the wrapper tube is cut by the slit-cut to pull it out, then the fuel pins are cut by the crop-cut at their end-plugs to separate them from the entrance nozzle. The pins are transferred to the magazine of the chopping machine. A series of tests were performed to develop this procedure. As the result of mechanical cutting tests, the CBN wheel was selected. The slit-cut tests were carried out to evaluated the cutting performance of the wheel. The wrapper tube is normally slit-cut in the circumferential direction. One CBN wheel could cut more than 5 fuel assemblies in this direction. The slit-cut in the axial direction is prepared as provision when the tube is difficult to put out. More work is needed to cut 5mm thick PNC-FMS plate in this direction without damaging the pins beneath it. As the result of the crop-cut tests of end-plugs made of ODS steel, the CBN wheel could cut the 61 pin bundle by two strokes. More work is needed to cut the 217 pin bundle. Fuel pin handling tests were performed to transfer them from the disassembly machine to the chopping machine. The Saucer tray was selected to receive the disassembled pins. All the pins were transferred and loaded into a magazine of the chopping machine. Fuel pin loading tests were conducted to optimize the magazine configuration to make the chopping length within 1.0±0.5 cm. In order to decrease the disturbance during chopping, the width of the magazine was adjusted to be 12 cm and installation of a height adjuster is favourable to control the free space above the pins. (author)

Effects of variations in fuel pellet composition and size on mixed-oxide fuel pin performance

International Nuclear Information System (INIS)

Makenas, B.J.; Jensen, B.W.; Baker, R.B.

1980-10-01

Experiments have been conducted which assess the effects on fuel pin performance of specific minor variations from nominal in both fuel pellet size and pellet composition. Such pellets are generally referred to in the literature as rogue pellets. The effect of these rogue pellets on fuel pin and reactor performance is shown to be minimal

Thermochemistry of nuclear fuels in advanced reactors

International Nuclear Information System (INIS)

Agarwal, Renu

2015-01-01

The presence of a large number of elements, accompanied with steep temperature gradient results in dynamic chemistry during nuclear fuel burn-up. Understanding this chemistry is very important for efficient and safe usage of nuclear fuels. The radioactive nature of these fuels puts lot of constraint on regulatory bodies to ensure their accident free operation in the reactors. One of the common aims of advanced fuels is to achieve high burn-up. As burn-up of the fuel increases, chemistry of fission-products becomes increasingly more important. To understand different phenomenon taking place in-pile, many out of-pile experiments are carried out. Extensive studies of thermodynamic properties, phase analysis, thermophysical property evaluation, fuel-fission product clad compatibility are carried out with relevant compounds and simulated fuels (SIMFUEL). All these data are compiled and jointly evaluated using different computational methods to predict fuel behaviour during burn-up. Only when this combined experimental and theoretical information confirms safe operation of the pin, a test pin is prepared and burnt in a test reactor. Every fuel has a different chemistry and different constraints associated with it. In this talk, various thermo-chemical aspects of some of the advanced fuels, mixed carbide, mixed nitride, 'Pu' rich MOX, 'Th' based AHWR fuels and metallic fuels will be discussed. (author)

Target-fueled nuclear reactor for medical isotope production

Science.gov (United States)

Coats, Richard L.; Parma, Edward J.

2017-06-27

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

The velocity dependent dissolution of spent nuclear fuel in a geologic repository

International Nuclear Information System (INIS)

Nutt, W.M.

1990-02-01

A model describing the dissolution of fission products and transuranic isotopes from spent nuclear fuel into flowing ground water has been developed. This model is divided into two parts. The first part of the model calculates the temperature within a consolidated spent fuel waste form at a given time and ground water velocity. This model was used to investigate whether water flowing at rates representative of a geological repository located at Yucca Mountain, Nevada, will cool a wasteform consisting of consolidated spent nuclear fuel pins. Time and velocity dependent temperature profiles were generated. These profiles were input into the second model, which calculates the dissolution rate of waste isotopes from a spent fuel pin. Two dissolution limiting processes were modeled; the processes are dissolution limited by the solubility limit of an isotopes in the ground water, and dissolution limited by the diffusion of waste isotopes from the interior of a spent fuel pin to the surface where dissolution can occur

Automation of FBTR fuel pin inspection using FPGA

International Nuclear Information System (INIS)

Khare, K.M.; Pai, Siddhesh; Pant, Brijesh; Sendhil Raja, S.; Gupta, P.K.

2011-01-01

A non-contact metrology system for inspection of FBTR fuel pins has been developed. The system consists of a stepper motors driven mechanism for orientation and positioning of FBTR fuel pin, a telecentric imaging system, absolute linear encoder with 0.1 μm resolution and a Field Programmable Gate Array (FPCA) based controller. The FBTR pin assembly is telecentrically illuminated from bottom by a red LED and its shadow graph is imaged using a CCD camera through telecentric imaging lens system. For system control and automation we have used a FPGA that has integrated soft picoblaze processor, X-θ axis motion controller, custom IPs for encoder data acquisition, synchronization circuit, RS485 interface along with other l/Os. Using the Graphical User Interface (GUI) on a PC the system is initialized at home position and the controller provides the trigger signal for start of data acquisition of CCD camera. CCD image of pin and the corresponding X-θ information is captured. After the acquisition of one set of images, the imaging module is moved with a step size pre-programmed to ensure proper stitching of acquired images. The GUI is programmed to analyze these X-θ Images to calculate the required parameters of the fuel pin like the diameter variation, pitch and bow. The details of the instrument and measurements made with it will be presented. (author)

Development of a numerical experimentation method for thermal hydraulics design and evaluation of high burn-up and innovative fuel pins

International Nuclear Information System (INIS)

Ninokata, Hisashi; Misawa, Takeharu; Baglietto, Emilio; Sorokin, A.P.; Maekawa, Isamu; Ohshima, Hiroyuki; Yamaguchi, Akira

2003-03-01

A method of large scale direct numerical simulation of turbulent flows in a high burn-up fuel pin bundle is proposed to evaluate wall shear stress and temperature distributions on the pin surfaces as well as detailed coolant velocity and temperature distributions inside subchannels under various thermal hydraulic conditions. This simulation is aimed at providing a tool to confirm margins to thermal hydraulics design limits of the nuclear fuels and at the same time to be used in design-by-analysis approaches. The method will facilitate thermal hydraulic design of high performance LMFR core fuels characterized by high burn-up, ultra long life, high reliable and safe performances, easiness of operation and maintenance, minimization of radio active wastes, without much relying on such empirical approach as hot spot factor and sub-factors, and above all the high cost mock up experiments. A pseudo direct numerical simulation of turbulence (DNS) code is developed, first on the Cartesian coordinates and then on the curvilinear boundary fit coordinates that enables us to reproduce thermal hydraulics phenomena in such a complicated flow channel as subchannels in a nuclear fuel pin assembly. The coordinate transformation is evaluated and demonstrated to yield correct physical quantities by carrying out computations and comparisons with experimental data with respect to the distributions of various physical quantities and turbulence statistics for fluid flow and heat transfers in various kinds of simple flow channel geometry. Then the boundary fitted pseudo DNS for flows inside an infinite pin array configuration is carried out and compared with available detailed experimental data. In parallel similar calculations are carried out using a commercial code STAR-CD to cross-check the DNS performances. As a results, the pseudo DNS showed reasonable comparisons with experiments as well as the STAR-CD results. Importance of the secondary flow influences is emphasized on the momentum

RAGRAF: a computer code for calculating temperature distributions in multi-pin fuel assemblies in a stagnant gas atmosphere

International Nuclear Information System (INIS)

Eastham, A.

1979-02-01

A method of calculating the temperature distribution in a cross-section of a multi-pin nuclear reactor fuel assembly has been computerised. It utilises the thermal radiation interchange between individual fuel pins in either a square or triangular pitched lattice. A stagnant gas atmosphere within the fuel assembly is assumed which inhibits natural convection but permits thermal conduction between adjacent fuel pins. no restriction is placed upon the shape of wrapper used, but its temperature must always be uniform. RAGRAF has great flexibility because of the many options it provides. Although, essentially, it is a transient code, steady state solutions may be readily identified from successive temperature prints. An enclosure for the assembly wrapper is available, to be included or discarded at will during transient calculations. outside the limit of the assembly wrapper, any type or combination of heat transfer mode may be included. Transient variations in boundary temperature may be included if required. (author)

Power release estimation inside of a fuel pin neighbouring a WWER-440 control rod

International Nuclear Information System (INIS)

Mikus, J.

2006-01-01

This work presents an estimation of the control rod (CR) influence in the WWER-440 core on the power release inside of a fuel pin neighbouring CR, that can have some consequences due to possible static and cyclic loads, for example fuel pin / fuel assembly bowing. For this purpose detailed (usual) axial power distribution measurements were performed in a WWER-440 type core on the light water, zero-power research reactor LR-0 in fuel pins near to an authentic CR model at zero boron concentration in moderator, modelling the conditions at the end of fuel cycle. To demonstrate the CR influence on power distribution inside of one fuel pin neighbouring CR, results of above measurements were used for estimation of the: 1) Axial power distribution inside of the investigated fuel pin in both opposite positions on its pellets surface that are situated to- and outwards CR and corresponding gradient of the (r, z) - power distribution in above opposite positions and 2) Azimuthal power distributions on pellet surface of the investigated fuel pin in horizontal planes at selected axial coordinates. Similar information can be relevant from the viewpoint of the fuel pin failures occurrence investigation

Reirradiation of mixed-oxide fuel pins at increased temperatures

International Nuclear Information System (INIS)

Lawrence, L.A.; Weber, E.T.

1976-05-01

Mixed-oxide fuel pins from EBR-II irradiations were reirradiated in the General Electric Test Reactor (GETR) at higher temperatures than experienced in EBR-II to study effects of the increased operating temperatures on thermal/mechanical and chemical behavior. The response of a mixed-oxide fuel pin to a power increase after having operated at a lower power for a significant portion of its life-time is an area of performance evaluation where little information currently exists. Results show that the cladding diameter changes resulting from the reirradiation are strongly dependent upon both prior burnup level and the magnitude of the temperature increase. Results provide the initial rough outlines of boundaries within which mixed-oxide fuel pins can or cannot tolerate power increases after substantial prior burnup at lower powers

HLM fuel pin bundle experiments in the CIRCE pool facility

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

FABRICE process for the refrabrication of experimental pins in a hot cell, from pins pre-irradiated in power reactors

International Nuclear Information System (INIS)

Vignesoult, N.; Atabek, R.; Ducas, S.

1982-06-01

The Fabrice ''hot cell refabrication'' process for small pins from very long irradiated fuel elements was developed at the CEA to allow parametric studies of the irradiation behavior of pins from nuclear power plants. Since this operation required complete assurance of the validity of the process, qualification of the fabrication was performed on test pins, refabricated in the hot cell, as well as irradiation qualification. The latter qualification was intended to demonstrate that, in identical experimental irradiation conditions, the refabricated Fabrice pins behaved in the same way as whole pins with the same initial characteristics. This qualification of the Fabrice process, dealing with more than twenty pins at different burnups, showed that fabrication did not alter: the inherent characteristics of the sampled fuel element and the irradiation behavior of the sampled fuel element [fr

TRANSPA: a code for transient thermal analysis of a single fuel pin

International Nuclear Information System (INIS)

Prenger, F.C.

1985-02-01

An analytical model (TRANSPA) for the transient thermal analysis of a single uranium carbide fuel pin was developed. This model uses thermal boundary conditions obtained from COBRA-WC output and calculates the transient thermal response of a single fuel pin to changes in internal power generation, coolant flowrate, or fuel pin physical configuration. The model uses the MITAS finite difference thermal analyzer. MITAS provides the means to input separate conductance models through the use of a user subroutine input capability. The model is a lumped-mass representation of the fuel pin using 26 nodes and 42 conductors. Run time for each transient analysis is approximately one minute of central processor time on the NOS operating system

Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

1980-05-01

The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin.

Fuel pin failure in the PFR/TREAT experiments

International Nuclear Information System (INIS)

Herbert, R.; Hunter, C.W.; Kramer, J.M.; Wood, M.H.; Wright, A.E.

1986-01-01

The PFR/TREAT safety testing programme involves the transient testing of fresh and pre-irradiated UK and US fuel pins. This paper summarizes the experimental and calculational results obtained to date on fuel pin failure during transient overpower (resulting from an accidental addition of resolivity) and transient undercooling followed by overpower (arising from an accidental stoppage of the primary sodium circulating pumps) accidents. Companion papers at this conference address: (I) the progress and future plans of the programme, and (II) post-failure material movements

Serviceability of rod ceramic fuel pins on motoring conditions of FTP or NEMF reactor

International Nuclear Information System (INIS)

Deryavko, I.I.

2004-01-01

The operation conditions of rod ceramic fuel pins in the running hydrogen-cooled technological canals of FTP or NEMF reactor on the motoring conditions are considered. The available postreactor researches of the fuel pins are presented and the additional postreactor researches of fuel pins, tested on this mode in IVG.1 and IRGIT reactors, are carried out. The fuel pins serviceability on motoring conditions of FTP or NEF reactor operation is concluded. (author)

Fuel pin response to an overpower transient in an LMFBR

International Nuclear Information System (INIS)

Grosberg, A.J.; Head, J.L.

1979-01-01

This paper describes a method by which the ability of a whole-core code accurately to predict the time and location of the first fuel pin failures may be tested. The method involves the use of a relatively simple whole-core code to 'drive' a sophisticated fuel pin code, which is far too complex to be used within a whole-core code but which is potentially capable of modelling reliably the response of an individual fuel pin. The method cannot follow accurately the subsequent course of the transient because the simple whole-core code does not model the reactivity effects of events which may follow pin failure. The codes used were the simple whole-core code FUTURE and the fuel pin behaviour code FRUMP. The paper describes an application of the method to analyse a hypothetical LMFBR accident in which the control rods were assumed to be driven from the core at maximum speed, with all trip circuits failed. Taking 0.5% clad strain as a clad failure criterion, failure was predicted to occur at the top of the active core at about 10s into the transient. A repeat analysis, using an alternative clad yield criterion which is thought to be more realistic, indicated failure at the same position but 24s into the transient. This is after the onset of sodium boiling. Pin failure at the top of the core are likely to cause negative reactivity changes. In this hypothetical accident, pin failures are likely, therefore, to have a moderating effect on the course of the transient. (orig.)

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

Nodal methods for calculating nuclear reactor transients, control rod patterns, and fuel pin powers

International Nuclear Information System (INIS)

Cho, Byungoh.

1990-01-01

Nodal methods which are used to calculate reactor transients, control rod patterns, and fuel pin powers are investigated. The 3-D nodal code, STORM, has been modified to perform these calculations. Several numerical examples lead to the following conclusions: (1) By employing a thermal leakage-to-absorption ratio (TLAR) approximation for the spatial shape of the thermal fluxes for the 3-D Langenbuch-Maurer-Werner (LMW) and the superprompt critical transient problems, the convergence of the conventional two-group scheme is accelerated. (2) By employing the steepest-ascent hill climbing search with heuristic strategies, Optimum Control Rod Pattern Searcher (OCRPS) is developed for solving control rod positioning problem in BWRs. Using the method of approximation programming the objective function and the nuclear and thermal-hydraulic constraints are modified as heuristic functions that guide the search. The test calculations have demonstrated that, for the first cycle of the Edwin Hatch Unit number-sign 2 reactor, OCRPS shows excellent performance for finding a series of optimum control rod patterns for six burnup steps during the operating cycle. (3) For the modified two-dimensional EPRI-9R problem, the least square second-order polynomial flux expansion method was demonstrated to be computationally about 30 times faster than a fine-mesh finite difference calculation in order to achieve comparable accuracy for pin powers. The basic assumption of this method is that the reconstructed flux can be expressed as a product of an assembly form function and a second-order polynomial function

Method and device for cleaning fuel pins

International Nuclear Information System (INIS)

Matsumoto, Kaname; Oohigashi, Yoshiaki.

1985-01-01

Purpose: To remove clads or scales deposited on the outer surface of fuel pins in BWR type reactors. Method: A fuel assembly taken out of a reactor core is vertically contained without detaching a channel box in a scrubber tower disposed in a liquid tight manner within a fuel pool. Then, a specifically prepared slurry is caused to flow and uprise from the bottom of the scrubber tower into the channel box and then discharged from the top of the tower. The slurry is prepared by mixing pure water and granules (for example, as activated carbon, ion exchanger resin, iron and molecular sieve) of such a granular size as not causing clogging in the channel box of the fuel assembly and having a larger specific gravity than pure water. The slurry flown into the channel box scrubs the surface of fuel pins to scrape off clads or scales. Then, discharged slurry is sent to a hydraulic cyclone to separate the granules from the clads or scales. (Ikeda, J.)

Assessment of clad integrity of PHWR fuel pin following a postulated severe accident

International Nuclear Information System (INIS)

Dutta, B.K.; Kushwaha, H.S.; Venkat Raj, V.

2000-01-01

A mechanistic fuel performance analysis code FAIR has been developed. The code can analyse fuel pins with free standing as well as collapsible clad under normal, off-normal and accident conditions of reactors. The code FAIR is capable of analysing the effects of high burnup on fuel behaviour. The code incorporates finite element based thermo-mechanical module for computing transient temperature distribution and thermal-elastic-plastic stresses in the fuel pin. A number of high temperature thermo-physical and thermo-mechanical models also have been incorporated for analysing fuel pins subjected to severe accident scenario. The present paper describes salient features of code FAIR and assessment of clad integrity of PHWR fuel pins with different initial burnup subjected to severe accident scenario. (author)

Evaluation of bundle duct interaction by out-of-pile compression test of FBR fuel pin bundles

Energy Technology Data Exchange (ETDEWEB)

Tanaka, Kosuke; Yamamoto, Yuji; Nagamine, Tsuyoshi; Maeda, Koji [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

2001-06-01

Bundle duct interaction (BDI) caused by expansion of fuel pin bundle is a main factor to limit the fuel lifetime. Therefore, it is important for the design of fast reactor fuel assembly to understand the fuel pin deformation behavior under BDI condition. In order to understand the fuel pin deformation behavior under BDI condition, out-of-pile compression tests were conducted for FBR fuel pin bundle by use of X-ray CT equipment. In these compression tests, two kinds of fuel pin bundles were conducted. One was the fuel pin bundle with the short wire-pitch and the other was the fuel pin bundle with the short wire-pitch and large diameter claddings. The general discussions were also performed based on the results of out-of-pile compression tests obtained by use of X-ray CT equipment in the previous work. Following results were obtained. 1) The occurrence of the pin-to-duct contact depends on the wire-pitch. In the fuel pin bundle with large wire-pitch, the pin-to-duct contact occurred at the early stage of BDI. The reason of this result is due to the low bowing rigidity of the fuel pins with long wire-pitch. 2) The value of the ovalation stiffness strongly depends on the geometry of cladding (diameter, thickness) and especially on wire-pitch. This result in this work revealed that the occurrence of the pin-to-duct contact depends on the value of the ovalation stiffness. 3) The occurrence of wire dispersion and dispersive displacement of pins depends on the wire-pitch strongly. In the fuel pin bundle with the long wire-pitch, the occurrence of the above-mentioned suppression mechanism to BDI is remarkable. 4) The suppression mechanism to BDI of the fuel pin bundle with the long wire-pitch is elastic oval deformation of cladding, wire dispersion and dispersive displacement of pins. On the other hand, the elastic and plastic oval deformation of cladding is the major suppression mechanism to BDI in the fuel pin bundle with the short wire-pitch. 5) The appearance of

Encapsulation of spent nuclear fuel in ceramic materials

International Nuclear Information System (INIS)

Forberg, S.; Westermark, T.

1983-03-01

The international situation with regard to deposition of spent nuclear fuel is surveyed, with emphasis on encapsulation in ceramic materials. The feasibility and advantages of ceramic containers, thermodynamic stable in groundwater, are discussed as well as the possibility to ensure that stability for longevity by engineered measures. The design prerequisite are summarized and suggestions are made for a conceptual design, comprising rutile containers with stacks of coiled fuel pins. A novel technique is suggested for the homogeneous sealing of rutile containers at low temperatures. acceptable also for the fuel pin package. Key points are given for research, demonstration and verifications of the design foundations and for future improvements. Of which a few ideas are exemplified. (author)

Gamma scanning of mixed carbide and oxide fuel pins irradiated in FBTR

International Nuclear Information System (INIS)

Jayaraj, V.V.; Padalakshmi, M.; Ulaganathan, T.; Venkiteswaran, C.N.; Divakar, R.; Joseph, Jojo; Bhaduri, A.K.

2016-01-01

Fission in nuclear fuels results in a number of fission products that are gamma emitters in the energy range of 100 keV to 3 MeV. The gamma emitting fission products are therefore amenable for detection by gamma detectors. Assessment of the fission product distribution and their migration behavior through gamma scanning is important for characterizing the in reactor behavior of the fuel. Gamma scanning is an important non destructive technique used to evaluate the behavior of irradiated fuels. As a part of Post Irradiation Examinations (PIE), axial gamma scanning has been carried out on selected fuel pins of the FBTR Mark I mixed carbide fuel sub-assemblies and PFBR MOX test fuel sub-assembly irradiated in FBTR. This paper covers the results of gamma scanning and correlation of gamma scanning results with other PIE techniques

Beginning-of-life gap closure behaviour of experimental PFBR MOX fuel pin

International Nuclear Information System (INIS)

Jayaraj, V.V.; Padalakshmi, M.; Ojha, B.K.; Padma Prabu, C.; Saravanan, T.; Venkiteswaran, C.N.; Philip, John; Muralidharan, N.G.; Joseph, Jojo; Kasiviswanathan, K.V.; Jayakumar, T.

2011-01-01

Mixed oxide fuel with 22 % and 29% plutonium is chosen as the fuel for PFBR for the two fissile zones. Due to the fabrication tolerances in the pellet diameter, fuel has to be preconditioned at a lower linear power for a brief period before raising the power to the rated value of 450 W/cm. PIE was done on an experimental MOX fuel pin irradiated in FBTR for 13 days at a linear power of 400 W/cm for gap closure studies with the objective of optimising the duration of pre-conditioning before raising the power to the design value of 450 W/cm. X-radiography and remote metallography was done on the fuel pin to estimate the axial fuel column elongation and fuel-clad gap. Remote metallography of the fuel pin cross-sections at five axial locations of the fuel column and the subsequent fuel-clad gap measurement has indicated that the average radial gap has reduced from the pre-irradiation value of 75-110 microns to around 12-13 microns along the entire length of the fuel column. This paper will describe the details of examinations and results of the PIE carried out on the MOX fuel pin. (author)

One- and two-dimension effects on fuel pin lifetime

International Nuclear Information System (INIS)

Stephen, J.D.; Biancheria, A.; Leibnitz, D.; O'Reilly, B.D.; Liu, Y.Y.; Labar, M.P.; Gneiting, B.C.

1979-01-01

Lifetime, or breach of the cladding, is a difficult performance limit to establish in fuel pin design. The significant benefits of high plant capacity factor favor conservative design to eliminate downtime or partial power operation caused by the breach limit; however, overly conservative design produces significant penalties. The LIFE system is being applied to help understand the range between operation and breach so that appropriate design margins can be selected. Standards are being developed in the USA to assure the structural integrity of all core components. These standards will provide guidelines to account for the failure mechanisms observed in the high temperature, high fluence core environment. The work to date indicates that creep rupture is the most important failure mechanism for mixed-oxide fuel pins during normal operation and slow power changes. The local cumulative creep rupture damage fraction (CDF) has been adopted as the parameter to assess the approach to failure. Several oxide breached pins and siblings have been studied For example, the P23B-73 pin was an FFTR driver design pin irradiated in EBR-II which failed at 10 at,% burnup. Initial evaluation based on LIFE3 led to the conclusion that the pin should not have failed. Further analyses determined the sensitivity of the breach prediction to the time-to-rupture correlation, cladding temperature, and fuel-fission product swelling (which had not been modeled in LIFE3). The uncertainties in the time-to-rupture correlation have been established. But LIFE is a one-dimensional model. The TWOD code is complete, and development of the best way to couple LIFE and TWOD for lifetime analysis is in progress. Two preliminary conclusions from analysis of representative oxide pin geometries are, first, that the circumferential stress distribution may not peak at the hot spot, but the damage (CDF) does. And second, that the effect of stress concentrations near fuel cracks on cladding creep damage is small

Investigation of the ramp testing behaviour of fuel pins with different diameters

International Nuclear Information System (INIS)

Pott, G.; Herren, M.; Wigger, B.

1979-09-01

The aim of these experiments was the investigation of the influence of different fuel pin diameter on the ramp testing behaviour. Fuel elements with diameter between 10,75 and 15,6 mm and different cladding thickness had been ramptested in the HBWR (Halden Boiling Water Reactor) after preirradiated in the same facility. Fuel pins with the smallest diameter of 10,75 mm failed. This was indicated by fission gas release measurement. Metallographic examination showed these failure were caused by hydride blisters. A systematic influence of fuel pin diameter and cladding thickness on the ramptesting behaviour was not observed. (orig.) [de

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-02-01

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

Advanced control system for the Integral Fast Reactor fuel pin processor

International Nuclear Information System (INIS)

Lau, L.D.; Randall, P.F.; Benedict, R.W.; Levinskas, D.

1993-01-01

A computerized control system has been developed for the remotely-operated fuel pin processor used in the Integral Fast Reactor Program, Fuel Cycle Facility (FCF). The pin processor remotely shears cast EBR- reactor fuel pins to length, inspects them for diameter, straightness, length, and weight, and then inserts acceptable pins into new sodium-loaded stainless-steel fuel element jackets. Two main components comprise the control system: (1) a programmable logic controller (PLC), together with various input/output modules and associated relay ladder-logic associated computer software. The PLC system controls the remote operation of the machine as directed by the OCS, and also monitors the machine operation to make operational data available to the OCS. The OCS allows operator control of the machine, provides nearly real-time viewing of the operational data, allows on-line changes of machine operational parameters, and records the collected data for each acceptable pin on a central data archiving computer. The two main components of the control system provide the operator with various levels of control ranging from manual operation to completely automatic operation by means of a graphic touch screen interface

TEMP: a computer code to calculate fuel pin temperatures during a transient

International Nuclear Information System (INIS)

Bard, F.E.; Christensen, B.Y.; Gneiting, B.C.

1980-04-01

The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method

Development of 3D pseudo pin-by-pin calculation methodology in ANC

International Nuclear Information System (INIS)

Zhang, B.; Mayhue, L.; Huria, H.; Ivanov, B.

2012-01-01

Advanced cores and fuel assembly designs have been developed to improve operational flexibility, economic performance and further enhance safety features of nuclear power plants. The simulation of these new designs, along with strong heterogeneous fuel loading, have brought new challenges to the reactor physics methodologies currently employed in the industrial codes for core analyses. Control rod insertion during normal operation is one operational feature in the AP1000 R plant of Westinghouse next generation Pressurized Water Reactor (PWR) design. This design improves its operational flexibility and efficiency but significantly challenges the conventional reactor physics methods, especially in pin power calculations. The mixture loading of fuel assemblies with significant neutron spectrums causes a strong interaction between different fuel assembly types that is not fully captured with the current core design codes. To overcome the weaknesses of the conventional methods, Westinghouse has developed a state-of-the-art 3D Pin-by-Pin Calculation Methodology (P3C) and successfully implemented in the Westinghouse core design code ANC. The new methodology has been qualified and licensed for pin power prediction. The 3D P3C methodology along with its application and validation will be discussed in the paper. (authors)

Setting for technological control of vibropacked uranium-plutonium fuel pins

International Nuclear Information System (INIS)

Golushko, V.V.; Semenov, A.L.; Chukhlova, O.P.; Kuznetsov, A.M.; Korchkov, Yu.N.; Kandrashina, T.A.

1991-01-01

Scanning set-up providing for control of fuel pins by quality of fuel distribution in them is described. The gamma absorption method of fuel density measurement and the method of its own radiation registration are applied. Scintillation detection blocks are used in the measuring equipment mainly consisting of standard CAMAC blocks. Automation of measurements is performed on the basis of the computer complex MERA-60. A complex of programs for automation of the procedures under way is developed, when the facility operates within the test production line of vibroracked uranium-plutonium fuel pins. 6 refs.; 4 figs.; 1 tabs

Nuclear reactor fuel element assemblies

International Nuclear Information System (INIS)

Raven, L.F.

1975-01-01

A spacer grid for a nuclear fuel element comprises a plurality of cojointed cylindrical ferrules adapted to receive a nuclear fuel pin. Each ferrule has a pair of circumferentially spaced rigid stop members extending inside the ferrule and a spring locating member attached to the ferrule and also extending from the ferrule wall inwardly thereof at such a circumferential spacing relative to the rigid stop members that the line of action of the spring locating member passes in opposition to and between the rigid stop members which lie in the same diametric plane. At least some of the cylindrical ferrules have one rim shaped to promote turbulence in fluid flowing through the grid. (Official Gazette)

Apparatus and process for separating end of nuclear fuel assembly

International Nuclear Information System (INIS)

Beneck, J.A.; Quayre, C.; Moreau, J.R.M.; Vermeille, D.E.A.

1989-01-01

The apparatus for cutting the guide tubes of a nuclear fuel assembly below the bottom nozzle has a framework with removable fixation and centering means on the bottom nozzle. Cutting devices in the form of hollow pins are inserted simultaneously into all the guide tubes. Each pin contains a shaft that is moved axially inside it by the expansion system to deploy or retract a cutting edge near the lower end of the pin. A single motor using a gear wheel system, rotates all the cutting pins simultaneously [fr

Fast reactor fuel pin behavior analyses in a LOF type transient event

International Nuclear Information System (INIS)

Mizuno, Tomoyasu; Koyama, Shin-ichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

2013-06-01

In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, fuel pin behavior calculations were made using the fast reactor fuel pin performance code CEDAR. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross and Stoute type gap conductance model and constant gap conductance model, which is used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross and Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross and Stoute type gap conductance model which is thought to be realistic. (author)

Integrated quality status and inventory tracking system for FFTF driver fuel pins

International Nuclear Information System (INIS)

Gottschalk, G.P.

1979-11-01

An integrated system for quality status and inventory tracking of Fast Flux Test Facility (FFTF) driver fuel pins has been developed. Automated fuel pin identification systems, a distributed computer network, and a data base are used to implement the tracking system

Grids for nuclear fuel elements

International Nuclear Information System (INIS)

Nicholson, G.

1980-01-01

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

Top-nozzle mounted replacement guide pin assemblies

International Nuclear Information System (INIS)

Gilmore, C.B.; Andrews, W.H.

1993-01-01

A replacement guide pin assembly is provided for aligning a nuclear fuel assembly with an upper core plate of a nuclear reactor core. The guide pin assembly includes a guide pin body having a radially expandable base insertable within a hole in the top nozzle, a ferrule insertable within the guide pin base and capable of imparting a radially and outwardly directed force on the expandable base to expand it within the hole of the top nozzle and thereby secure the guide pin body to the top nozzle in response to a predetermined displacement of the ferrule relative to the guide pin body along its longitudinal axis, and a lock screw interfitted with the ferrule and threaded into the guide pin body so as to produce the predetermined displacement of the ferrule. (author)

Parallel two-phase-flow-induced vibrations in fuel pin model

International Nuclear Information System (INIS)

Hara, Fumio; Yamashita, Tadashi

1978-01-01

This paper reports the experimental results of vibrations of a fuel pin model -herein meaning the essential form of a fuel pin from the standpoint of vibration- in a parallel air-and-water two-phase flow. The essential part of the experimental apparatus consisted of a flat elastic strip made of stainless steel, both ends of which were firmly supported in a circular channel conveying the two-phase fluid. Vibrational strain of the fuel pin model, pressure fluctuation of the two-phase flow and two-phase-flow void signals were measured. Statistical measures such as power spectral density, variance and correlation function were calculated. The authors obtained (1) the relation between variance of vibrational strain and two-phase-flow velocity, (2) the relation between variance of vibrational strain and two-phase-flow pressure fluctuation, (3) frequency characteristics of variance of vibrational strain against the dominant frequency of the two-phase-flow pressure fluctuation, and (4) frequency characteristics of variance of vibrational strain against the dominant frequency of two-phase-flow void signals. The authors conclude that there exist two kinds of excitation mechanisms in vibrations of a fuel pin model inserted in a parallel air-and-water two-phase flow; namely, (1) parametric excitation, which occurs when the fundamental natural frequency of the fuel pin model is related to the dominant travelling frequency of water slugs in the two-phase flow by the ratio 1/2, 1/1, 3/2 and so on; and (2) vibrational resonance, which occurs when the fundamental frequency coincides with the dominant frequency of the two-phase-flow pressure fluctuation. (auth.)

Calculation of fuel pin failure timing under LOCA conditions

International Nuclear Information System (INIS)

Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.

1991-10-01

The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B ampersand W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs

WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

International Nuclear Information System (INIS)

Vitkova, M.; Manolova, M.; Stefanova, S.; Simeonova, V.; Passage, G.; Lassmann, K.

1994-01-01

PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: 1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; 2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; 3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs

AGR fuel pin pellet-clad interaction failure limits and activity release fractions

International Nuclear Information System (INIS)

Hughes, H.; Hargreaves, R.

1985-01-01

The limiting conditions beyond which pellet-clad interaction can flail AGR fuel are described. They have been determined by many experiments involving post-irradiation examination and testing, loop experiments and cycling and up-rating of both individual fuel stringers and the whole WAGR core. The mechanisms causing this interaction are well understood and are quantitatively expressed in computer codes. Strain concentration effects over fuel cracks determine power cycling endurance while additional strain concentrations at clad ridges and from cross pin temperature gradients contribute to up-rating failures. An equation summarising tube burst test data so as to determine the ductility available at any transient is given. The hollow fuel and more ductile clad of the Civil AGR fuel pins leads to a much improved performance over the original fuel design. The Civil AGRs operate well within these limiting conditions and substantial increases beyond the design burn-up are confidently expected. The activity release on pin failure and its development during continued operation of failed fuel have also been investigated. A retention of radioiodine and caesium of 90-99% compared to the noble gases has been demonstrated. Measured fission gas releases into the free volume of Civil AGR fuel pins have been very low (< 0.1%)

WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

Energy Technology Data Exchange (ETDEWEB)

Vitkova, M; Manolova, M; Stefanova, S; Simeonova, V; Passage, G [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika; Kharalampieva, Ts [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Lassmann, K [European Atomic Energy Community, Karlsruhe (Germany). European Inst. for Transuranium Elements

1994-12-31

PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: (1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; (2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; (3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs.

Laser pulse heating of nuclear fuels for simulation of reactor power ...

Indian Academy of Sciences (India)

As a prelude to work on irradiated nuclear fuel specimens, pilot studies on ... But facilities for making such fuel pins, and carrying out test irradiations for the .... To graduate to a complete simulation set-up, the following are needed: (a) A means.

Inserts for nuclear fuel elements

International Nuclear Information System (INIS)

Cragg, P.J.

1982-01-01

An insert for a nuclear fuel pin which comprises a strip. The strip carries notches, which enable a coding arrangement to be carried on the strip. The notches may be of differing sizes and the coding on the strip includes identification and identification checking data. Each notch on the strip may give rise to a signal pulse which is counted by a detector to avoid errors. (author)

SIFAIL: a subprogram to calculate cladding deformation and damage for fast reactor fuel pins

International Nuclear Information System (INIS)

Wilson, D.R.; Dutt, D.S.

1979-05-01

SIFAIL is a series of subroutines used in conjunction with the thermal performance models of SIEX to assist in the evaluation of mechanical performance of mixed uranium plutonium oxide fuel pins. Cladding deformations due to swelling and creep are calculated. These have been compared to post-irradiation data from fuel pin tests in EBR-II. Several fuel pin cladding failure criteria (cumulative damage, total strain, and thermal creep strain) are evaluated to provide the fuel pin designer with a basis to select design parameters. SIFAIL allows the user many property options for cladding material. Code input is limited to geometric and environmental parameters, with a consistent set of material properties provided by the code. The simplified, yet adequate, thin wall stress--strain calculations provide a reliable estimate of fuel pin mechanical performance, while requiring a small amount of core storage and computer running time

Fabrication of uranium-plutonium mixed nitride fuel pins (88F-5A) for first irradiation test at JMTR

International Nuclear Information System (INIS)

Suzuki, Yasufumi; Iwai, Takashi; Arai, Yasuo; Sasayama, Tatsuo; Shiozawa, Ken-ichi; Ohmichi, Toshihiko; Handa, Muneo

1990-07-01

A couple of uranium-plutonium mixed nitride fuel pins was fabricated for the first irradiation tests at JMTR for the purpose of understanding the irradiation behavior and establishing the feasibility of nitride fuels as advanced FBR fuels. The one of the pins was fitted with thermocouples in order to observe the central fuel temperature. In this report, the fabrication procedure of the pins such as pin design, fuel pellet fabrication and characterizations, welding of fuel pins, and inspection of pins are described, together with the outline of the new TIG welder installed recently. (author)

Creep relaxation of fuel pin bending and ovalling stresses

International Nuclear Information System (INIS)

Chan, D.P.; Jackson, R.J.

1979-06-01

Analytical methods for calculating fuel pin cladding bending and ovalling stresses due to pin bundle-duct mechanical interaction taking into account nonlinear creep are presented. Calculated results are in close agreement with finite element results by MARC-CDC program. The methods are used to investigate the effect of creep on the FTR fuel cladding bending and ovalling stresses. It is concluded that the cladding of 316 SS 20% CW and reference design has high creep rates in the FTR core region to keep the bending and ovalling stresses to low levels

Modeling of WWER-440 Fuel Pin Behavior at Extended Burn-up

International Nuclear Information System (INIS)

El-Koliel, M.S.; Abou-Zaid, A.A.; El-Kafas, A.A.

2004-01-01

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWER's as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased to 60 to 70 Mwd/kg U. The change in the fuel radial power distribution as a function of fuel burn up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO 2 fuel pin were evaluated using MCNP 4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted fission gas release calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. a computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented

Metallographic examinations of the wear-marks on fuel pins of the KNK II/2 fuel assembly NY-308

International Nuclear Information System (INIS)

Patzer, G.

1987-12-01

On the fuel pins and pin spacers of the fuel assembly NY-308 of the second core of KNK II pronounced wear marks had been found in the area of the contact points. In order to determine the exact form of the marks, metallographic investigations were performed on two test pieces of fuel pins in the Hot Cells of the KfK Karlsruhe. It was found that the wear marks did show the already observed stratified structure. Next to the unchanged cladding area there is a peripheral zone with modified grain structure, followed by a layer of moved material and finally there is a flake-like zone of accumulated cladding material at the lower end of the wear marks. Longitudinal cuts do not show grain deformations, which could indicate axial friction forces between pin and spacer. The wear marks are rapidly dropping to their maximum depth at the ends and the depth shows a relatively uniform pattern between both. The findings are confirming the picture, that a stirring movement of the fuel pins took place, which caused adhesive wear [de

Fuel transfer manipulator for liquid metal nuclear reactors

International Nuclear Information System (INIS)

Sturges, R.H.

1983-01-01

A manipulator for transferring fuel assemblies between inclined fuel chutes of a liquid metal nuclear reactor installation. Hoisting means are mounted on a mount supported by beams pivotably attached by pins to the mount and to the floor in such a manner that pivoting of the beams causes movement and tilting of a hoist tube between positions of alignment with the inclined chutes. (author)

FFTF metal fuel pin sodium bond quality verification

International Nuclear Information System (INIS)

Pitner, A.L.; Dittmer, J.O.

1988-12-01

The Fast Flux Test Facility (FFTF) Series III driver fuel design consists of U-10Zr fuel slugs contained in a ferritic alloy cladding. A liquid metal, sodium bond between the fuel and cladding is required to prevent unacceptable temperatures during operation. Excessive voiding or porosity in the sodium thermal bond could result in localized fuel melting during irradiation. It is therefore imperative that bond quality be verified during fabrication of these metal fuel pins prior to irradiation. This document discusses this verification

French approach in fuel pin modelling for fast reactors

Energy Technology Data Exchange (ETDEWEB)

Pascard, R [CEA-Centre de Fontenay-aux-Roses, Fontenay-aux-Roses (France)

1979-12-01

The purpose of this paper is to present the general philosophy on the problem of fuel modelling now prevailing in France after a twelve years period of tremendously increasing knowledge on fuel behavior. When the Rapsodie fuel pin was designed in 1962 , little was known about the behavior of a mixed oxide fuel pin under fast flux ; but a large body of knowledge on UO{sub 2} behavior in thermal reactor was available together with some sparse irradiation results on (U Pu)O{sub 2} in French experimental reactors. The performances assigned to the pin were then rather modest in rating (400 w/cm) and in burnup (30,000 MWd/t). The AISI 316 steel in solution annealed state was chosen as cladding material. The clad itself was supposed to deform by thermal creep due to fission gas pressure (100% release), and was affected consequently by a strain limit criteria. The importance of clad temperature ({approx}650 deg.) was considered only in connection with thermal creep, the possibility of a chemical reaction between mixed oxide and clad being at that time hardly suspected. Rapsodie had only been at full power for a few months when appeared the evidence of stainless steel swelling under a fast neutrons flux. This swelling was observed on Rapsodie pins as soon as they experienced sufficient neutrons dose, roughly one year later. This entirely new problem came immediately in the front stage (and is still of major importance today), and was at the origin of the change from the Rapsodie to the Fortissimo core in order to accelerate materials testing versus void swelling by multiplying the flux by a factor two. Even with unforeseen swelling, the design of the Rapsodie and later on Fortissimo pin, allowed not only to reach the goal burnup, but to increase it steadily to roughly 100,000 MWd/t. Since then, the French approach in fuel pin design has still retained something of its original simplicity, and technological efficiency, attitude which is justified by the following

Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins

International Nuclear Information System (INIS)

Makenas, B.J.; Lawrence, L.A.; Jensen, B.W.

1982-05-01

Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 (E > .1 MeV) and 8.8 x 10 22 n/cm 2 (E > .1 MeV) respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %

Radiographic examination methods for fuel pins

International Nuclear Information System (INIS)

Smirnov, V.P.; Dvoretskii, V.G.

1987-11-01

To study the fast neutron reactor fuel pins structure the NIIAR Institute used x diffraction, neutronic radiography and autoradiographies. The two first methods are used for internal macrostructure studies, the third method for the plutonium and uranium radial distribution. These methods and the main results are indicated in this document [fr

Fuel pin behavior under slow ramp-type transient-overpower conditions in the CABRI-FAST experiments

International Nuclear Information System (INIS)

Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, Jean

2009-01-01

In the CABRI-FAST experimental program, four in-pile tests were performed with slow power-ramp-type transient-overpower conditions (called hereafter as 'slow TOP') to study transient fuel pin behavior under inadvertent control rod withdrawal events in liquid metal cooled fast breeder reactors. Annular-pellet fuel pins were used in three tests, while a solid-pellet fuel pin was used in the other test. All of these pins were pre-irradiated in Phenix. The slow TOP test with a solid-pellet fuel pin was realized as a comparatory test against an existing test (E12) in the CABRI-2 program. In the CABRI-FAST test (BCF1), a power ramp rate of 3% Po/s was applied, while in the CABRI-2 test, 1% Po/s was adopted. Moreover, overpower condition was maintained for a few seconds beyond the observed pin failure in the BCF1 test. In spite of the different power ramp rates, evaluated fuel thermal conditions at the observed failure time are quite similar. The continued overpower condition in the BCF1 test resulted in gradual degradation of the pin structure providing information effective for evaluation of various accident scenarios. Three slow TOP tests with the annular fuel in the CABRI-FAST program resulted in no pin failure showing high failure threshold. Based on post-test examination data and a theoretical evaluation, it was concluded that intra-pin free spaces, such as central hole, macroscopic cracks and fuel-cladding gap effectively mitigated fuel cladding mechanical interaction. It was also clarified that cavity pressurization became effective only in case of very large amount of fuel melting. Furthermore, such cavity pressurization was effectively mitigated by a molten-fuel squirting into the upper blanket region pushing the blanket pellets upward. These CABRI FAST slow TOP tests, in combination with the existing CABRI and TREAT tests, provided an extended slow TOP test database with various fuel and transient conditions. (author)

Nuclear fuel sub-assemblies

International Nuclear Information System (INIS)

Dodd, J.A.; Butterfield, C.E.; Waite, E.

1979-01-01

A fast reactor fuel sub-assembly has honeycomb grids for laterally supporting the fuel pins. The grids are of two series and are arranged alternately along the bundle. The grids of a first series provide a discrete cell for each pin but the grids of the second series have a peripheral group of cells only. The grids of the second series provide intermediate support of the edge pins to restrain bow. (author)

FEA stress analysis considering cavity formation of metallic fuel pin under transient state

Energy Technology Data Exchange (ETDEWEB)

Jung, Hyun-Woo; Oh, Young-Ryun; Kim, Yun-Jae [Korea University, Seoul (Korea, Republic of)

2016-05-15

The aim of this research is to study the stress state of the fuel and the cladding under transient state using the commercial finite element analysis software, ABAQUS v6.13. It is checked out that the gap distance between the fuel and the cladding is a major factor determining FCMI stress. In this regard, initial boundary condition of the fuel pin such as the initial gap distance should be set carefully when the stress analysis of the fuel pin under transient state is conducted. In case of simulating cavity formation, it is confirmed that the new cavity simulation model that elements in cavity region lose their stiffness is valid. There is a great deal of research into SFR, which is one of GEN IV reactors. When it comes to the accidents of SFR, there are two cases of accident process. One of them is In-pin process that molten fuel is discharged into upper plenum. The other is Ex-pin process that the molten fuel is discharged into coolant because of breakage of cladding.

Infinite fuel element simulation of pin power distributions and control blade history in a BWR fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Li, J.; Nuenighoff, K.; Allelein, H.J. [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energie- und Klimaforschung (IEK), Sicherheitsforschung und Reaktortechnik (IEK-6)

2011-07-01

Pellet-Cladding Interaction (PCI) is a well known effect in fuel pins. One possible reason for PCI-effects could be local power excursions in the fuel pins, which can led to a rupture of the fuel cladding tube. From a reactor safety point of view this has to be considered as a violence of the barrier principal in order to retain fission products in the fuel pins. This paper focuses on the pin power distributions in a 2D infinite lattice of a BWR fuel element. Lots of studies related PCI effect can be found in the literature. In this compact, coupled neutronic depletion calculations taking the control history effect into account are described. Depletion calculations of an infinite fuel element of a BWR were carried out with controlled, uncontrolled and temporarily controlled scenarios. Later ones are needed to describe the control blade history (CBH) effect. A Monte-Carlo approach is mandatory to simulate the neutron physics. The VESTA code was applied to couple the Monte-Carlo-Code MCNP(X) with the burnup code ORIGEN. Additionally, CASMO-4 is also employed to verify the method of simulation results from VESTA. The cross sections for Monte Carlo and burn-up calculations are derived from ENDF/B-VII.0. (orig.)

Comparative prediction of irradiation test of CNFT and Cise prototypes of CIRENE fuel pins, a prediction by transuranus M1V1J12 code

International Nuclear Information System (INIS)

Suwardi

2014-01-01

A prototype of fuel pin design for HWR by CIRENE has been realized by Center for Nuclear Fuel Technology CNFT-BATAN. The prototype will be irradiated in PRTF Power Ramp Test (PRTF). The facility has been installed inside RSG-GA Siwabessy at Serpong. The present paper reports the preparation of experimentation and prediction of irradiation test. One previous PCI test report is found in, written by Lysell G and Valli G in 1973. The CNFT fuel irradiation test parameter is adapted to both PRTF and power loop design for RSG-GAS reactor in Serpong mainly the maxima of: rod length, neutrons flux, total power of rod, and power ramp rate. The CNFT CIRENE prototype design has been reported by Futichah et al 2007 and 2010. The AEC-India HWR fuel pin is of 19/22 fuel bundle design has also been evaluated as comparison. The first PCI test prediction has experiment comparison for Cise pin. The second prediction will be used for optimizing the design of ramp test for CNFT CIRENE fuel pin prototype. (author)

The design of electrical heater pins to simulate transient dryout and post-dryout of water reactor fuel

International Nuclear Information System (INIS)

Burgess, M.H.; Butcher, A.A.; Sidoli, J.E.A.

1978-11-01

A theoretical assessment of indirect and direct filled heater simulations of nuclear reactor fuel pins is described. For reasons of fast temperature response, a direct unfilled heater, with thermocouples buried in the walls, is recommended for studies of Loss-of-Coolant Accidents leading to dryout, post-dryout and rewetting. A design of heater pins, for use in SGHWR or PWR experiments, and compatible with existing 9MW power supplies, is described. Experiments to confirm collapse pressure calculations at 1000 0 C and thermocouple response times are also reported. (author)

Passive measurements of mixed-oxide fuel for nuclear nonproliferation

International Nuclear Information System (INIS)

Dolan, Jennifer L.; Flaska, Marek; Pozzi, Sara A.; Chichester, David L.

2013-01-01

We present new results on passive measurements and simulations of mixed-oxide fuel-pin assemblies. Potential tools for mixed-oxide fuel pin characterization are discussed for future nuclear-nonproliferation applications. Four EJ-309 liquid scintillation detectors coupled with an accurate pulse timing and digital, offline and optimized pulse-shape discrimination method were used. Measurement analysis included pulse-height distributions to distinguish between purely fission neutron sources and alpha-n plus fission neutrons sources. Time-dependent cross-correlation functions were analyzed to measure the fission neutron contribution to the measured sample's neutron source. The use of Monte Carlo particle transport code MCNPX-PoliMi is discussed in conjunction with the measurements

Developments in the LASL Fuel Pin Imaging System: PINEX-3A

International Nuclear Information System (INIS)

Lumpkin, A.H.; Berzins, G.J.; Cosimi, R.A.; O'Hare, T.E.; Davidson, J.R.

1979-01-01

The LASL Fuel Pin Imaging System was evaluated using a series of 10 TREAT transients, each of approx. 240-MW peak power. HEDL provided the fuel-ejection type capsule with annular fuel pellets. The pin visibility threshold was determined to be approx. 20-MW of TREAT power (approx. 130 W/g), almost an order of magnitude improvement over our PINEX-2 threshold. The impact of changes in instrumentation, imaging apertures, and fluors that produced the improved sensitivity are reported. Results of a time-integrated imaging technique are also presented

Comparative analysis of different methods of modelling of most loaded fuel pin in transients

International Nuclear Information System (INIS)

Ovdiyenko, Y.; Khalimonchuk, V.; Ieremenko, M.

2007-01-01

Different methods of modeling of most loaded fuel pin are presented at the work. Calculation studies are performed on example of accident related to WWER-1000 cluster rod ejection with using of spatial kinetic code DYN3D that uses nodal method to calculate distribution of neutron flux in the core. Three methods of modeling of most loaded fuel pin are considered - flux reconstruction in fuel macrocell, pin-by-pin calculation by using of DYN3D/DERAB package and by introducing of additional 'hot channel'. Obtained results of performed studies could be used for development of calculation kinetic models during preparing of safety analysis report (Authors)

Analytic models for fuel pin transient performance

International Nuclear Information System (INIS)

Bard, F.E.; Fox, G.L.; Washburn, D.F.; Hanson, J.E.

1976-09-01

HEDL's ability to analyze various mechanisms that operate within a fuel pin has progressed substantially through development of codes such as PECTCLAD, which solves cladding response, and DSTRESS, which solves fuel response. The PECTCLAD results show good correlation with a variety of mechanical tests on cladding material and also demonstrate the significance of cladding strength when applying the life fraction rule. The DSTRESS results have shown that fuel deforms sufficiently during overpower transient tests that available volumes are filled, whether in the form of a central cavity or start-up cracks

Program of quality management when fabricating fast reactor vibropack oxide fuel pins

International Nuclear Information System (INIS)

Mayorshin, A.A.; Kisly, V.A.; Sudakov, L.V.

2000-01-01

There are presented main principles of creation and operation of Quality Management Program in fabricating vibropack oxide fuel pins for BOR-60 and BN-600 being in force in SSC RF RIAR. There is given structure of documentation for QS principal elements. Under Quality System there are defined all the procedures, assuring that fuel pin meets the normative requirements. The system model is complied with the standard model IS 9001. There are shown technologic flowchart and check operation, statistic results of pin critical parameter check as well as main results of in-pile tests. (author)

Method for monitoring irradiated nuclear fuel using cerenkov radiation

International Nuclear Information System (INIS)

Caldwell, J.T.; Dowdy, E.J.; Nicholson, N.

1983-01-01

A method is provided for monitoring irradiated nuclear fuel inventories located in a water-filled storage pond wherein the intensity of the cerenkov radiation emitted from the water in the vicinity of the nuclear fuel is measured. This intensity is then compared with the expected intensity for nuclear fuel having a corresponding degree of irradiation exposure and time period after removal from a reactor core. Where the nuclear fuel inventory is located in an assembly having fuel pins or rods with intervening voids, the cerenkov light intensity measurement is taken at selected bright spots corresponding to the water-filled interstices of the assembly in the water storage, the waterfilled interstices acting as cerenkov light channels so as to reduce cross-talk. On-line digital analysis of an analog video signal is possible, or video tapes may be used for later measurement using a video editor and an electrometer. Direct measurement of the cerenkov radiation intensity also is possible using spot photometers pointed at the assembly

Theoretical studies of the influence of filler material gas gap and cladding material on rewetting rate of nuclear reactor fuel pins

International Nuclear Information System (INIS)

Blackburn, D.; Pearson, K.G.; Shires, G.L.

1977-03-01

Theoretical studies of the effect of fuel and gas gap on the rewetting rate of overheated fuel pins quenched by a falling film of water are presented. Two approaches have been made: a finite difference technique and an approximate analytical solution. The results obtained by the two methods for the case of a uranium-dioxide-filled Zircaloy clad fuel pin are in close agreement. The paper shows that under high pressure conditions the delaying effect of the stored heat within the fuel on the wetting rate is relatively small, particularly if a gas gap is present between the clad and the fuel. At low pressure conditions, however, the effect of the fuel may be very important. Simplification of the analytical solution shows that at low wetting rates a constant fractional reduction in wetting speed may be anticipated the magnitude of which depends only on the relative thermal diffusivities and heat capacities of the fuel and cladding. (author)

The lumped parameter model for fuel pins

Energy Technology Data Exchange (ETDEWEB)

Liu, W S [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

The use of a lumped fuel-pin model in a thermal-hydraulic code is advantageous because of computational simplicity and efficiency. The model uses an averaging approach over the fuel cross section and makes some simplifying assumptions to describe the transient equations for the averaged fuel, fuel centerline and sheath temperatures. It is shown that by introducing a factor in the effective fuel conductivity, the analytical solution of the mean fuel temperature can be modified to simulate the effects of the flux depression in the heat generation rate and the variation in fuel thermal conductivity. The simplified analytical method used in the transient equation is presented. The accuracy of the lumped parameter model has been compared with the results from the finite difference method. (author). 4 refs., 2 tabs., 4 figs.

Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors

International Nuclear Information System (INIS)

Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan

1999-02-01

Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs

Image analysis for remote examination of fuel pins

International Nuclear Information System (INIS)

Cook, J.H.; Nayak, U.P.

1982-01-01

An image analysis system operating in the Wing 9 Hot Cell Facility at Los Alamos National Laboratory provides quantitative microstructural analyses of irradiated fuels and materials. With this system, fewer photomicrographs are required during postirradiation microstructural examination and data are available for analysis much faster. The system has been used successfully to examine Westinghouse Advanced Reactors Division experimental fuel pins

Numerical solution of diffusion equation to study fast neutrons flux distribution for variant radii of nuclear fuel pin and moderator regions

Energy Technology Data Exchange (ETDEWEB)

Mousavi Shirazi, Seyed Alireza [Islamic Azad Univ. (I.A.U.), Dept. of Physics, Tehran (Iran, Islamic Republic of)

2015-07-15

In this symbolic investigation, a cylindrical cell in a LWR, which consists of one fuel pin and moderator (water), is considered. The width of this cylindrical cell is divided into 100 equal units. Since the neutron flux in a cylindrical fuel pin is resulting from the diffusion equation: -(1)/(r)(d)/(dr)Dr(d)/(dr)φ(r) + Σ{sub a}φ(r) = S(r), the amount of fast neutron fluxes are obtained on the basis of the numeric solution of this equation, and the applied boundary conditions are considered: φ'(0) = φ'(1) = 0. This differential equation is solved by the tridiagonal method for variant enrichments of uranium. Neutron fluxes are obtained in variant radii of fuel pin and moderator and are finally compared with each other. There are some interesting outcomes resulting from this investigation. It can be inferred that because of the fuel enrichment increment, the fast neutron flux increases significantly at the centre of core, while many of the fast neutrons produced are absorbed after entering the water region, moderation of lots of them causes the reduced neutron flux to get improved in this region.

Axial migratin of cesium in LMFBR fuel pins

International Nuclear Information System (INIS)

Karnesky, R.A.; Bridges, A.E.; Jost, J.W.

1981-11-01

A correlated model for quantitatively predicting the behavior of cesium in LMFBR fuel pins has been developed. This correlation was shown to be in good agreement with experimental data. It has been used to predict the behavior of cesium in the FFTF driver fuel and as the result of this analysis it has been shown that the accumulation of cesium in the insulator pellets at the ends of the fuel column will not be life limiting

Analysis of fuel pin mechanics in case of flow blockage of a single RBMK channel

International Nuclear Information System (INIS)

Pierro, F.; Moretti, F.; Mazzini, D.; D'Auria, F.

2005-01-01

The evaluation of the consequences of the pressure tube rupture due to accidental overheating is one of the key elements for addressing an RBMK safety analysis, since it causes the lost of design boundaries against the fission products release. Several events are expected to take place: thermal hydraulic crisis (energy unbalance), fuel overheating, fuel rod damage, pressure tube overheating, pressure tube failure and graphite stack damage, Hydrogen and fission products release. The present work deals with the research activity carried out at ''Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione'' (DIMNP) of the University of Pisa aimed at assessing numerical models for safety analysis of the RBMK-1000. The attention is focused on the modelling of (1) a single fuel channel and its surrounding graphite column for evaluating the transient conditions enabling the different damaging phenomena, (2) a single fuel rod for investigating fuel pin behaviour, (3) the ruptured fuel channel for figuring the magnitude of the hydrodynamic loads acting on fuel rods. Different codes were employed to cover the competences for the investigation of each field; in particular, RELAP5 code for thermal-hydraulics, FRAPCON-3 and FRAPTRAN1-2 codes for fuel pin mechanics, FLUENT-6 for fluid dynamics. The paper discusses the numerical models, the analysis capabilities of numerical models in comparison with available data about the Leningrad NPP 1992 accident. Furthermore, the possibility to draw a failure map identifying the range of the cladding safety respect to the transient condition is outlined. (author)

Material accountancy for metallic fuel pin casting

International Nuclear Information System (INIS)

Bucher, R.G.; Orechwa, Y.; Beitel, J.C.

1995-01-01

The operation of the Fuel Conditioning Facility (FCF) is based on the electrometallurgical processing of spent metallic reactor fuel. The pin casting operation, although only one of several operations in FCF, was the first to be on-line. As such, it has served to demonstrate the material accountancy system in many of its facets. This paper details, for the operation of the pin casting process with depleted uranium, the interaction between the mass tracking system (MTG) and some of the ancillary computer codes which generate pertinent information for operations and material accountancy. It is necessary to distinguish between two types of material balance calculations -- closeout for operations and material accountancy for safeguards. The two have much in common, for example, the mass tracking system database and the calculation of an inventory difference, but, in general, are not congruent with regard to balance period and balance spatial domain. Moreover, the objective, assessment, and reporting requirements of the calculated inventory difference are very different in the two cases

The fuel to clad heat transfer coefficient in advanced MX-type fuel pins

International Nuclear Information System (INIS)

Caligara, F.; Campana, M.; Mandler, R.; Blank, H.

1979-01-01

Advanced fuels (mixed carbides, nitrides and carbonitrides) are characterised by a high thermal conductivity compared to that of oxide fuels (5 times greater) and their behaviour under irradiation (amount of swelling, fracture behaviour, restructuring) is far more sensitive to the design parameters and to the operating temperature than that of oxide fuels. The use of advanced fuels is therefore conditioned by the possibility of mastering the above phenomena, and the full exploitation of their favorable neutron characteristics depends upon a good understanding of the mutual relationships of the various parameters, which eventually affect the mechanical stability of the pin. By far the most important parameter is the radial temperature profile which controls the swelling of the fuel and the build-up of stress fields within the pin. Since the rate of fission gas swelling of these fuels is relatively large, a sufficient amount of free space has to be provided within the pin. This space originally appears as fabrication porosity and as fuel-to-clad clearance. Due to the large initial gap width and to the high fuel thermal conductivity, the range of the fuel operating temperatures is mainly determined by the fuel-to-clad heat transfer coefficient h, whose correct determination becomes one of the central points in modelling. During the many years of modelling activity in the field of oxide fuels, several theoretical models have been developed to calculate h, and a large amount of experimental data has been produced for the empirical adjustment of the parameters involved, so that the situation may be regarded as rather satisfactory. The analysis lead to the following conclusions. A quantitative comparison of experimental h-values with existing models for h requires rather sophisticated instrumented irradiation capsules, which permit the measurement of mechanical data (concerning fuel and clad) together with heat rating and temperatures. More and better well

Review of HEDL fuel pin transient analyses analytical programs

International Nuclear Information System (INIS)

Scott, J.H.; Baars, R.E.

1975-05-01

Methods for analysis of transient fuel pin performance are described, as represented by the steady-state SIEX code and the PECT series of codes used for steady-state and transient mechanical analyses. The empirical fuel failure correlation currently in use for analysis of transient overpower accidents is described. (U.S.)

Development of a Fast Breeder Reactor Fuel Bundle Deformation Analysis Code - BAMBOO: Development of a Pin Dispersion Model and Verification by the Out-of-Pile Compression Test

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ito, Masahiro; Ukai, Shigeharu

2004-01-01

To analyze the wire-wrapped fast breeder reactor fuel pin bundle deformation under bundle/duct interaction conditions, the Japan Nuclear Cycle Development Institute has developed the BAMBOO computer code. This code uses the three-dimensional beam element to calculate fuel pin bowing and cladding oval distortion as the primary deformation mechanisms in a fuel pin bundle. The pin dispersion, which is disarrangement of pins in a bundle and would occur during irradiation, was modeled in this code to evaluate its effect on bundle deformation. By applying the contact analysis method commonly used in the finite element method, this model considers the contact conditions at various axial positions as well as the nodal points and can analyze the irregular arrangement of fuel pins with the deviation of the wire configuration.The dispersion model was introduced in the BAMBOO code and verified by using the results of the out-of-pile compression test of the bundle, where the dispersion was caused by the deviation of the wire position. And the effect of the dispersion on the bundle deformation was evaluated based on the analysis results of the code

Calculation of fission gases internal pressure in nuclear fuel rods

International Nuclear Information System (INIS)

Vasconcelos Santana, M. de.

1981-12-01

Models concerning the principal phenomena, particularly thermal expansion, fuel swelling, densification, reestructuring, relocation, mechanical strain, fission gas production and release, direct or indirectly important to calculate the internal pressure in nuclear fuel rods were analysed and selected. Through these analyses a computer code was developed to calculate fuel pin internal pressure evolution. Three different models were utilized to calculate the internal pressure in order to select the best and the most conservative estimate. (Author) [pt

Influence of LMFBR fuel pin temperature profiles on corrosion rate

International Nuclear Information System (INIS)

Shiels, S.A.; Bagnall, C.; Schrock, S.L.; Orbon, S.J.

1976-01-01

The paper describes the sodium corrosion behavior of 20 percent cold worked Type 316 stainless steel fuel pin cladding under a simulated reactor thermal environment. A temperature gradient, typical of a fuel pin, was generated in a 0.9 m long heater section by direct resistance heating. Specimens were located in an isothermal test section immediately downstream of the heater. A comparison of the measured corrosion rates with available data showed an enhancement factor of between 1.5 and 2 which was attributed to the severe axial temperature gradient through the heater. Differences in structure and surface chemistry were also noted

Velocity distribution measurement in wire-spaced fuel pin bundle

International Nuclear Information System (INIS)

Mizuta, Hiroshi; Ohtake, Toshihide; Uruwashi, Shinichi; Takahashi, Keiichi

1974-01-01

Flow distribution measurement was made in the subchannels of a pin bundle in air flow. The present paper is interim because the target of this work is the decision of temperature of the pin surface in contact with wire spacers. The wire-spaced fuel pin bundle used for the experiment consists of 37 simulated fuel pins of stainless steel tubes, 3000 mm in length and 31.6 mm in diameter, which are wound spirally with 6 mm stainless steel wire. The bundle is wrapped with a hexagonal tube, 3500 mm in length and 293 mm in flat-to-flat distance. The bundle is fixed with knock-bar at the entrance of air flow in the hexagonal tube. The pitch of pins in the bundle is 37.6 mm (P/D=1.19) and the wrapping pitch of wire is 1100 mm (H/D=34.8). A pair of arrow-type 5-hole Pitot tubes are used to measure the flow velocity and the direction of air flow in the pin bundle. The measurement of flow distribution was made with the conditions of air flow rate of 0.33 m 3 /sec, air temperature of 45 0 C, and average Reynolds number of 15100 (average air velocity of 20.6 m/sec.). It was found that circular flow existed in the down stream of wire spacers, that axial flow velocity was slower in the subchannels, which contained wire spacers, than in those not affected by the wire, and that the flow angle to the axial velocity at the boundary of subchannels was two thirds smaller than wire wrapping angle. (Tai, I.)

Specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1979-12-01

The purpose of the meeting was to provide an opportunity for exchanging views of theoretical modelling of LMFBR fuel pin behaviour and to summarise the IWGFR member countries' knowledge in this field. The special emphasis was placed on normal operating conditions. The technical part of the meeting was divided into six sessions, as follows: An overview of fuel modelling studies; Key factors and basic phenomena relevant to fuel pin behaviour modelling; Application to steady state operation and normal transients; Experimental validation through pins in service and specific irradiation experiments; Advanced fuels; and Brief review of existing codes. During the meeting, papers were presented by the delegates on behalf of their countries or organization. The papers, which are included in this report, were either in the form of a general survey of the subject, or on specific technical subjects. In each subject area presentations appropriate to the subject were made from the submitted papers. The presentations were followed by discussions of the questions raised and summary is made.

Specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour. Summary report

International Nuclear Information System (INIS)

1979-12-01

The purpose of the meeting was to provide an opportunity for exchanging views of theoretical modelling of LMFBR fuel pin behaviour and to summarise the IWGFR member countries' knowledge in this field. The special emphasis was placed on normal operating conditions. The technical part of the meeting was divided into six sessions, as follows: An overview of fuel modelling studies; Key factors and basic phenomena relevant to fuel pin behaviour modelling; Application to steady state operation and normal transients; Experimental validation through pins in service and specific irradiation experiments; Advanced fuels; and Brief review of existing codes. During the meeting, papers were presented by the delegates on behalf of their countries or organization. The papers, which are included in this report, were either in the form of a general survey of the subject, or on specific technical subjects. In each subject area presentations appropriate to the subject were made from the submitted papers. The presentations were followed by discussions of the questions raised and summary is made

FFTF [Fast Flux Test Facility]/IEM [Interim Examination and Maintenance] Cell Fuel Pin Weighing System

International Nuclear Information System (INIS)

Gibbons, P.W.

1987-09-01

A Fuel Pin Weighing Machine has been developed for use in the Fast Flux Test Facility (FFTF) Interim Examination and Maintenance (IEM) Cell to assist in identifying an individual breached fuel pin from its fuel assembly pin bundle. A weighing machine, originally purchased for use in the Fuels and Materials Examination Facility (FMEF) at Hanford, was used as the basis for the IEM Cell system. Design modifications to the original equipment were centered around: 1) adapting the FMEF machine for use in the IEM Cell and 2) correcting operational deficiencies discovered during functional testing in the IEM Cell Mockup

Fast reactor fuel pin behaviour modelling in the UK

Energy Technology Data Exchange (ETDEWEB)

Matthews, J R [UKAEA, Harwell, Didcot, Oxon (United Kingdom); Hughes, H [Springfields Nuclear Power Development Laboratories, Springfields, Salwick, Preston (United Kingdom)

1979-12-01

Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

Fast reactor fuel pin behaviour modelling in the UK

International Nuclear Information System (INIS)

Matthews, J.R.; Hughes, H.

1979-01-01

Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

Energy Technology Data Exchange (ETDEWEB)

Boussard, F.; Huillery, R. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Combustibles; Averseng, J.L.; Serpantie, J.P. [Novatome Industries, 92 - Le Plessis-Robinson (France)

1994-12-31

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs.

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

International Nuclear Information System (INIS)

Boussard, F.; Huillery, R.

1994-01-01

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs

Cyclic movement pin mechanism for controlling a nuclear reactor

International Nuclear Information System (INIS)

Joly, J.G.; Martin, Jean.

1981-01-01

This invention concerns a recurring movement pin mechanism for controlling a nuclear reactor by shifting a neutron absorbing assembly, vertically mobile in the nuclear reactor, to adjust the power and for emergency shut-down. This mechanism ensures a continuous movement and accurate shut-down at any level of the travel height of the absorbing assembly in the core. It also prevents the impacts of the pivoting pins in the control rod slots [fr

Axial gap formation in P.W.R. fuel pins

International Nuclear Information System (INIS)

Roberts, G.; Jones, K.W.

1978-07-01

The potential mechanisms of axial gap formation in PWR fuel pins are examined analytically and also using evidence from post-irradiation examination (p.i.e.) investigation. It is concluded that fuel and cladding cannot remain in contact during densification and so the settling of of the fuel stack, which forms the gaps, must be prevented by such things as asperities in the cladding, fuel chips or tilted pellets. Examples from the p.i.e. examination programme are used to support this conclusion. (author)

Contamination of a PWR primary circuit by fuel pins with failed cladding

International Nuclear Information System (INIS)

Janvier, J.C.; Chagrot, M.

1979-01-01

The safety authorities in the principal nuclear countries appear to be attaching increasing importance to keeping reactor primary circuits as contamination-free as possible. Therefore, the consequences of cladding failures and especially of those resulting from fabrication defects have to be evaluated, for when these failures become systematic in nature they constitute an important source of contamination in pressurized-water reactors. The Grenoble Nuclear Research Centre is implementing a programme on the study of such failures with a view to analysing the behaviour of failed fuel elements. A distinction is made between two types of cladding failure, depending on whether the primary water enters the fuel pin as soon as the circuits are pressurized (fabrication defect) or whether the failure is caused during operation. The emission of gaseous fission products and halogens has been analysed in different operating modes (steady-state or transient), and in spite of the complexity of the phenomena involved, some results have been obtained which already enable one to evaluate fission product contamination of the primary circuit. (author)

Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

International Nuclear Information System (INIS)

Thoms, K.R.

1975-04-01

Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 (Ta-8 percent W-2 percent Hf) and uranium dioxide (UO 2 ) fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1 percent Zr. A total of nine fuel pins was irradiated (four containing porous UN, two containing dense, nonporous UN, and three containing dense UO 2 ) at average cladding temperatures ranging from 931 to 1015 0 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO 2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of 10 -5 . (U.S.)

FFTF/IEM [Fast Flux Test Facility/Interim Examination and Maintenance] cell fuel pin weighing system: Remote maintenance design considerations

International Nuclear Information System (INIS)

Gibbons, P.W.

1986-06-01

A Fuel Pin Weighing Machine has been developed for use in the Fast Flux Test Facility (FFTF) Interim Examination and Maintenance (IEM) Cell to assist in identifying an individual breached fuel pin from its fuel assembly pin bundle. Optimum configuration for remote maintenance was a major consideration in the design of each element of the Pin Weighing System

High burnup, high power irradiation behavior of helium-bonded mixed carbide fuel pins

International Nuclear Information System (INIS)

Levine, P.J.; Nayak, U.P.; Boltax, A.

1983-01-01

Large diameter (9.4 mm) helium-bonded mixed carbide fuel pins were successfully irradiated in EBR-II to high burnup (12%) at high power levels (100 kW/m) with peak cladding midwall temperatures of 550 0 C. The wire-wrapped pins were clad with 0.51-mm-thick, 20% cold-worked Type 316 stainless steel and contained hyperstoichiometric (Usub(0.8)Pusub(0.2))C fuel covering the smeared density range from 75-82% TD. Post-irradiation examinations revealed: extensive fuel-cladding mechanical interaction over the entire length of the fuel column, 35% fission gas release at 12% burnup, cladding carburization and fuel restructuring. (orig.)

Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins

International Nuclear Information System (INIS)

Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A.

1979-01-01

This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed (U,Pu)-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and (excess) liquid alkali metal (Na, K, Cs) were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate (equilibrium) phases has been analyzed. (orig./RW) [de

Device for a nuclear reactor. [Fuel element spacers

Energy Technology Data Exchange (ETDEWEB)

Foulds, R B; Kasberg, A H; Puechl, K H; Bleiberg, M L

1972-03-08

A spacer design for fuel element clusters for PWR type reactors is described. It consists of a frame supporting an egg-carton like grid each sector of which is provided with springs which grip the fuel pins. The spring design is such as to prevent fuel pin vibrations and at same time accommodate fuel pin deformations. Formulae for the calculation of natural frequencies, spring stiffness and friction loads are presented.

Fuel-cladding chemical interaction correlation for mixed-oxide fuel pins

International Nuclear Information System (INIS)

Lawrence, L.A.

1986-10-01

A revised wastage correlation was developed for FCCI with fabrication and operating parameters. The expansion of the data base to 305 data sets provided sufficient data to employ normal statistical techniques for calculation of confidence levels without unduly penalizing predictions. The correlation based on 316 SS cladding also adequately accounts for limited measured depths of interaction for fuel pins with D9 and HTq cladding

Calculational assessment of critical experiments with mixed-oxide fuel pin arrays moderated by organic solution

International Nuclear Information System (INIS)

Smolen, G.R.; Funabashi, H.

1987-01-01

Critical experiments have been conducted with organically moderated mixed-oxide (MOX) fuel pin assemblies at the Pacific Northwest Lab. Critical Mass Lab. These experiments are part of a joint exchange program between the US Dept. of Energy and the Power Reactor and Nuclear Fuel Development Corp. of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organically moderated systems. Calculations presented in this paper indicated that the Scale code system and the 27-energy-group cross-section library accurately compute k/sub eff/ for organically moderated MOX fuel pin assemblies. Furthermore, the reactivity of an organic solution with a 32 vol % TBP/68 vol% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water

Calculational assessment of critical experiments with mixed oxide fuel pin arrays moderated by organic solution

International Nuclear Information System (INIS)

Smolen, G.R.

1987-01-01

Critical experiments have been conducted with organic-moderated mixed oxide (MOX) fuel pin assemblies at the Pacific Northwest Laboratory (PNL) Critical Mass Laboratory (CML). These experiments are part of a joint exchange program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organic-moderated systems. Calculations presented in this paper indicated that the SCALE code system and the 27-energy-group cross-section accurately compute k-effectives for organic moderated MOX fuel-pin assemblies. Furthermore, the reactivity of an organic solution with a 32-vol-% TBP/68-vol-% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water. 5 refs

Design of unique pins for irradiation of higher actinides in a fast reactor

International Nuclear Information System (INIS)

Basmajian, J.A.; Birney, K.R.; Weber, E.T.; Adair, H.L.; Quinby, T.C.; Raman, S.; Butler, J.K.; Bateman, B.C.; Swanson, K.M.

1982-03-01

The actinides produced by transmutation reactions in nuclear reactor fuels are a significant factor in nuclear fuel burnup, transportation and reprocessing. Irradiation testing is a primary source of data of this type. A segmented pin design was developed which provides for incorporation of multiple specimens of actinide oxides for irradiation in the UK's Prototype Fast Reactor (PFR) at Dounreay Scotland. Results from irradiation of these pins will extend the basic neutronic and material irradiation behavior data for key actinide isotopes

Behavior of a bundle of fast fuel pins under irradiation

International Nuclear Information System (INIS)

Marbach, G.; Millet, P.; Robert, J.; Languille, A.

1979-01-01

In the French design of fuel elements for fast reactors, great deformation of pins can bring about interaction with the hexagonal tube through the spacer wires. The change in such bundles is described here when the diameter of the cladding increases and the outcome of this reaction (bending and ovalization of pins) is calculated with a simplified model. It is shown that the results achieved agree well with the experimental observations [fr

Fast breeder fuel pin bundle tests in the KNK II-reactor

International Nuclear Information System (INIS)

Haefner, H.E.; Bojarsky, E.

1986-11-01

Three variants of ring elements with test bundles will be reported in this paper: In a first step a ring element was built with a permanently integrated test bundle (19 carbide pins of the Karlsruhe reference concept) while the proven fuel element components have been largely maintained. This irradiation will be completed in autumn 1986 after 380 full power days of operation. The central topic of this paper will be the technique of reloadable ring elements with replaceable test bundles. A first experiment, TOAST, is in preparation. For this experiment, above all the components of the fuel element head and foot had to be newly developed and tested. A special version of double-walled replaceable test bundles to be used in the TETRA temperature transient experiments will be briefly mentioned. It is envisaged in these experiments to vary in a defined manner the coolant flow at remotely assembled test bundles consisting of 19 KNK pins each having undergone a high burnup and to use a measuring and control plug placed on the test bundle so that a variety of fuel pin temperature programs can be realized. Finally, some additional aspects of bundle design will be indicated. (orig./GL) [de

The deformation analysis of the KALIMER breakeven core driver fuel pin based on the axial power profile during irradiation

International Nuclear Information System (INIS)

Lee, Dong Uk; Lee, Byoung Oon; Kim, Young Kyun; Hong, Ser Gi; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

2003-03-01

In this study, material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the End Of Life(EOL) is predicted to be 68.61% and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is 1.928%, satisfying the preliminary design criterion (3%) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc

The deformation analysis of the KALIMER breakeven core driver fuel pin based on the axial power profile during irradiation

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Uk; Lee, Byoung Oon; Kim, Young Kyun; Hong, Ser Gi; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

2003-03-01

In this study, material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the End Of Life(EOL) is predicted to be 68.61% and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is 1.928%, satisfying the preliminary design criterion (3%) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc.

Results of transient overpower events on breached and unbreached fuel pins

International Nuclear Information System (INIS)

Strain, R.V.; Tsai, H.C.; Neimark, L.A.; Aratani, K.

1986-04-01

The objective of the extended overpower tests on intact pins was to determine the pin cladding breaching thresholds vis-a-vis the Plant Protection System (PPS) trip settings, typically at ∼10 to 15% overpower. These tests emphasize slow operational-type transients in light of earlier work which suggested that irradiated mixed-oxide fuel pins may be particularly vulnerable in the slow ramp-rate regime. An overview of the extended overpower test series was previously reported. More recent results on two of the tests in this series are included in this paper. These two tests, designated TOPI-1A and TOPI-1B, were each conducted on a 19-pin assembly with various pin design, operation and burnup variables. The overpower ramp rates for the TOPI-1A and -1B tests were 0.1%/s and 10%/s, respectively

Physics evaluation for testino. of RAPS and TAPS fuel pins in CIRUS pressurised water loop

International Nuclear Information System (INIS)

John, Benjamin; Paul, O.P.K.

1976-01-01

Relevant calculations carried out to assess the reactivity effect, heat generation and other parameters for testing of RAPS and TAPS fuel pins in the Cirus pressurised water loop are summarised. The Cirus neutron flux level being low, in order to simulate the RAPS design heat rating of ∫ Kdtheta = 40 w/cm, the required plutonium enrichment in mixed plutonium uranium oxide fuel pin was worked out. The results showed that a PuO 2 enrichment of 1.5 wt percent would be necessary to meet the above requirement. The analysis for the TAPS pin indicated that the desired heat flux of 115w/cm 2 cannot be obtained in the Cirus loop with either a 7 pin cluster geometry, or with a single pin with the enrichment level as used in TAPS pin. Lattice code DUMLAC and the core simulation code AECLHEX were used for these studies. (author)

FFTF fuel pin design procedure verification for transient operation

International Nuclear Information System (INIS)

Baars, R.E.

1975-05-01

The FFTF design procedures for evaluating fuel pin transient performance are briefly reviewed, and data where available are compared with design procedure predictions. Specifically, burst conditions derived from Fuel Cladding Transient Tester (FCTT) tests and from ANL loss-of-flow tests are compared with burst pressures computed using the design procedure upon which the cladding integrity limit was based. Failure times are predicted using the design procedure for evaluation of rapid reactivity insertion accidents, for five unterminated TREAT experiments in which well characterized fuel failures were deliberately incurred. (U.S.)

Irradiation of a 19 pin subassembly with mixed carbide fuel in KNK II

Science.gov (United States)

Geithoff, D.; Mühling, G.; Richter, K.

1992-06-01

The presentation deals with the fabrication, irradiation and nondestructive postirradiation examinations of LMR fuel pins with mixed (U, Pu)-carbide fuels. The mixed carbide fuel was fabricated by the European Institute of Transuranium Elements using various fabrication procedures. Fuel composition varied therefore in a wide range of tolerances with respect to oxygen and phase content and microstructure. The 19 carbide pins were irradiated in the fast neutron flux of the KNK II reactor to a burn-up of about 7 at% without any failure in the centre of a KNK "carrier element" at a maximum linear rating of 800 W/cm. After dismantling in the Hot Cells of KfK nondestructive examinations were carried out comprising dimensional controls, radiography, γ-scanning and eddy-current testing. The results indicate differences in fuel behaviour with respect to composition of the fuel.

Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

International Nuclear Information System (INIS)

Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

2014-01-01

A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

Energy Technology Data Exchange (ETDEWEB)

Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

Pressure and Temperature of the Room 1 for the Pipe Break Accidents of the 3-Pin Fuel Test Loop

Energy Technology Data Exchange (ETDEWEB)

Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

2005-08-15

This report deals with the prediction of the pressure and temperature of the room 1 for the pipe break accidents of the 3-pin fuel test loop. The 3-pin fuel test loop is an experimental facility for nuclear fuel tests at the operation conditions similar to those of PWR and CANDU power plants. Because the most processing systems of the 3-pin fuel test loop are placed in the room 1. The structural integrity of the room 1 should be evaluated for the postulated accident conditions. Therefore the pressures and temperatures of the room 1 needed for the structural integrity evaluation have been calculated by using MARS code. The pressures and temperatures of the room 1 have been calculated in various conditions such as the thermal hydraulic operation parameters, the locations of pipe break, and the thermal properties of the room 1 wall. It is assumed that the pipe break accident occurs in the letdown operation without regeneration, because the mass and energy release to the room 1 is expected to be the largest. As a result of the calculations the maximum pressure and temperature are predicted to be 208kPa and 369.2K(96.0 .deg. C) in case the heat transfer is considered in the room 1 wall. However the pressure and temperature are asymptotically 243kPa and 378.1K(104.9 .deg. C) assuming that the heat transfer does not occur in the room 1 wall.

Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

Energy Technology Data Exchange (ETDEWEB)

Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ito, Masahiro; Maeda, Koji [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

2011-09-30

Highlights: > We evaluated diametral strain of fast reactor MOX fuel pins irradiated to 130 GWd/t. > The strain was due to cladding void swelling and irradiation creep. > The irradiation creep was caused by internal gas pressure and PCMI. > The PCMI was associated with pellet swelling by rim structure or by cesium uranate. > The latter effect tended to increase the cumulative damage fraction of the cladding. - Abstract: The C3M irradiation test, which was conducted in the experimental fast reactor, 'Joyo', demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, 'Monju'. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and {sup 137}Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

Analysis of the porosity distribution of mixed oxide pins

International Nuclear Information System (INIS)

Lieblich, M.; Lopez, J.

1987-01-01

In the frame of the Joint Irradiation Program IVO-FR2-Vg7 between the Centre of Nuclear Research of Karlsruhe (KfK), the irradiation of 30 mixed-oxide fuel rods in the FR2 experimental reactor was carried out. The pins were located in 10 single-walled NaK capsules. The behaviour of the fuel during its burnup was studied, mainly, the rest-porosity and cracking distribution in the pellet, partial densification, etc. In this work 3 pins from the capsule No. 165 were analyzed. The experimental results (pore and cracking profiles) were interpreted by the fuel rod code SATURN. (Author) 20 refs

Fuel sub-assembly

International Nuclear Information System (INIS)

Jolly, R.

1988-01-01

A nuclear fuel sub-assembly includes a hexagonal bundle of parallel, spaced apart fuel pins coupled at one end to an end-holding grid comprising a number of transverse spaced apart rails to each of which is connected a series of pin-receiving cells which render the pins axially captive with the rails. The series of cells are defined by a pair of metal strips each of which has a series of pocket formations such that when the pocket formations are in registry they define cylindrical shaped cells provided with internal projections which engage annular recesses in the end caps of the fuel pins to effect axial constraint of the pins. (author)

Comparison of reconstructed radial pin total fission rates with experimental results in full scale BWR fuel elements

Energy Technology Data Exchange (ETDEWEB)

Giust, Flavio [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne (Switzerland); Nordostschweizerische Kraftwerke AG, Parkstrasse 23, CH-5401 Baden (Switzerland); Grimm, Peter; Jatuff, Fabian [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne (Switzerland)

2008-07-01

Total fission rate measurements have been performed on full size BWR fuel assemblies of type SVEA-96+ in the zero power reactor PROTEUS at the Paul Scherrer Institute. This work presents comparisons of reconstructed 2D pin fission rates in two configurations, I-1A and I-2A. Both configurations contain, in the central test zone, an array of 3x3 SVEA-96+ fuel elements moderated with light water at 20 deg. C. In configuration I-2A, an L-shaped hafnium control blade (half of a real cruciform blade) is inserted adjacent to the NW corner of the central fuel element. To minimize the impact of the surroundings, all measurements were done in fuel pins belonging to the central assembly. The 3x3 experimental configuration was modeled using the core monitoring and design tools that are applied at the Leibstadt Nuclear Power Plant (KKL). These are the 2D transport code HELIOS, used for the cross-section generation, and the 3D, 2-group nodal diffusion code PRESTO-2. The exterior is represented, in the axial and radial directions, by 2-group albedos calculated at the test zone boundary using a full-core 3D MCNPX model. The calculated-to-experimental (C/E) ratios of the total fission rates have a standard deviation of 1.3% in configuration I-1A (uncontrolled) and 3.2% in configuration I-2A (controlled). Sensitivity cases are analyzed to show the impact of certain parameters on the calculated fission rate distribution and reactivity. It is shown that the relative pin fission rate is only weakly dependent on these parameters. In cases without a control blade, the pin power reconstruction methodology delivers the same level of accuracy as 2D transport calculations. On the other hand, significant deviations, that are inherent to the use of reflected geometry in the lattice calculations, are observed in cases when the control blade is inserted. (authors)

Post-irradiation examination of fifteen UO2/PuO2-fuel pins from the experiment DFR-350

International Nuclear Information System (INIS)

Geithoff, D.

1975-06-01

Within the framework of the fuel pin development for a sodium-cooled fast reactor a subassembly containing 77 fuel pins has been irradiated up to 5.65% fima in the Dounreay fast reactor. The pins were prototypes in terms of fuel and cladding material. The fuel consisted of mechanically mixed UO 2 (80%) and PuO 2 (20%) pressed into pellets whereas austenitic steels (W.-No. 1,4961 and 1,4988) were used as cladding material. Furthermore a blanket column of UO 2 pellets and a gas plenum were incorporated in the pin. For irradiation the conditions in a fast breeder were simulated by a linear rod power of 450 W/cm and a maximum cladding temperature of 630 0 C. After the successful completion of the irradiation, the subassembly was dismantled and fifteen pins were selected for a nondestructive and destructive examination. The tests included visual control, measurement of external dimensions, γ-spectroscopy, X-ray radiography, fission gas measurement, ceramography, radiochemical burn-up measurement. The results are presented. The most important results of the examinations seem to be the migration of fission product cesium and the fact that no signs of impending pin failure have been found. Thus the pin specification tested in this experiment is capable of achieving higher burnups under the irradiation conditions described above. (orig./AK) [de

Measurement of fission gas release, internal pressure and cladding creep rate in the fuel pins of PHWR bundle of normal discharge burnup

Energy Technology Data Exchange (ETDEWEB)

Viswanathan, U.K. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Sah, D.N., E-mail: dnsah@barc.gov.i [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Rath, B.N.; Anantharaman, S. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

2009-08-01

Fuel pins of a Pressurised Heavy Water Reactor (PHWR) fuel bundle discharged from Narora Atomic Power Station unit no. 1 after attaining a fuel burnup of 7528 MWd/tU have been subjected to two types of studies, namely (i) puncture test to estimate extent of fission gas release and internal pressure in the fuel pin and (ii) localized heating of the irradiated fuel pin to measure the creep rate of the cladding in temperature range 800 deg. C - 900 deg. C. The fission gas release in the fuel pins from the outer ring of the bundle was found to be about 8%. However, only marginal release was found in fuel pins from the middle ring and the central fuel pin. The internal gas pressure in the outer fuel pin was measured to be 0.55 +- 0.05 MPa at room temperature. In-cell isothermal heating of a small portion of the outer fuel pins was carried out at 800 deg. C, 850 deg. C and 900 deg. C for 10 min and the increase in diameter of the fuel pin was measured after heat treatment. Creep rates of the cladding obtained from the measurement of the diameter change of the cladding due to heating at 800 deg. C, 850 deg. C and 900 deg. C were found respectively to be 2.4 x 10{sup -5} s{sup -1}, 24.6 x 10{sup -5} s{sup -1} and 45.6 x 10{sup -5} s{sup -1}.

Computational and experimental analysis of causes for local deformation of research reactor U-Mo fuel pin claddings in case of high burn-ups

International Nuclear Information System (INIS)

Popov, V.V.; Khmelevsky, M.Ya.; Lukichev, V.A.; Golosov, O.A.

2005-01-01

Post-reactor investigations of (U-Mo) fuel pins irradiated in the IVV-2M reactor have allowed to determine: the change in a fuel pin volume; the dimensions and the kind of the local deformation of fuel pin claddings; the amount of gases released under the cladding from the fuel composition, the thickness and appearance of the interaction layer of between the (U-Mo) particles and aluminium as a matrix material. The computational analysis of the stressed-strained state of fuel pins has shown that the major contribution to the increase of the fuel pin volume is made by the fuel swelling caused by the solid products of fission being formed in the process of operation. The emergence of the (U-Mo) fuel-aluminium matrix interaction layers around the (U-Mo) particles results in formation and evolution of lamination cavities inside the fuel composition under the joint action of the pressure of process gases and gaseous fission products. In case of high burn-up a local bulge of a fuel pin cladding is being formed in the fuel lamination area caused by the pressure of gases in the presence of creep in the fuel pin cladding material. The computational results relating to the local strain in a research reactor (U-Mo) fuel pin are in a good accordance with the results of the post-reactor investigations. (author)

Report of 5th new nuclear fuel research meeting, Yayoi Research Group. Trend of advanced basic research in nuclear fuel technical development

International Nuclear Information System (INIS)

1994-03-01

Theme of this meeting is 'Trend of advanced basic research in nuclear fuel technical development', and it was attempted to balance both sides of the basic research and the development. At the meeting, lectures were given on the chemical form of FPs in oxide fuel pins, the absorption of hydrogen of fuel cladding tubes, the application of hydride fuel to thorium cycle, the thermal properties of fuel cladding tubes, the preparation of NpN and heat conductivity, the high temperature chemical reprocessing of nitride fuel, the research on the annihilation treatment of minor actinide in fast reactors, the separation of TRU by dry process and the annihilation using a metallic fuel FBR. In this report, the summaries of the lectures are collected, and also the program of the meeting and the list of attendants are shown. (K.I.)

Preparation of a thermal-hydraulic design method for driver core fuel pins of a new in-pile experimental reactor for FBR safety research

International Nuclear Information System (INIS)

Mizuno, Masahiro; Yamaguchi, Katsuhisa; Uto, Nariaki

1999-07-01

A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under quasi-steady state and various transient operation modes. In order to evaluate the driver core performance in conducting such experiments, clarify the relating design issues to be resolved and refine the experimental needs, it is indispensable to comprehend the allowable margin for the thermal-hydraulic fuel pin design since it largely affects the strategy for the driver core design. This report presents a thermal-hydraulic design method for the driver core fuel pins, which is a combination of a two-dimensional time-dependent heat transfer analysis code TAC-2D and a general non-linear finite-element structural analysis code FINAS. In TAC-2D, the allowable spatial mesh and the time step sizes are evaluated. The code is modified so as to treat time-dependent thermal properties, include an improved gap heat-transfer model and treat the change of intra-pin gap width under transient modes, for the purpose of improving the accuracy of evaluating heat transfer characteristics which gives a significant impact on the thermal-hydraulic design. As for FINAS, the number of element nodes and spatial meshes required to obtain adequate accuracy for the thermal stress characteristics of a fuel pellet during transient modes are investigated. In addition, post-processing tools are newly developed to process the calculation results obtained from these codes. The results of this work contribute to advancing the fuel pin design study for SERAPH as well with the investigation on the technique of manufacturing fuel pins. (author)

Establishment of technological basis for fabrication of U-Pu-Zr ternary alloy fuel pins for irradiation tests in Japan

International Nuclear Information System (INIS)

Kikuchi, Hironobu; Iwai, Takashi; Nakajima, Kunihisa; Arai, Yasuo; Nakamura, Kinya; Ogata, Takanari

2011-01-01

A high-purity Ar gas atmosphere glove box accommodating injection casting and sodium-bonding apparatuses was newly installed in the Plutonium Fuel Research Facility of Oarai Research and Development Center, Japan Atomic Energy Agency, in which several nitride and carbide fuel pins were fabricated for irradiation tests. The experiences led to the establishment of the technological basis of the fabrication of U-Pu-Zr alloy fuel pins for the first time in Japan. After the injection casting of the U-Pu-Zr alloy, the metallic fuel pins were fabricated by welding upper and lower end plugs with cladding tubes of ferritic-martensitic steel. Subsequent to the sodium bonding for filling the annular gap region between the U-Pu-Zr alloy and the cladding tube with the melted sodium, the fuel pins for irradiation tests are inspected. This paper shows the apparatuses and the technological basis for the fabrication of U-Pu-Zr alloy fuel pins for the irradiation test planned at the experimental fast test reactor Joyo. (author)

COMETHE III J a computer code for predicting mechanical and thermal behaviour of a fuel pin

International Nuclear Information System (INIS)

Verbeek, P.; Hoppe, N.

1976-01-01

The design of fuel pins for power reactors requires a realistic evaluation of their thermal and mechanical performances throughout their irradiation life. This evaluation involves the knowledge of a number of parameters, very intricate and interconnected, for example, the temperature, the restructuring and the swelling rates of the fuel pellets, the dimensions, the stresses and the strains in the clad, the composition and the properties of gases, the inner gas pressure etc. This complex problem can only be properly handled by a computer programme which analyses the fuel pin thermal and mechanical behaviour at successive steps of its irradiation life. This report presents an overall description of the COMETHE III-J computer programme, designed to calculate the integral performance of oxide fuel pins with cylindrical metallic cladding irradiated in thermal or fast flux. (author)

Fuel pin behaviour under conditions of control rod withdrawal accident in CABRI-2 experiments

International Nuclear Information System (INIS)

Papin, Joelle; Lemoine, Francette; Sato, Ikken; Struwe, Dankward; Pfrang, Werner

1994-01-01

Simulation of the control rod withdrawal accident has been performed in the international CABRI-2 experimental programme. The tests realized with industrial pins led to clarification of the influence of the pellet design and have shown the important role of fission products on the solid fuel swelling which promotes early pin failure with solid fuel pellet. With annular pellet design, large fuel swelling combined to low smear density leads to degradation of fuel thermal conductivity and thus reduces power to melt. However, the high margin to deterministic failure is confirmed with hollow pellets. Improvements of the modelling were necessary to describe such behaviours in computer codes as SAS-4A, PAPAS-2S and PHYSURAC. (author)

UO2-PuO2 fuel pin capsule-irradiations of the test series FR 2-5a

International Nuclear Information System (INIS)

Dienst, W.; Goetzmann, O.; Schulz, B.

1975-06-01

In the capsule-irradiation test series FR 2-5a, short UO 2 -PuO 2 fuel pins (80 mm fuel length) of 7 mm diameter were irradiated in a thermal neutron flux at mean rod powers of 400 - 450 W/cm and mean cladding surface temperatures of 500 - 550 0 C to burnups of 0.6, 1.8 and 5.0 at% (U + Pu). Void volume redistribution in the fuel pins was examined in micrographs of cross-sections by measuring crack widths, central void diameters, and fuel porosity. The width of the radial cracks at the outer fuel rim was taken as a basis for measuring the irradiation-induced densification of the UO 2 -PuO 2 fuel. The result was that the final fuel density after irradiation-induced densification amounted to 92 - 94% TD and had already been reached after 0.6 at% burnup. The porosity measurement on fuel cross-sections was to show a possible dependence of the radial porosity redistribution on the initial sintered density. Examining the fuel pin diameters after irradiation showed permanent cladding strains after 5 at% burnup, which must be due to mechanical interaction with the fuel. To judge if the chemical compatibility between the fuel and the cladding of Cr-Ni-stainless steel 1.4988, the depths of chemical attack on the cladding inside was measured by micrographs of fuel pin cross-sections. (orig./GSC) [de

Development of a FBR fuel pin bundle deformation analysis code 'BAMBOO' . Development of a dispersion model and its validation

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ukai, Shigeharu; Asaga, Takeo

2002-03-01

Bundle Duct Interaction (BDI) is one of the life limiting factors of a FBR fuel subassembly. Under the BDI condition, the fuel pin dispersion would occur mainly by the deviation of the wire position due to the irradiation. In this study the effect of the dispersion on the bundle deformation was evaluated by using the BAMBOO code and following results were obtained. (1) A new contact analysis model was introduced in BAMBOO code. This model considers the contact condition at the axial position other than the nodal point of the beam element that composes the fuel pin. This improvement made it possible in the bundle deformation analysis to cause fuel pin dispersion due to the deviations of the wire position. (2) This model was validated with the results of the out-of-pile compression test with the wire deviation. The calculated pin-to-duct and pin-to-pin clearances with the dispersion model almost agreed with the test results. Therefore it was confirmed that the BAMBOO code reasonably predicts the bundle deformation with the dispersion. (3) In the dispersion bundle the pin-to-pin clearances widely scattered. And the minimum pin-to-duct clearance increased or decreased depending on the dispersion condition compared to the no-dispersion bundle. This result suggests the possibility that the considerable dispersion would affect the thermal integrity of the bundle. (author)

Fuel sub-assembly

International Nuclear Information System (INIS)

Jolly, R.

1982-01-01

A fuel sub-assembly for a liquid metal cooled nuclear reactor is described in which the bundle of fuel pins are braced apart by a series of spaced grids. The grids at the lower end are capable of yielding, thus allowing pins swollen by irradiation to be withdrawn with a reduced risk of damage. (U.K.)

Mechanical behavior of fast reactor fuel pin cladding subjected to simulated overpower transients

International Nuclear Information System (INIS)

Johnson, G.D.; Hunter, C.W.

1978-06-01

Cladding mechanical property data for analysis and prediction of fuel pin transient behavior were obtained under experimental conditions in which the temperature ramps of reactor transients were simulated. All cladding specimens were 20% CW Type 316 stainless steel and were cut from EBR-II irradiated fuel pins. It was determined that irradiation degraded the cladding ductility and failure strength. Specimens that had been adjacent to the fuel exhibited the poorest properties. Correlations were developed to describe the effect of neutron fluence on the mechanical behavior of the cladding. Metallographic examinations were conducted to characterize the failure mode and to establish the nature of internal and external surface corrosion. Various mechanisms for the fuel adjacency effect were examined and results for helium concentration profiles were presented. Results from the simulated transient tests were compared with TREAT test results

Development of multi-dimensional thermal-hydraulic modeling using mixing factors for wire wrapped fuel pin bundles in fast reactors. Validation through a sodium experiment of 169-pin fuel subassembly

International Nuclear Information System (INIS)

Nishimura, M.; Kamide, H.; Miyake, Y.

1997-04-01

Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature. The prediction of temperature distribution in the subassembly is, therefore one of the important issues for the reactor safety assessment. Mixing factors were applied to multi-dimensional thermal-hydraulic code AQUA to enhance the predictive capability of simulating maximum cladding temperature in the fuel subassemblies. In the previous studies, this analytical method had been validated through the calculations of the sodium experiments using driver subassembly test rig PLANDTL-DHX with 37-pin bundle and blanket subassembly test rig CCTL-CFR with 61-pin bundle. The error of the analyses were comparable to the error of instrumentation's. Thus the modeling was capable of predicting thermal-hydraulic field in the middle scale subassemblies. Before the application to large scale real subassemblies with more than 217 pins, accuracy of the analytical method have to be inspected through calculations of sodium tests in a large scale pin bundle. Therefore, computations were performed on sodium experiments in the relatively large 169-pin subassembly which had heater pins sparsely within the bundle. The analysis succeeded to predict the experimental temperature distributions. The errors of temperature rise from inlet to maximum values were reduced to half magnitudes by using mixing factors, compared to those of analyses without mixing factors. Thus the modeling is capable of predicting the large scale real subassemblies. (author)

Development of a coupling scheme between MCNP5 and subchanflow for the PIN- and fuel Assembly-Wise simulation of LWR and innovative reactors

International Nuclear Information System (INIS)

Ivanov, A.; Sanchez, V.; Imke, U.

2011-01-01

In order to increase the accuracy and the degree of spatial resolution of core design studies, coupled 3D neutronic (deterministic and Monte Carlo) and 3D thermal hydraulics (CFD and subchannel) codes are being developed worldwide. At KIT both deterministic and Monte Carlo codes were coupled with subchannel codes and applied to predict the safety-related design parameters such as pin power, maximal cladding and fuel temperature, DNB. These coupling approaches were revised and improved based on the experience gained. One particular example is replacing COBRA-TF with SUBCHANFLOW, in-house development subchannel code, in the COBRA-TF/MCNP coupling, accompanied with new way of radial mapping between the neutronic and thermal hydraulic domains. The new coupled system MCNP5/SUBCHANFLOW makes it possible to investigate variety of fuel assembly types (BWR, PWR or SCFR). Key issues in such a coupled system are the way in which thermal-hydraulic/neutronic feedbacks, accuracy of the Monte Carlo solutions and observation of convergence during the iterative solution are handled. Another key issue that might be considered is the optimal application of parallel computing. Using multi-processor computer architectures, it is possible to reduce the Monte- Carlo running time and obtain converged results within reasonable time limit. In particular it is shown that by exploiting the capabilities of multi-processor calculation, it is possible to investigate large fuel assemblies in a pin-by-pin manner with a resolution at pin and subchannel level. One of the most important issues addressed in the current work is the temperature effects on nuclear data. For the particular studies pseudo material approach was used, which produces interpolated results for Doppler broadened cross sections from NJOY pre-generated nuclear data. (author)

Position-dependency of Fuel Pin Homogenization in a Pressurized Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Heo, Woong; Kim, Yonghee [Korea Advanced Institute of Science and Technolgy, Daejeon (Korea, Republic of)

2016-05-15

By considering the multi-physics effects more comprehensively, it is possible to acquire precise local parameters which can result in a more accurate core design and safety assessment. A conventional approach of the multi-physics neutronics calculation for the pressurized water reactor (PWR) is to apply nodal methods. Since the nodal methods are basically based on the use of assembly-wise homogenized parameters, additional pin power reconstruction processes are necessary to obtain local power information. In the past, pin-by-pin core calculation was impractical due to the limited computational hardware capability. With the rapid advancement of computer technology, it is now perhaps quite practical to perform the direct pin-by-pin core calculation. As such, fully heterogeneous transport solvers based on both stochastic and deterministic methods have been developed for the acquisition of exact local parameters. However, the 3-D transport reactor analysis is still challenging because of the very high computational requirement. Position-dependency of the fuel pin homogenized cross sections in a small PWR core has been quantified via comparison of infinite FA and 2-D whole core calculations with the use of high-fidelity MC simulations. It is found that the pin environmental affect is especially obvious in FAs bordering the baffle reflector regions. It is also noted that the downscattering cross section is rather sensitive to the spectrum changes of the pins. It is expected that the pinwise homogenized cross sections need to be corrected somehow for accurate pin-by-pin core calculations in the peripheral region of the reactor core.

Course of pin fuel test In WWR-M reactor core

International Nuclear Information System (INIS)

Zakharov, A.S.; Kirsanov, G.A.; Konoplev, K.A.

2005-01-01

Pin type fuel element (FE) of square form with twisted ribs was developed in VNIINM as an alternative for tube type FE of research reactors. Two variants of full-scale fuel assemblies (FA) are under test in the core of PNPI WWR-M reactor. One FA contains FE with UO 2 LEU and other - UMo LEU. Both types of FE have an aluminum matrix. Results of the first stages of the test are presented. (author)

A three-dimensional pin-wise analysis for CEA ejection accident

Energy Technology Data Exchange (ETDEWEB)

Park, Guen-Tae; Park, Min-Ho; Park, Jin-Woo; Um, Kil-Sup; Choi, Tong-Soo [KEPCO NF, Daejeon (Korea, Republic of)

2016-10-15

The ejection of a control element assembly (CEA) with high reactivity worth causes the sudden insertion of reactivity into the core. Immediately after the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the doppler effect becomes important and turns the reactivity balance and power down to lower levels. The 3-D CEA ejection analysis methodology has been developed using the multi-dimensional code coupling system, CHASER, which couples three dimensional core neutron kinetics code ASTRA, subchannel analysis code THALES, and fuel performance analysis code FROST using message passing interface (MPI). This paper presents the pin-by-pin level analysis result with the 3-D CEA ejection analysis methodology using the CHASER. The pin-by-pin level analysis consists of DNBR, enthalpy and Pellet/Clad Mechanical Interaction (PCMI) analysis. All the evaluations are simulated for APR1400 plant loaded with PLUS7 fuel. In this paper, the pin-by-pin analysis using the multidimensional core transient code, CHASER, is presented with respect to enthalpy, DNBR and PCMI for APR1400 plant loaded with PLUS7 fuel. For the pin-by-pin enthalpy and DNBR analysis, the quarter core for HFP case or 15 - 20 assemblies around the most severe assembly for part powers or HZP cases are selected. And PCMI calculation is performed for all the rods in the whole core during a conservative time period. The pin-by-pin analysis results show that the regulatory guidelines of CEA ejection accident are satisfied.

Sensitivity analysis of fuel pin failure performance under slow-ramp type transient overpower condition by using a fuel performance analysis code FEMAXI-FBR

International Nuclear Information System (INIS)

Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

2012-01-01

The FEMAXI-FBR is a fuel performance analysis code and has been developed as one module of core disruptive evaluation system, the ASTERIA-FBR. The FEMAXI-FBR has reproduced the failure pin behavior during slow transient overpower. The axial location of pin failure affects the power and reactivity behavior during core disruptive accident, and failure model of which pin failure occurs at upper part of pin is used by reflecting the results of the CABRI-2 test. By using the FEMAXI-FBR, sensitivity analysis of uncertainty of design parameters such as irradiation conditions and fuel fabrication tolerances was performed to clarify the effect on axial location of pin failure during slow transient overpower. The sensitivity analysis showed that the uncertainty of design parameters does not affect the failure location. It suggests that the failure model with which locations of failure occur at upper part of pin can be adopted for core disruptive calculation by taking into consideration of design uncertainties. (author)

Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO₂ Fuel Pin Lattices in Air

Energy Technology Data Exchange (ETDEWEB)

Trumbull, TH [Rensselaer Polytechnic Inst., Troy, NY (United States)

2004-10-18

The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called “channeling” effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute’s Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice.Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins

High dose stainless steel swelling data on interior and peripheral oxide fuel pins

International Nuclear Information System (INIS)

Boltax, A.; Foster, J.P.; Nayak, U.P.

1983-01-01

High dose (2 x 10 23 n/cm 2 , E > 0.1 Mev) swelling data obtained on 20% cold-worked AISI 316 stainless steel (N-lot) cladding from mixed-oxide fuel pins show large differences in swelling incubation dose due to pre-incubation dose temperature changes. Circumferential swelling variations of 1.5 to 4 times were found in peripheral fuel pin cladding which experienced 30 to 60 deg C temperature changes due to movement in a temperature gradient. Consideration is given to the implications of these results to low swelling materials development and core design. (author)

Neutron radiography for quality assurance of PHWR fuel pins

International Nuclear Information System (INIS)

Chandrasekharan, K.N.; Patil, B.P.; Ghosh, J.K.; Ganguly, C.

1993-01-01

Neutron radiography was employed for quality assurance (QA) for advanced PHWR experimental fuel pins containing mixed uranium-plutonium dioxide and thorium-plutonium dioxide pellets. Direct, transfer and track-etch techniques were utilised. The thermal neutron beam facility of APSARA research reactor at Bhabha Atomic Research Centre was used. (author). 5 refs., 16 figs., 2 tabs

Spent-fuel disassembly and canning programs at the Barnwell Nuclear Fuel Plant (BNFP)

International Nuclear Information System (INIS)

Townes, G.A.

1980-02-01

Doubling pool storage and tripling truck cask shipping capability appear feasible based on preliminary development work performed at the BNFP. This would be accomplished by disassembly of the fuel assembly and canning of the fuel pins. The dry, remote disassembly and canning in the headend cells offer adequate rates, easily controlled operator exposure, and visual inspection of the fuel pins and operations through shielded windows

Application of the pulsed magnetic welding process to nuclear breeder reactor fuel pin end closures

International Nuclear Information System (INIS)

Brown, W.F.

1984-01-01

The pulsed magnetic welding process is a solid state welding process in which metallurgical bonding is effected by impacting metal or alloy parts against each other at high velocity by use of controlled high frequency, high intensity pulsed magnetic fields. This process is similar to the explosive welding process except that magnetic energy is used for impacting the parts together instead of using explosive energy. The pulsed magnetic welding (PMW) process is readily applied to the welding of cylindrical plugs to small diameter tubes. Although breeder reactor fuel pin design may vary in size, the application described here consisted of cladding tubes approximately 6.4 mm in diameter by 244 cm long with a wall thickness of 0.38 mm. After the cladding tubes are filled with fuel pellets and associated metal hardware, tapered end plugs are inserted into the end of the tubes and welded. A typical setup for PMW is described

Nuclear fuels

International Nuclear Information System (INIS)

Gangwani, Saloni; Chakrabortty, Sumita

2011-01-01

Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

International experience with the bundle behavior of fuel elements of sodium cooled reactors; derivation of a figure of merit for the judgement of fuel pin bundle parameters with respect to abrasion due to thermoelastic pin-pin interaction

International Nuclear Information System (INIS)

Toebbe, H.

1987-10-01

The report describes the status of experience with respect to the abrasion behavior of bundles in standard fuel elements and test elements with wire or grid spacing in the reactors Rapsodie fortissimo, Phenix, DFR, PFR, EBR-II, FFTF, JOYO and KNK II. With the help of simple considerations concerning thermoelastic pin-pin interactions a figure of merit is deduced from the different bundle parameters, which allows a comparative judgement of the parameters of different bundle concepts [de

Defect pin behaviour in the DFR

International Nuclear Information System (INIS)

Sloss, W.M.; Bagley, K.Q.; Edmonds, E.; Potter, P.E.

1979-01-01

A program of defective fuel pin irradiations has been carried out in the DFR. This program employed fuel pins which had failed during previous irradiations (natural defects) and pins in which simulated failures (artificial defects) had been induced prior to irradiation or during an intermediate examination stage at moderate or substantial burnups. The artificial defects simulated longitudinal ruptures and were normally located at positions near the top, middle and bottom of the pin where clad temperatures were 450, 540 and 630 0 C respectively. The fuel was mixed U-Pu oxide, and fuel form, stoichiometry, clad type, pin diameter, linear rating, and burnup were among the variables examined. The defect pin tests were normally carried out in single pin or trefoil type vehicles. After irradiation all the pins were subjected to the normal nondestructive examination procedures and the visual, radiographic, gamma-scanning, and dimensional change results are presented. Several pins were destructively examined and the metallographic data are discussed

Fuel-sodium reaction product formation in breached mixed-oxide fuel

International Nuclear Information System (INIS)

Bottcher, J.H.; Lambert, J.D.B.; Strain, R.V.; Ukai, S.; Shibahara, S.

1988-01-01

The run-beyond-cladding-breach (RBCB) operation of mixed-oxide LMR fuel pins has been studied for six years in the Experimental Breeder Reactor-II (EBR-II) as part of a joint program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan. The formation of fuel-sodium reaction product (FSRP), Na 3 MO 4 , where M = U/sub 1-y/Pu/sub y/, in the outer fuel regions is the major phenomenon governing RBCB behavior. It increases fuel volume, decreases fuel stoichiometry, modifies fission-product distributions, and alters thermal performance of a pin. This paper describes the morphology of Na 3 MO 4 observed in 5.84-mm diameter pins covering a variety of conditions and RBCB times up to 150 EFPD's. 8 refs., 1 fig

Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

Energy Technology Data Exchange (ETDEWEB)

Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

International Nuclear Information System (INIS)

Unal, Cetin; Pasamehmetoglu, Kemal; Carmack, Jon

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R and D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

Mechanical energy release in CABRI-2 experiments with Viggen-4 fuel pins

International Nuclear Information System (INIS)

Wolff, J.

1993-07-01

The results of mechanical energy release evaluations in CABRI-2 experiments with Viggen-4 fuel pins (12 atom % burnup) are described. In general the experience gained by the CABRI-1 experiments is confirmed. Those physical phenomena are enhanced which are influenced by the release of fission products. Especially the late blow-out of pressurized fission gases from the lower test pin plenum led to large flow variations. The corresponding mechanical power releases are low

Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code 'FEMAXI-FBR'

International Nuclear Information System (INIS)

Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

2012-01-01

FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0 /s (P 0 : steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)

The Effect of Material Homogenization in Calculating the Gamma-Ray dose from Spent PWR Fuel Pins in an Air Medium

International Nuclear Information System (INIS)

TH Trumbull

2005-01-01

The effect of material homogenization on the calculated dose rate was studied for several arrangements of typical PWR spent fuel pins in an air medium using the Monte Carlo code, MCNP. The models analyzed increased in geometric complexity, beginning with a single fuel pin, progressing to ''small'' lattices, i.e., 3x3, 5x5, 7x7 fuel pins, and culminating with a full 17x17 pin PWR bundle analysis. The fuel pin dimensions and compositions were taken directly from a previous study and efforts were made to parallel this study by specifying identical flux-to-dose functions and gamma-ray source spectra. The analysis shows two competing components to the overall effect of material homogenization on calculated dose rate. Homogenization of pin lattices tends to lower the effect of radiation ''channeling'' but increase the effect of ''source redistribution.'' Depending on the size of the lattice and location of the detectors, the net effect of material homogenization on dose rate can be insignificant or range from a 6% decrease to a 35% increase relative to the detailed geometry model

Fuel assembly for a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Ferrari, H M; Miller, D L; Tong, L S

1973-09-06

The subject of the patent is a spacer design applicable, primarily, to LWR, and especially, though not specifically PWR, fuel assemblies. The spacer consists of an egg-box type of assembly formed of interlocking pressed plates giving a square lattice whose openings accommodate fuel pins or regulating rods. The pressed plates are formed to provide pressed-out spring-like flanges which hold the fuel pins in position and guide the regulating rods. Additional pressed-out flanges ensure the correct configuration of the spacer structure. The spacer is designed to present as little resistance as possible to coolant flow.

Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

International Nuclear Information System (INIS)

Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye

2013-01-01

It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

Energy Technology Data Exchange (ETDEWEB)

Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

Microstructure of irradiated Inconel 706 fuel pin cladding

International Nuclear Information System (INIS)

Yang, W.J.S.; Makenas, B.J.

1983-08-01

A fuel pin from the HEDL-P-60 experiment with a cladding of solution-annealed Inconel 706 breached in an apparently brittle manner at a position 12.7 cm above the bottom of the fuel column with a crack of 5.72 cm in length after 5.0 atomic percent burnup in EBR-II. Temperatures (time-averaged midwall) and fast fluences for the fractured area range from 447 0 C and 5.5 x 10 22 n/cm 2 to 526 0 C and 6.1 x 10 22 n/cm 2 (E > 0.1 MeV). Specimens of the fractured fuel pin section were successfully prepared and examined in both a scanning electron microscope and a transmission electron microscope. The fracture surfaces of the breached section showed brittle intergranular fracture characteristics for both the axial and circumferential cracks. Formation of γ' in the matrix near the breach confirmed that the irradiation temperature at the breached area was below 500 0 C, in agreement with other estimates of the temperature for the area, 447 to 526 0 C. A hexagonal eta-phase, Ni 3 (Ti,Nb), precipitated at boundaries near the breach. A more extensive eta-phase coating at grain boundaries was found in a section irradiated at 650 0 C. The eta-phase plates at grain boundaries are expected to have a detrimental effect on alloy ductility. A plane of weakness in this region along the (111) slip planes will develop in Inconel 706 because the eta-plates have a (111) habit relationship with the matrix

A comparative CFD investigation of helical wire-wrapped 7, 19 and 37 fuel pin bundles and its extendibility to 217 pin bundle

International Nuclear Information System (INIS)

Gajapathy, R.; Velusamy, K.; Selvaraj, P.; Chellapandi, P.; Chetal, S.C.

2009-01-01

Preliminary investigations of sodium flow and temperature distributions in heat generating fuel pin bundles with helical spacer wires have been carried out. Towards this, the 3D conservation equations of mass, momentum and energy have been solved using a commercial computational fluid dynamics (CFD) code. Turbulence has been accounted through the use of high Reynolds number version of standard k-ε model, with uniform mesh density respecting wall function requirements. The geometric details of the bundle and the heat flux in are similar to that of the Indian Prototype Fast Breeder Reactor (PFBR) that is currently under construction. The mixing characteristics of the flow among the peripheral and central zones are compared for 7, 19 and 37 fuel pin bundles and the characteristics are extended to a 217 pin bundle. The friction factors of the pin bundles obtained from the present study is seen to agree well with the values derived from experimental correlations. It is found that the normalized outlet velocities in the peripheral and central zones are nearly equal to 1.1-0.9, respectively which is in good agreement with the published hydraulic experimental measurements of 1.1-0.85 for a 91 pin bundle. The axial velocity is the maximum in the peripheral zone where spacer wires are located and minimum in the zones which are diametrically opposite to the respective zone of maximum velocity. The sodium temperature is higher in the zones where the flow area and mass flow rates are less due to the presence of the spacer wires though the axial velocity is higher there. It is the minimum in the peripheral zones where the circumferential flow is larger. Based on the flow and temperature distributions obtained for 19 and 37 pin bundles, a preliminary extrapolation procedure has been established for estimating the temperatures of peripheral and central zones of 217 pin bundle.

Recoil release of fission products from nuclear fuel

International Nuclear Information System (INIS)

Wise, C.

1985-01-01

An analytical approximation is developed for calculating recoil release from nuclear fuel into gas filled interspaces. This expression is evaluated for a number of interspace geometries and shown to be generally accurate to within about 10% by comparison with numerical calculations. The results are applied to situations of physical interest and it is demonstrated that recoil can be important when modelling fission product release from low temperature CAGR pin failures. Furthermore, recoil can contribute significantly in experiments on low temperature fission product release, particularly where oxidation enhancement of this release is measured by exposing the fuel to CO 2 . The calculations presented here are one way of allowing for this, other methods are suggested. (orig.)

Performance of electrical contact pins near a nuclear explosion

International Nuclear Information System (INIS)

Ragan, C.E.; Silbert, M.G.; Ellis, A.N.; Robinson, E.E.; Daddario, M.J.

1977-09-01

The pressures attainable in equation-of-state studies using nuclear-explosion-driven shock waves greatly exceed those that can be reached in normal laboratory conditions. However, the diagnostic instrumentation must survive in the high-radiation environment present near such an explosion. Therefore, a set of experiments were fielded on the Redmud event to test the feasibility of using electrical contact pins in this environment. In these experiments a 60-cm-high shield of boron-lead was placed on the rack lid approximately 1 m from the device. A sample consisting of slabs of molybdenum and 238 U was placed on top of the shield, and twelve electrical contact pins were embedded to five different depths in the materials. Five different multiplexing-charging circuits were used for the pins, and a piezoelectric quartz gauge was placed on top of the uranium to obtain an estimate of the fission-energy deposition. All of the charged pins survived the radiation and produced signals indicating shock arrival. The uncertainty in determining the pin-closure time was approximately 3 ns. The signal from the quartz gauge corresponded to a pressure that was consistent with the calculated neutron fluence

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.)

User's guide to EPIC, a computer program to calculate the motion of fuel and coolant subsequent to pin failure in an LMFBR

Energy Technology Data Exchange (ETDEWEB)

Pizzica, P.A.; Garner, P.L.; Abramson, P.B.

1979-10-01

The computer code EPIC models fuel and coolant motion which results from internal fuel pin pressure (from fission gas or fuel vapor) and possibly from the generation of sodium vapor pressure in the coolant channel subsequent to pin failure in a liquid-metal fast breeder reactor. The EPIC model is restricted to conditions where fuel pin geometry is generally preserved and is not intended to treat the total disruption of the pin structure. The modeling includes the ejection of molten fuel from the pin into a coolant channel with any amount of voiding through a clad breach which may be of any length or which may extend with time. One-dimensional Eulerian hydrodynamics is used to treat the motion of fuel and fission gas inside a molten fuel cavity in the fuel pin as well as the mixture of two-phase sodium and fission gas in the coolant channel. Motion of fuel in the coolant channel is tracked with a type of particle-in-cell technique. EPIC is a Fortran-IV program requiring 400K bytes of storage on the IBM 370/195 computer. 21 refs., 2 figs.

Spent nuclear fuel structural response when subject to an end impact accident

Energy Technology Data Exchange (ETDEWEB)

Tang, D.T.; Guttmann, J. [United States Nuclear Regulatory Commission, Rockville, MD (United States)]|[United States Nuclear Regulatory Commission, Washington, DC (United States); Koeppel, B.J.; Adkins, H.E.

2004-07-01

The US Nuclear Regulatory Commission (USNRC) is responsible for licensing spent fuel storage and transportation systems. A subset of this responsibility is to investigate and understand the structural performance of these systems. Studies have shown that the fuel rods of intact spent fuel assemblies with burn-ups up to 45 gigawatt days per metric ton of uranium (Gwd/MTU) are capable of resisting the normally expected impact loads subjected during drop accident conditions. However, effective cladding thickness for intact spent fuel assemblies with burn ups greater than 45 Gwd/MTU can be reduced due to corrosion. The capability of the fuel rod to withstand the expected loads encountered under normal and accident conditions may also be reduced, given degradation of the material properties under extended use, such as decrease in ductility. The USNRC and Pacific Northwest Laboratory (PNNL) performed computational studies to predict the structural response of spent nuclear fuel in a transport system that is subjected to a hypothetical regulatory impact accident, as defined in 10 CFR71.73. This study performs a structural analysis of a typical high burn up Pressurized Water Reactor (PWR) fuel assembly using the ANSYS {sup registered} ANSYS {sup registered} /LS- DYNA {sup registered} finite element analysis (FEA) code. The material properties used in the analyses were based on expert judgment and included uncertainties. Ongoing experimental programs will reduce the uncertainties. The current evaluations include the pins, spacer grids, and tie plates to assess possible cladding failure/rupture under hypothetical impact accident loading. This paper describes the USNRC and PNNL staff's analytical approach, provides details on the single pin model developed for this assessment, and presents the results.

Test Specifications and the Design of the Wire Wrapped 37-Pin Fuel Assembly for Hydrodynamic Experiments

International Nuclear Information System (INIS)

Chang, S. K.; Euh, D. J.; Bae, H.; Lee, H. Y.; Choi, S. R.

2013-01-01

Most influencing parameters on uncertainties and sensitivities of the CFD analyses are the friction coefficient and the mixing coefficient. The friction coefficient is related to the flow distribution in reactor sub-channels. The mixing coefficient is defined with the cross flow between neighboring sub-channels. The eventual purpose of the thermal hydraulic design considering these parameters is to guarantee the fuel cladding integrity as the design limit parameter. At the moment, the experimental program is being undertaken to quantify these friction and mixing parameters which characterize the flow distribution in sub-channels, and the wire wrapped 37-pin rod assembly and its hexagonal test rig have been designed and fabricated. The quantified thermal hydraulic experimental data from this program are utilized primarily to estimate the accuracy of the safety analysis codes and their thermal hydraulic model. A wire wrapped 37 pin fuel assembly has been designed for the measurements of the flow distribution, where the measurements are utilized to quantify the friction coefficient and the mixing coefficient. The test rig of the wire wrapped 37 pin fuel assembly has been fabricated considering the geometric and flow dynamic similarities. It comprises four components i. e., the upper plenum, the fuel housing, the lower plenum, and the wire wrapped 37 pin fuel assembly. At further works, the quantified friction and mixing coefficients through the experiments are going to be utilized for insuring the reliability of the CFD analysis results

SIEX3: A correlated computer code for prediction of fast reactor mixed oxide fuel and blanket pin performance

International Nuclear Information System (INIS)

Baker, R.B.; Wilson, D.R.

1986-04-01

The SIEX3 computer program was developed to calculate the fuel and cladding performance of oxide fuel and oxide blanket pins irradiated in the fast neutron environment of a liquid metal cooled reactor. The code is uniquely designed to be accurate yet quick running and use a minimum of computer core storage. This was accomplished through the correlation of physically based models to very large data bases of irradiation test results. Data from over 200 fuel pins and over 800 transverse fuel microscopy samples were used in the calibrations

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Fabrication and post-irradiation examination of a zircaloy-2 clad UO2-1.5 wt% PuO2 fuel pin irradiated in PWL, CIRUS

International Nuclear Information System (INIS)

Sah, D.N.; Sahoo, K.C.; Chatterjee, S.; Majumdar, S.; Kamath, H.S.; Ramachandran, R.; Bahl, J.K.; Purushottam, D.S.C.; Ramakumar, M.S.; Sivaramakrishnan, K.S.; Roy, P.R.

1977-01-01

A zircaloy-2 clad UO 2 -1.5 wt% PuO 2 fuel pin was fabricated at the Radiometallurgy Section of the Bhabha Atomic Research Centre, Bombay, for irradiation in the pressurised water loop in CIRUS. Requisite development work related to powder conditioning, blending, pressing and sintering parameters was carried out to meet the exacting fuel pellet specifications of CANDU fuel. The fuel pin ruptured while being irradiated in the pressurised water loop in CIRUS, after experiencing a low burn-up of 507 MWD/MTM and was subsequently examined at the Radiometallurgy Hot Cells Facility. The results showed that internal clad hydriding led to primary failure of the fuel pin. Subsequent ingress of the coolant water caused excessive swelling of the thermal insulating magnesia pellets located at the ends of the fuel column. The swelling of magnesia pellets caused severe rupturing of the fuel pin at the two ends. The delayed rupturing of the fuel pin at the upper end, caused the fuel column to be displaced downwards by 5.85mm. (author)

Immersed multiple device for the control of the irradiated PWR fuel pins in the reloadable loop in the OSIRIS pond

International Nuclear Information System (INIS)

Farny, G.

1983-01-01

With respect to the dynamics of the degradation of the PWR fuel in transient, normal and abnormal regions, a new multi-device immersed in the cooling pond of the OSIRIS reactor, is studied. The multiple device is subjected to three examinations: (1) visual studying and video-recording of the appearance of the fuel pins, (2) metrology of the pins, (3) investigation of the induced Foucault currents in the fuel cans. Attention is chiefly paid to the last point; the other ones - being closely related - are only touched on whenever needed. It is concluded that quality control of the fuel pins is possible by means of Foucault currents without applying mechanical constraints and without interfering with the cooling rate. (Auth.)

Grid spacers for use in a nuclear fuel assembly

International Nuclear Information System (INIS)

Kuwako, Akira.

1987-01-01

Purpose: To obtain spacers capable of reducing the pressure loss by enlarging coolant flow channels when the fuel temperature is high, while capable of reliably maintaining the fuel pins with no vibrations when the fuel temperature is low. Constitution: This invention concerns grid spacers for constituting fuel assemblies for use in water cooled reactors. Memory shape alloys are disposed at least a portion of a spacer element that takes such a shape as urging the pin when the fuel temperature is low, while enlarging the coolant flow channel to reduce the pressure loss when the fuel temperature is high. (Ikeda, J.)

Fuel or irradiation subassembly

International Nuclear Information System (INIS)

Seim, O.S.; Hutter, E.

1975-01-01

A subassembly for use in a nuclear reactor is described which incorporates a loose bundle of fuel or irradiation pins enclosed within an inner tube which in turn is enclosed within an outer coolant tube and includes a locking comb consisting of a head extending through one side of the inner sleeve and a plurality of teeth which extend through the other side of the inner sleeve while engaging annular undercut portions in the bottom portion of the fuel or irradiation pins to prevent movement of the pins

Application of vacuum technology during nuclear fuel fabrication, inspection and characterization

International Nuclear Information System (INIS)

Majumdar, S.

2003-01-01

Full text: Vacuum technology plays very important role during various stages of fabrication, inspection and characterization of U, Pu based nuclear fuels. Controlled vacuum is needed for melting and casting of U, Pu based alloys, picture framing of the fuel meat for plate type fuel fabrication, carbothermic reduction for synthesis of (U-Pu) mixed carbide powder, dewaxing of green ceramic fuel pellets, degassing of sintered pellets and encapsulation of fuel pellets inside clad tube. Application of vacuum technology is also important during inspection and characterization of fuel materials and fuel pins by way of XRF and XRD analysis, Mass spectrometer Helium leak detection etc. A novel method of low temperature sintering of UO 2 developed at BARC using controlled vacuum as sintering atmosphere has undergone successful irradiation testing in Cirus. The paper will describe various fuel fabrication flow sheets highlighting the stages where vacuum applications are needed

A thermodynamic model for the attack behaviour in stainless steel clad oxide fuel pins

International Nuclear Information System (INIS)

Goetzmann, O.

1979-01-01

So far, post irradiation examination of burnt fuel pins has not revealed a clear cut picture of the cladding attack situation. For seemingly same conditions sometimes attack occurs, sometimes not. This model tries to depict the reaction possibilities along the inner cladding wall on the basis of thermodynamic facts in the fuel pin. It shows how the thermodynamic driving force for attack changes along the fuel column, and with different initial and operational conditions. Two criteria for attack are postulated: attack as a result of the direct reaction of reactive elements with cladding components; and attack as a result of the action of a special agent (CsOH). In defining a reaction potenial the oxygen potential, the temperature conditions (cladding temperature and fuel surface temperature), and the fission products are involved. For the determination of the oxygen potential at the cladding, three models for the redistribution of oxygen across the fuel/clad gap are offered. The effect of various parameters, like rod power, gap conductance, oxygen potential, inner wall temperature, on the thermodynamic potential for attack is analysed. (Auth.)

Nuclear fuel

International Nuclear Information System (INIS)

D Hondt, P.

1998-01-01

The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

Development of Melting Crucible Materials of Metallic Fuel Slug for SFR

International Nuclear Information System (INIS)

Kim, K. H.; Lee, C. T.; Oh, S. J.; Kim, S. K.; Lee, C. B.; Ko, Y. M.; Woo, W. M.

2010-01-01

The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr-Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized

The thermal-mechanical behavior of fuel pins during power's maneuvering regime at stationary core loading on 2nd unit of KHNPP

International Nuclear Information System (INIS)

Ieremenko, M.; Ovdiyenko, Y.; Khalimonchuk, V.

2007-01-01

Results of thermal-mechanical behaviour of fuel pins during daily power's maneuvering regime that were proposed for second unit of Khmelnitsky NPP are presented. Calculations were performed for campaign's moments 100 and 160 fpd and for different type of regulation. Additionally calculations were performed for campaign 7. It is the design variant of the campaign and reactor core contains the high burnt fuel. Calculations of macro-core parameters (Kq, Kv) was performed by spatial computer code DYN3D. Calculations of micro-core parameters (fuel pin power) was performed by computer code DERAB. Calculations of thermal-mechanical behaviour of fuel pins was performed by computer code TRANSURANUS (Authors)

Cesium relocation in mixed-oxide fuel pins resulting from increased temperature reirradiation

International Nuclear Information System (INIS)

Lawrence, L.A.; Woodley, R.E.; Weber, E.T.

1976-06-01

Mixed-oxide fuel pins from EBR-II test subassemblies PNL-3 and PNL-4 were reirradiated in the GETR to study effects of increased fuel and cladding temperatures on chemical and thermomechanical behavior. Radial and axial distributions of cesium were obtained using postirradiation nondestructive precision gamma-scanning techniques. Data presented relate to the dependence of cesium distribution and transport processes on temperature gradients which were altered after substantial steady-state operation

DIMCO. A new system for mechanical and bidimensional of nuclear fuel pins

International Nuclear Information System (INIS)

Moreno, A.

1977-01-01

The system developed in JEN, for: the mechanical analysis uni and bidimensional, of nuclear fuels is presented. The mathematical and numerical foundations used, are here described. And so the models developed for effects such as swelling, cracking, clad growth etc. Numerical results for several cases are presented. a) Numerical test in one and two dimensions. b) Applicability range, c) Interaction effects. d) Influence of the power history. (Author) 17 refs

Development of end plug welding method in the fabrication of FBR fuel pins

International Nuclear Information System (INIS)

Ohtani, Seiji; Sawayama, Takeo; Tateishi, Yoshinori

1977-01-01

As a part of the development of the automatic and remote controlled fabrication of FBR fuel pins, welding of fuel pin end plugs has been examined. Cladding tubes and end plugs used for this experiment are made of SUS 316, and they are the components of fuel pins for the prototype fast breeder reactor (Monju) or the second core of Joyo (Joyo MK-II). The welding tests of cladding tubes and four kinds of end plugs were carried out by means of two techniques; tungsten inert gas welding and laser welding. It can be said that no considerable difference was observed in weld penetration, occurrence rate of weld defects and breaking strength between the tight fit and the loose fit plugs. The face-to-face fit welding requires the least welding heat input, but involves much difficulty in the control of weld penetration and bead zone diameter. The good concentrative property and high energy density of laser beam make the face of weld hollow due to the vaporization of weld metal. However, this problem can be easily solved by changing the shape of end plugs. Good results in the other characteristics of the weld also were obtained by this laser welding. Further experiment is needed in connection with the compatibility of weld metal with sodium and neutron irradiation before final judgement is made on the laser welding technique. (Nakai, Y.)

Fuel pin behavior of a pressurizer water reactor with load following

International Nuclear Information System (INIS)

Perrotta, J.A.

1980-10-01

The performance of a PWR fuel pin was evaluated, during power cycles that occur in normal operations, excluding accident cases. A code to perform the mechanical analysis of the cladding was developed using the Finite Element Method to take into account local effects of pellet-cladding interaction (PCI). (E.G.) [pt

SIEX design predictions for the PNC fuel pins in the HEDL P-E01 power-to-melt test

International Nuclear Information System (INIS)

1979-01-01

During the design phase of the HEDL P-E01 power-to-melt test, a series of design predictions were generated for the three PNC pins using the SIEX fuel pin modeling code. This document tabulates a series of selected PNC pin design predictions as requested by M. Shinohara during his visit to HEDL

Tests of the SNR fuel pin behaviour in case of operational transients in the HFR Petten

International Nuclear Information System (INIS)

Plitz, H.

1989-05-01

The loadings on fast reactor fuel pins under operational transients (power and temperature increases in the design area) have been studied in the High-Flux-Reactor HFR in Petten with sodium cooled irradiation capsules. The results of the first campaign of transient experiments are described in the report. No cladding defects have been observed, and the fuel pins of the Mark-I and Mark-II type resisted to linear power levels of more than 800 W/cm, thus demonstrating the required design margins. The plans for further experiments are outlined

PIN99W, Modelling of VVER and PWR Fuel Rod Thermomechanical Behaviour

International Nuclear Information System (INIS)

Valach, M.; Strizhov, P.; Svoboda, R.

2000-01-01

1 - Description of program or function: The Code is developed to describe fuel rod thermomechanical behaviour in operational conditions. The main goal of this code is to calculate fuel temperature, gap conductivity, fission gas release and inner gas pressure. 2 - Methods: - fuel rod temperature response is solved by using one-dimensional finite element method combined with weighted residuals method; - the code involves models describing physical phenomena typical for the fuel irradiated in Light Water Power Reactors (densification, restructuring, fission gas release, swelling and relocation) ; - this code is updated and improves PIN-micro code. 3 - Restrictions on the complexity of the problem: - simplified mechanistic solution; - only steady-state solution; - no cladding failure criterion; - no model for axial fuel-cladding interaction

Investigation of velocity distribution in an inner subchannel of wire wrapped fuel pin bundle of sodium-cooled fast reactor

International Nuclear Information System (INIS)

Nishimura, Masahiro; Kamide, Hideki; Ohshima, Hiroyuki; Kobayashi, Jun; Sato, Hiroyuki

2011-01-01

A sodium cooled fast reactor is designed to attain a high burn-up of core fuel in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the detail of flow velocity distribution in a wire wrapped pin bundle. In this study, water experiments were carried out to investigate the detailed velocity distribution in a subchannel of nominal pin geometry as the first step. These basic data are not only useful for understanding of pin bundle thermal hydraulics but also a code validation. A wire-wrapped 3-pin bundle water model was applied to investigate the detailed velocity distribution in the subchannel which is surrounded by 3 pins with wrapping wire. The test section consists of an irregular hexagonal acrylic duct tube and three pins made of fluorinated resin pins which has nearly the same refractive index with that of water and a high light transmission rate. This enables to visualize the central subchannel through the pins. The velocity distribution in the central subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through a side wall of the duct tube. Typical flow velocity conditions in the pin bundle were 0.36m/s (Re=2,700) and 1.6m/s (Re=13,500). Influence of the wrapping wire on the velocity distributions in vertical and horizontal directions was confirmed. A clockwise swirl flow around the wire was found in subchannel. Significant differences were not recognized between the two cases of Re=2,700 and 13,500 concerning flow patterns. (author)

Nuclear fuel string assembly

International Nuclear Information System (INIS)

Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

1976-01-01

A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

Annual report of the working group 'fuel pin and fuel element mechanics' of the Institut fuer Reaktortechnik (IRT) of the Technische Hochschule Darmstadt for the Fast Breeder Project

International Nuclear Information System (INIS)

Fabian, H.; Humbach, W.; Lassmann, K.; Mueller, J.J.; Preusser, T.; Schmelz, K.

1978-09-01

This report comprises six single lectures given at an information meeting organized by the Institut fuer Reaktortechnik der Technischen Hochschule Darmstadt (IRT) in Darmstadt on April 24, 1978. The lectures are an account of work performed at IRT on the mechanics of fuel pins and fuel elements and supported by the Fast Breeder Project (PSB) of KfK. These activities can be broken down into studies of the integral fuel pin (URANUS computer code) and into multidimensional studies of the fuel pin using the finite-element method (FINEL and ZIDRIG computer codes). Moreover, a report is presented of the status of the test facility for simulation of out-of-pile cladding tube loads and of the IRT project on the simulation and analysis of radiation damage. (orig./GL) [de

Study of fuel bundle geometry on inter subchannel flow in a 19 pin wire wrapped bundle

International Nuclear Information System (INIS)

Naveen Raj, M.; Velusamy, D.K.

2015-01-01

In typical sodium cooled fast reactor (SFR) fuel pin bundle, gap between the pins is maintained by helically wound wire wrap around each pin. The presence of wire induces large inter-subchannel transverse flow, eventually promoting mixing and heat transfer. The magnitude of the transverse flow is highly dependent on the various pin-bundle dimensions. Appropriate modeling of these transverse flows in subchannel codes is necessary to predict realistic temperature distribution in pin bundle. Hence, detailed parametric study of transverse flow on pin-bundle geometric parameters has been conducted. The parameters taken for the present study are pin diameter, wire diameter, helical wire pitch and edge gap. Towards this 3-D computational fluid dynamic analysis on a structured mesh of 19 pin bundle is carried out using k-epsilon turbulence model. Periodic oscillations along the primacy flow direction were found in subchannel transverse flow and peripheral pin clad temperatures with periodicity over one pitch length. Based on parametric studies, correlations for transverse flow in central subchannels are proposed. (author)

Improvements in nuclear fuel assembly cages

Energy Technology Data Exchange (ETDEWEB)

Eaton, C.W.; Seeley, T.A.; Ince, G.; Speakman, W.T.

1986-03-12

The fuel pin/guide tube supporting grids of an assembly cage for a multi pin fuel element or a reflector element for a stringer are mounted in the moderator sleeve by way of mounting assemblies engaged in grooves machined into the interior surface of the sleeve, each mounting assembly including a split ring which is assembled into its groove by being radially contracted, pushed along the sleeve into registry with the groove and allowed to radially expand. The split ring may carry burnable neutron absorber. The region of the sleeve between two adjacent grids may be of smaller internal diameter than the remainder of the sleeve.

Development and testing of high-performance fuel pin simulators for boiling experiments in liquid metal flow

International Nuclear Information System (INIS)

Casal, V.

1976-01-01

There are unknown phenomena, about local and integral boiling events in the core of sodium cooled fast breeder reactors. Therefore at GfK depend out-of-pile boiling experiments have been performed using electrically heated dummies of fuel element bundles. The success of these tests and the amount of information derived from them depend exclusively on the successful simulation of the fuel pins by electrically heated rods as regards the essential physical properties. The report deals with the development and testing of heater rods for sodium boiling experiments in bundles including up to 91 heated pins

MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR

Directory of Open Access Journals (Sweden)

Suwardi Suwardi

2015-03-01

Full Text Available MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR*. The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared at the CNFT (Center for Nuclear Fuel Technology then a ramp test will be performed. The present work is part of designing first irradiation experiments in the PRTF (Power Ramp Test Facility of RSG-GAS 30 MW reactor. The thermal mechanic of the pin during irradiation has simulated. The geometry variation of pellet and cladding is modeled by taking into account different phenomena such as thermal expansion, densification, swelling by fission product, thermal creep and radiation growth. The cladding variation is modeled by thermal expansion, thermal and irradiation creeps. The material properties are modeled by MATPRO and standard numerical parameter of TRANSURANUS code. Results of irradiation simulation with 9 kW/m LHR indicates that pellet-clad contacts onset from 0.090 mm initial gaps after 806 d, when pellet radius expansion attain 0.015 mm while inner cladding creep-down 0.075 mm. A newer computation data show that the maximum measured LHR of n-UO2 pin in the PRTF 12.4 kW/m. The next simulation will be done with a higher LHR, up to ~ 25 kW/m. MODEL SIMULASI VARIASI GEOMETRI DAN STRESS-STRAIN DARI PROTOTIP BAHAN BAKAR PIN BATAN SELAMA UJI IRADIASI DI REAKTOR RSG-GAS. Pusat Teknologi Bahan Bakar Nuklir (PTBBN telah menyiapkan tangkai (pin bahan bakar pendek perdana yang berisi pelet UO2 alam dalam kelongsong paduan zircaloy untuk dilakukan uji iradiasi daya naik. Penelitian ini merupakan bagian dari perancangan percobaan iradiasi pertama di PRTF (Power Ramp Test Fasility yang terpasang di reaktor serbaguna RSG-GAS berdaya 30 MW. Telah dilakukan pemodelan dan simulasi kinerja termal mekanikal pin selama iradiasi. Variasi geometri pelet dan kelongsong selama pengujian dimodelkan dengan memperhatikan fenomena ekspansi termal

Nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Nakano, H [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

1976-10-01

It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts.

The Nuclear Fuel Cycle

International Nuclear Information System (INIS)

2011-08-01

This brochure describes the nuclear fuel cycle, which is an industrial process involving various activities to produce electricity from uranium in nuclear power reactors. The cycle starts with the mining of uranium and ends with the disposal of nuclear waste. The raw material for today's nuclear fuel is uranium. It must be processed through a series of steps to produce an efficient fuel for generating electricity. Used fuel also needs to be taken care of for reuse and disposal. The nuclear fuel cycle includes the 'front end', i.e. preparation of the fuel, the 'service period' in which fuel is used during reactor operation to generate electricity, and the 'back end', i.e. the safe management of spent nuclear fuel including reprocessing and reuse and disposal. If spent fuel is not reprocessed, the fuel cycle is referred to as an 'open' or 'once-through' fuel cycle; if spent fuel is reprocessed, and partly reused, it is referred to as a 'closed' nuclear fuel cycle.

Post-Irradiation Examination of Fuel Pin R54-F20A, Irradiated in a NaK Environment. RCN Report

International Nuclear Information System (INIS)

Kwast, H.

1972-12-01

Fuel pin R54-F20A has been irradiated in a NaK-environment. Temperature measurements in the NaK were carried out at average linear fission powers of 552 and 825 W/cm respectively. A maximum average canning temperature of 920°C was reached. The fuel pin was irradiated for about 50 minutes at the maximum irradiation conditions, while the total irradiation time was two hours. The irradiation had to be broken off before the end condition was reached because of malfunctioning of the fuelfailure detection system. No power peaking did occur at the upper and lower interfaces between the 50%-enriched UO 2 - and the natural UO 2 + 8 w/o UB 4 pellet. About 35% of the fuel has molten, but the fuel pin did not fail. The irradiation has been carried out in the Poolside Facility (PSF) of the High Flux Reactor (HFR) at Petten. (author)

The KNK II/1 fuel assembly NY-205: Compilation of the irradiation history and the fuel and fuel pin fabrication data of the INTERATOM data bank system BESEX

International Nuclear Information System (INIS)

Patzer, G.; Geier, F.

1988-01-01

The fuel assembly NY-205 has been irradiated during the first and the second core of KNK II with a total residence time of 832 equivalent full-power days. A maximum burnup of 175.000 MWd/tHM or 18.6 % was reached with a maximum steel damage of 66 dpa-NRT. For the cladding the materials 1.4970 and 1.4981 have been used in different metallurgical conditions, and for the Uranium/Plutonium mixed- oxide fuel the most important variants of the major fabrication parameters had been realized. The assembly will be brought to the Hot Cells of the KfK Karlsruhe for post-irradiation examination in February 1988, so that the knowledge of the fabrication data is of interest for the selection of fuel pins and for the evaluation of the examination results. Therefore this report compiles the fuel and fuel pin fabrication data from the INTERATOM data bank system BESEX and additionally, an overview of the irradiation history of the assembly is given [de

Electric arc apparatus for severing split-pin assemblies of guide tubes of nuclear reactors

International Nuclear Information System (INIS)

Burns, D.C.; Kauric, C.E.; Persang, J.C.

1987-01-01

This patent describes an apparatus for use in the replacement of an old split-pin assembly of a guide tube of a nuclear reactor by a new split-pin assembly, the old split-pin assembly including an old split pin and an old nut securing the old split pin to the guide tube, the old split-pin assembly and the guide tube being radioactive. The apparatus includes a metal disintegration machining tool, the tool having an electrode, means for mounting the tool submerged in a pool of water in engagement with the guide tube and with the old split-pin assembly secured to the guide tube, the tool being so mounted with the electrode in position to coact electrically with the last-named old split-pin assembly but not with the guide tube, and means, connected to the tool, for firing a disintegrating arc between the electrode and the assembly to disintegrate the assembly into readily removable fragments

Materials properties utilization in a cumulative mechanical damage function for LMFBR fuel pin failure analysis

International Nuclear Information System (INIS)

Jacobs, D.C.

1977-01-01

An overview is presented of one of the fuel-pin analysis techniques used in the CRBRP program, the cumulative mechanical damage function. This technique, as applied to LMFBR's, was developed along with the majority of models used to describe the mechanical properties and environmental behavior of the cladding (i.e., 20 percent cold-worked, 316 stainless steel). As it relates to fuel-pin analyses the Cumulative Mechanical Damage Function (CDF) continually monitors cladding integrity through steady state and transient operation; it is a time dependent function of temperature and stress which reflects the effects of both the prior mechanical history and the variations in mechanical properties caused by exposure to the reactor environment

Dynamic behaviour of FBR fuel pin bundles

International Nuclear Information System (INIS)

Martin, P.H.; Van Dorsselaere, J.P.; Ravenet, A.

1990-01-01

A programme of shock tests on a fast neutron reactor subassembly model (SPX1 geometry) including a complete bundle of fuel pins (dummy elements) is being carried out in the BELIER test facility at Cadarache. The purpose of these tests is: to determine the distribution of dynamic forces applied to the fuel rod clads under the impact conditions encountered in a reactor during a earthquake; to reduce as much as possible the conservatism of the methods presently used for the calculation of those forces. The test programme, now being completed, consists of the following steps: impacts on the mock-up in air with an non-compact bundle (situation of the subassembly at beginning of life (BOL) with clearances within the bundle); impacts under the same conditions but with fluid (water) in the subassembly; impacts on the mock-up in air and with a compacted bundle (simulating the conditions of an end-of-life (EOL) bundle with no clearance within the bundle). The accelerations studied in these tests cover the range encountered in design calculations for the subassembly frequencies in beam mode. (author)

Nuclear fuels

International Nuclear Information System (INIS)

2008-01-01

The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

Fuel and nuclear fuel cycle

International Nuclear Information System (INIS)

Prunier, C.

1998-01-01

The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

A device for supporting a pin bundle in a nuclear reactor assembly casing

International Nuclear Information System (INIS)

Marmonier, Pierre; Mesnage, Bernard; Teulon, Jean; Vayra, Jean; Venobre, Henri.

1974-01-01

Description is given of a device for supporting a pin-bundle in a nuclear reactor assembly casing. That device comprises a member coaxially mounted at the bottom of the vertically mounted casing, adapted to support a plurality of parallel rails along whose edges slide grooves made in the pin-plugs. It is characterized in that said supporting member is provided with a lateral groove open toward its periphery, cooperating with clamping-lugs that form extensions of the rail-sides and comprise an inwardly directed portion adapted to be engaged in the groove. This can be applied to fast neutron nuclear reactors [fr

Power release estimation inside of fuel pins neighbouring fuel pin with gadolinium in a WWER-1000 type core

International Nuclear Information System (INIS)

Mikus, J.

2006-01-01

The purpose of this work consists in investigation of the gadolinium fuel pin (fps) influence on space power distribution, especially from viewpoint of the values and gradient occurrence inside of neighbouring FPs that could result in static loads with some consequences, e.g., FP bowing. Since detailed power distributions cannot be obtained in the NPPs, needed information is provided by means of experiments on research reactors. As for the power release measurement inside of FPs, some special (e.g. track) detectors placed between fuel pellets are usually used. Since such works are relatively complicated and time consuming, an evaluation method based on mathematical modelling and numerical approximation was proposed by means of that, and using measured (integral) power release in selected FPs, relevant information about power release inside of needed (investigated) FP, can be obtained. For this purpose, an experiment on light water, zero-power research reactor LR-0 was realized in a WWER-1000 type core with 7 fuel assemblies at zero boron concentration and containing gadolinium FPs. Application of the above evaluation method is demonstrated on investigated FP neighbouring a FP with gadolinium by means of the 1) Azimuthal power distribution inside of investigated FP on their fuel pellet surface in horizontal plane and 2) Gradient of the power distribution inside of investigated FP in two opposite positions on pellets surface that are situated to- and outwards a FP with gadolinium. Similar information can be relevant from the viewpoint of the FP failures occurrence investigation (Authors)

On the behaviour of dissolved fission gases prior to transient testing of fuel pins

International Nuclear Information System (INIS)

Wood, M.H.; Matthews, J.R.

1978-10-01

The TREAT and CABRI series of reactor safety experiments on irradiated fuel require the transfer of fuel pins from the reactor in which the fuel has achieved some burn-up to the test facility. Subsequently, the fuel is restored to power in the test facility for some time before transient heating is initiated. Such pre-test manoeuvres, where the fuel is subjected to changes in the fission rate and temperature, may have important consequences for the fission gas behaviour during the transient experiment. The results of rate theory calculations are used to assess these effects. (author)

Romanian nuclear fuel program

International Nuclear Information System (INIS)

Budan, O.

1999-01-01

The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

Implementation, verification, and validation of the FPIN2 metal fuel pin mechanics model in the SASSYS/SAS4A LMR transient analysis codes

International Nuclear Information System (INIS)

Sofu, T.; Kramer, J.M.

1994-01-01

The metal fuel version of the FPIN2 code which provides a validated pin mechanics model is coupled with SASSYS/SAS4A Version 3.0 for single pin calculations. In this implementation, SASSY/SAS4A provides pin temperatures, and FPIN2 performs analysis of pin deformation and predicts the time and location of cladding failure. FPIN2 results are also used for the estimates of axial expansion of fuel and associated reactivity effects. The revalidation of the integrated SAS-FPIN2 code system is performed using TREAT tests

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

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

International Nuclear Information System (INIS)

Blanchard, A.; Campion, P.

1980-01-01

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

Application of core structural design guidelines in conceptual fuel pin design

International Nuclear Information System (INIS)

Patel, M.R.; Stephen, J.D.

1979-01-01

The paper describes an application of the Draft RDT Standards F9-7, -8, and -9 to conceptual design of Fast Breeder Reactor (FBR) fuel pins. The Standards are being developed to provide guidelines for structural analysis and design of the FBR core components which have limited ductility at high fluences and are not addressed by the prevalent codes. The development is guided by a national working group sponsored by the Division of Reactor Researcch and Technology of the Department of Energy. The development program summarized in the paper includes establishment of design margins consistent with the test data and component performance requirements, and application of the design rules in various design activities. The application program insures that the quantities required for proper application of the design rules are available from the analysis methods and test data, and that the use of the same design rules in different analysis tools used at different stages of a component design producees consistent results. This is illustrated in the paper by application of the design rules in the analysis methods developed for conceptual and more detailed designs of an FBR fuel pin

Conversion from film to image plates for transfer method neutron radiography of nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Craft, Aaron E.; Papaioannou, Glen C.; Chichester, David L.; Williams, Walter J.

2017-02-01

This paper summarizes efforts to characterize and qualify a computed radiography (CR) system for neutron radiography of irradiated nuclear fuel at Idaho National Laboratory (INL). INL has multiple programs that are actively developing, testing, and evaluating new nuclear fuels. Irradiated fuel experiments are subjected to a number of sequential post-irradiation examination techniques that provide insight into the overall behavior and performance of the fuel. One of the first and most important of these exams is neutron radiography, which provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Results from neutron radiography are often the driver for subsequent examinations of the PIE program. Features of interest that can be evaluated using neutron radiography include irradiation-induced swelling, isotopic and fuel-fragment redistribution, plate deformations, and fuel fracturing. The NRAD currently uses the foil-film transfer technique with film for imaging fuel. INL is pursuing multiple efforts to advance its neutron imaging capabilities for evaluating irradiated fuel and other applications, including conversion from film to CR image plates. Neutron CR is the current state-of-the-art for neutron imaging of highly-radioactive objects. Initial neutron radiographs of various types of nuclear fuel indicate that radiographs can be obtained of comparable image quality currently obtained using film. This paper provides neutron radiographs of representative irradiated fuel pins along with neutron radiographs of standards that informed the qualification of the neutron CR system for routine use. Additionally, this paper includes evaluations of some of the CR scanner parameters and their effects on image quality.

Nuclear-fuel-cycle education: Module 1. Nuclear fuel cycle overview

International Nuclear Information System (INIS)

Eckhoff, N.D.

1981-07-01

This educational module is an overview of the nuclear-fule-cycle. The overview covers nuclear energy resources, the present and future US nuclear industry, the industry view of nuclear power, the International Nuclear Fuel Cycle Evaluation program, the Union of Concerned Scientists view of the nuclear-fuel-cycle, an analysis of this viewpoint, resource requirements for a model light water reactor, and world nuclear power considerations

Two-dimensional steady-state thermal and hydraulic analysis code for prediction of detailed temperature fields around distorted fuel pin in LMFBR assembly: SPOTBOW

International Nuclear Information System (INIS)

Shimizu, T.

1983-01-01

SPOTBOW computer program has been developed for predicting detailed temperature and turbulent flow velocity fields around distorted fuel pins in LMFBR fuel assemblies, in which pin to pin and pin to wrapper tube contacts may occur. The present study started from the requirement of reactor core designers to evaluate local hot spot temperature due to the wire contact effect and the pin bowing effect on cladding temperature distribution. This code calculates for both unbaffled and wire-wrapped pin bundles. The Galerkin method and iterative procedure were used to solve the basic equations which govern the local heat and momentum transfer in turbulent fluid flow around the distorted pins. Comparisons have been made with cladding temperatures measured in normal and distorted pin bundle mockups to check the validity of this code. Predicted peak temperatures in the vicinity of wire contact point were somewhat higher than the measured values, and the shape of the peaks agreed well with measurement. The changes of cladding temperature due to the decrease of gap width between bowing pin and adjacent pin were predicted well

HEDL empirical correlation of fuel pin top failure thresholds, status 1976

International Nuclear Information System (INIS)

Baars, R.E.

1976-01-01

The Damage Parameter (DP) empirical correlation of fuel pin cladding failure thresholds for TOP events has been revised and recorrelated to the results of twelve TREAT tests. The revised correlation, called the Failure Potential (FP) correlation, predicts failure times for the tests in the data base with an average error of 35 ms for $3/s tests and of 150 ms for 50 cents/s tests

Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel bundle for SFR

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jae Ho; Yoo, Jin; Lee, Kwi Lim; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-08-15

Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel assembly mock-up of a Japanese loop-type sodium-cooled fast reactor, Monju, were investigated with a numerical analysis using a general-purpose commercial computational fluid dynamics code, CFX. Complicated and vortical flow phenomena in the wire-wrapped 37-pin fuel assembly were captured by a Reynolds-averaged Navier-Stokes flow simulation using a shear stress transport turbulence model. The main purpose of the current study is to understand the three-dimensional complex flow phenomena in a wire-wrapped fuel assembly to support the license issue for the core design. Computational fluid dynamics results show good agreement with friction factor correlation models. The secondary flow in the corner and edge subchannels is much stronger than that in an interior subchannel. The axial velocity averaged in the corner and edge subchannels is higher than that averaged in the interior subchannels. Three-dimensional multiscale vortex structures start to be formed by an interaction between secondary flows around each wire-wrapped pin. Behavior of the large-scale vortex structures in the corner and edge subchannels is closely related to the relative position between the hexagonal duct wall and the helically wrapped wire spacer. The small-scale vortex is axially developed in the interior subchannels. Furthermore, a driving force on each wire spacer surface is closely related to the relative position between the hexagonal duct wall and the wire spacer.

Calculation of DND-signals in case of fuel pin failures in KNK II with the computer code FICTION III

International Nuclear Information System (INIS)

Schmuck, I.

1990-11-01

In KNK II two delayed neutron detectors are installed for quick detection of fuel subassembly cladding failures. They record the release of the precursors of the emitters of delayed neutrons into the sodium. The computer code FICTION III calculates the expected delayed neutron signals for certain fuel pin failures, where the user has to set the boundary conditions interactively. In view of FICTION II the advancement of FICTION III consists of the following items: application of the data sets of 105 isotopes, distinction of thermal and fast neutron induced fission, partitioning of the sodium flow into two circuits, consideration of the specific fission rates in 10 fuel pin sections, elaboration of the user's interaction possibilities for input/ output. The capability of FICTION III is shown by means of two applications (UNi-test pin on position 100 and the third KNK fuel subassembly cladding failure). Object of further evaluations will be among other things the analysis of increased delayed neutron signals in regard to the fault location and dimension

Fuel pin bowing and related investigation of the gadolinium fuel pin influence on power release inside of neighbouring fuel pins in a WWER-440 type core

International Nuclear Information System (INIS)

Mikus, J.

2006-01-01

As known both the WWER-440 and WWER-1000 reactors are systematically modernized to enhance their safety and economical parameters of operation. For this purpose new fuel assemblies (FAs) were designed with improved technical parameters, e.g., containing fuel pins (FPs) in which Gd 2 O 3 burnable absorber is integrated into fuel. Presence of such FPs in reactor core results in a strong depression of thermal neutrons in their positions and corresponding high gradients in neighbouring FPs. Consequently, similar situation in neighbouring FPs can be expected as for both the power release and temperature gradients. The purpose of this work consists in investigation of the gadolinium FP influence on space power distribution, especially from viewpoint of the values and gradient occurrence inside of the neighbouring FPs that could result in static loads with some consequences, e.g., a contribution to FP/FA bowing. Since detailed power distributions cannot be obtained in the NPPs, needed information is provided by means of experiments on research reactors. As for the power release measurement inside of FPs, some special (e.g. track) detectors placed between fuel pellets are usually used. Since such works are relatively complicated and time consuming, an evaluation method based on mathematical modelling and numerical approximation was proposed by means of that, and using measured (integral) power release in selected FPs, needed power release values inside of investigated FPs, can be estimated. For this purpose, experimental results from light water, zero-power research reactor LR-0 obtained by measurements in a WWER-440 type core with 19 FAs at zero boron concentration and containing some FPs with gadolinium (Gd FPs) were utilized. Application of the proposed evaluation method is demonstrated on investigated FPs neighbouring a Gd FP by means of the: relative azimuthal power distribution estimation inside of investigated FPs on their fuel pellet surface in horizontal plane

Construction and tests of a gamma device for experimental measurements of burnup of nuclear reactor fuel

International Nuclear Information System (INIS)

Brandao Junior, F.A.

1982-01-01

The gamma-scanning method is an important tool for the measurement of burnup of nuclear reactor fuel. The adequate knowledge of burnup allows for a better inventory of 'sensitive' fissile materials, better fuel management and provides insight on fuel behaviour and safety margins. This paper is related to the description, construction and operation of a first gamma scanning device, tested by irradiation of prototype PWR fuel pins, 14 cm long, in a Triga Mark-I reactor at very low power. Despite the limitations imposed by the low burnup, the experiment permitted a good checking of the main physical concepts and devices involved in the method. (Author) [pt

Comparison of SCDAP/RELAP5/MOD3 to TRAC-PF1/MOD1 for timing analysis of PWR fuel pin failures

International Nuclear Information System (INIS)

Jones, K.R.; Katsma, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.

1991-01-01

A comparison has been made of SCDAP/RELAP5/MOD3- and TRAC-PF1/MOD1- based calculations of the fuel pin failure timing (time from containment isolation signal to first fuel pin failure) in a loss-of-coolant accident (LOCA). The two codes were used to calculate the thermal-hydraulic boundary conditions for a complete, double-ended, offset-shear break of a cold leg in a Westinghouse 4-loop pressurized water reactor. Both calculations used the FRAPCON-2 code to calculate the steady-state fuel rod behavior and the FRAP-T6 code to calculate the transient fuel rod behavior. The analysis was performed for 16 combinations of fuel burnups and power peaking factors extending up to the Technical Specifications limits. While all calculations were made on a best-estimate basis, the SCDAP/RELAP5/MOD3 code has not yet been fully assessed for large-break LOCA analysis. The results indicate that SCDAP/RELAP5/MOD3 yields conservative fuel pin failure timing results in comparison to those generated using TRAC-PF1/MOD1. 7 refs., 5 figs

Nuclear fuel element

International Nuclear Information System (INIS)

Yamamoto, Seigoro.

1994-01-01

Ultrafine particles of a thermal neutron absorber showing ultraplasticity is dispersed in oxide ceramic fuels by more than 1% to 10% or lower. The ultrafine particles of the thermal neutron absorber showing ultrafine plasticity is selected from any one of ZrGd, HfEu, HfY, HfGd, ZrEu, and ZrY. The thermal neutron absorber is converted into ultrafine particles and solid-solubilized in a nuclear fuel pellet, so that the dispersion thereof into nuclear fuels is made uniform and an absorbing performance of the thermal neutrons is also made uniform. Moreover, the characteristics thereof, for example, physical properties such as expansion coefficient and thermal conductivity of the nuclear fuels are also improved. The neutron absorber, such as ZrGd or the like, can provide plasticity of nuclear fuels, if it is mixed into the nuclear fuels for showing the plasticity. The nuclear fuel pellets are deformed like an hour glass as burning, but, since the end portion thereof is deformed plastically within a range of a repulsive force of the cladding tube, there is no worry of damaging a portion of the cladding tube. (N.H.)

Fuel pin transient behavior technology applied to safety analyses. Presentation to AEC Regulatory Staff 4th Regulatory Briefing on safety technology, Washington, D.C., November 19--20, 1974

International Nuclear Information System (INIS)

1974-11-01

Information is presented concerning LMFBR fuel pin performance requirements and evaluation; fuels behavior codes with safety interfaces; performance evaluations; ex-reactor materials and simulation tests; models for fuel pin failure; and summary of continuing fuels technology tasks. (DCC)

Innovate pin design for Sphere-pac fuel in sodium cooled fast reactors

International Nuclear Information System (INIS)

Pouchon, Manuel A.; Niceno, Bojan; Krepel, Jiri

2011-01-01

The paper discusses a new fuel element type, which combines a particle fuel concept, the Sphere-pac, with a new pin design which features internal cooling. Particle fuels are auspicious when considering a closed fuel cycle, where minor actinide containing fuels must be fabricated. The principle advantage lies in their production simplicity with much less maintenance intensive mechanical devices. Furthermore the Sphere-pac is usually produced by a wet and therefore powder-less route. Therefore the implementation in a remotely controlled and heavily shielded environment becomes easier to realize. Besides the advantages in the production process, the Sphere-pac bears one important disadvantage: the lower thermal conductivity of the particle arrangement, and the therefore higher peak temperatures in the fuel. Consequently a new fuel design is suggested in this paper. It offers an internal cooling channel and therefore smaller maximal fuel distances to the coolant. As the concept is new, the most important aspects are studied; these are the neutronics, the temperature profile in the fuel plus thermal-hydraulics aspects. (author)

The environmental impact of the nuclear fuel cycle

International Nuclear Information System (INIS)

Papp, R.

1976-01-01

Leaks in fuel-cans and in valve seals are the sources of nearly all the iodine and noble gases discharged from the reactor building's ventilation systems to the environment. Leakages play a greater role in BWR's than in PWR's due to the widespread primary circuit of BWR's. Another isotope of considerable concern during reactor operation is tritium which can be produced by different processes. The Zircaloy-cladding of LWR-fuel pins retains the tritium formed in the fuel much better than the stainless steel cladding of breeder reactors. The principal radionuclides discharged from reprocessing plants designed to process LWR uuel are 85 Kr, tritium, the iodines, and the actinides. Another isotope which should be considered because of its long half-live and its possible contribution to the whole-body dose is 14 C. The radionuclides released into the atmosphere in the form of gases and aerosols are diluted by the wind and decay by emitting β- and γ-radiation. This so-called submersion is the most important pathway of exposure during normal operation of nuclear facilities. (HP) [de

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydın; Kazimi, Mujid S.

2013-01-01

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydın, E-mail: karahan@alum.mit.edu; Kazimi, Mujid S.

2013-10-15

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

Nuclear Fuel Reprocessing

International Nuclear Information System (INIS)

Simpson, Michael F.; Law, Jack D.

2010-01-01

This is a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. Nuclear reprocessing is the chemical treatment of spent fuel involving separation of its various constituents. Principally, it is used to recover useful actinides from the spent fuel. Radioactive waste that cannot be re-used is separated into streams for consolidation into waste forms. The first known application of nuclear reprocessing was within the Manhattan Project to recover material for nuclear weapons. Currently, reprocessing has a peaceful application in the nuclear fuel cycle. A variety of chemical methods have been proposed and demonstrated for reprocessing of nuclear fuel. The two most widely investigated and implemented methods are generally referred to as aqueous reprocessing and pyroprocessing. Each of these technologies is described in detail in Section 3 with numerous references to published articles. Reprocessing of nuclear fuel as part of a fuel cycle can be used both to recover fissionable actinides and to stabilize radioactive fission products into durable waste forms. It can also be used as part of a breeder reactor fuel cycle that could result in a 14-fold or higher increase in energy utilization per unit of natural uranium. Reprocessing can also impact the need for geologic repositories for spent fuel. The volume of waste that needs to be sent to such a repository can be reduced by first subjecting the spent fuel to reprocessing. The extent to which volume reduction can occur is currently under study by the United States Department of Energy via research at various national laboratories and universities. Reprocessing can also separate fissile and non-fissile radioactive elements for transmutation.

Nuclear fuel elements

International Nuclear Information System (INIS)

Nakai, Keiichi

1983-01-01

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

Nuclear 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

Pin cell discontinuity factors in the transient 3-D discrete ordinates code TORT-TD - 237

International Nuclear Information System (INIS)

Seubert, A.

2010-01-01

This paper describes the application of generalized equivalence theory to the time-dependent 3-D discrete ordinates neutron transport code TORT-TD. The introduction of pin cell discontinuity factors into the discrete ordinates transport equation is described by assuming a linear dependence of the homogenized neutron angular flux within a pin cell which may be discontinuous at the interfaces to adjacent cells. The homogenized flux discontinuity at cell interfaces is expressed by pin cell discontinuity factors which in turn are determined from fuel assembly lattice calculations using HELIOS. Application of TORT-TD to the all rods in state of the PWR MOX/UO 2 Core Transient Benchmark with pin cell homogenized nuclear cross sections demonstrate the potential of pin cell discontinuity factors to reduce pin cell homogenization errors. (authors)

Investigation on macroscopic cross section model for BWR pin-by-pin core analysis - 118

International Nuclear Information System (INIS)

Fujita, T.; Tada, K.; Yamamoto, A.; Yamane, Y.; Kosaka, S.; Hirano, G.

2010-01-01

A cross section model used in the pin-by-pin core analysis for BWR is investigated. In the pin-by-pin core calculation method, pin-cell averaged cross sections are calculated for many combinations of state and history variables that have influences on the cross section and are tabulated prior to the core calculations. Variation of a cross section in a core simulator is classified into two different types, i.e., the instantaneous effect and the history effect. The instantaneous effect is incorporated by the variation of cross section which is caused by the instantaneous change of state variables. For this effect, the exposure, the void fraction, the fuel temperature, the moderator temperature and the control rod are used as indexes. The history effect is the cumulative effect of state variables. We treat this effect with a unified approach using the spectral history. To confirm accuracy of the cross section model, the pin-by-pin fission rate distribution and the k-infinity of fuel assembly which are obtained with the tabulated and the reference cross sections are compared. For the instantaneous effect, the present cross section model well reproduces the reference results for all off-nominal conditions. For the history effect, however, considerable differences both on the pin-by-pin fission rate distribution and the k-infinity are observed at high exposure points. (authors)

Nuclear power and the nuclear fuel cycle

Energy Technology Data Exchange (ETDEWEB)

NONE

1976-07-01

The IAEA is organizing a major conference on nuclear power and the nuclear fuel cycle, which is to be held from 2 to 13 May 1977 in Salzburg, Austria. The programme for the conference was published in the preceding issue of the IAEA Bulletin (Vol.18, No. 3/4). Topics to be covered at the conference include: world energy supply and demand, supply of nuclear fuel and fuel cycle services, radioactivity management (including transport), nuclear safety, public acceptance of nuclear power, safeguarding of nuclear materials, and nuclear power prospects in developing countries. The articles in the section that follows are intended to serve as an introduction to the topics to be discussed at the Salzburg Conference. They deal with the demand for uranium and nuclear fuel cycle services, uranium supplies, a computer simulation of regional fuel cycle centres, nuclear safety codes, management of radioactive wastes, and a pioneering research project on factors that determine public attitudes toward nuclear power. It is planned to present additional background articles, including a review of the world nuclear fuel reprocessing situation and developments in the uranium enrichment industry, in future issues of the Bulletin. (author)

Cell for receipting and dismantling nuclear fuel assembly

International Nuclear Information System (INIS)

Beneck, J.A.; Quayre, C.

1989-01-01

The cell has a vertical structure with a right section corresponding at that of the assembly to receive, a mechanism for keeping fuel pins at their nominal separation in the form of at least two combs and mechanisms of holding grids and bottom nozzle. The comb arrangements are moved into position by hydraulic actuators so that they cross each other to form a lattice round the fuel pins. The mechanism for holding grid assemblies consist of joints that articulate from a free position to a position where the joints press of the grid on all sides [fr

Nuclear power generation and nuclear fuel

International Nuclear Information System (INIS)

Okajima, Yasujiro

1985-01-01

As of June 30, 1984, in 25 countries, 311 nuclear power plants of about 209 million kW were in operation. In Japan, 27 plants of about 19 million kW were in operation, and Japan ranks fourth in the world. The present state of nuclear power generation and nuclear fuel cycle is explained. The total uranium resources in the free world which can be mined at the cost below $130/kgU are about 3.67 million t, and it was estimated that the demand up to about 2015 would be able to be met. But it is considered also that the demand and supply of uranium in the world may become tight at the end of 1980s. The supply of uranium to Japan is ensured up to about 1995, and the yearly supply of 3000 st U 3 O 8 is expected in the latter half of 1990s. The refining, conversion and enrichment of uranium are described. In Japan, a pilot enrichment plant consisting of 7000 centrifuges has the capacity of about 50 t SWU/year. UO 2 fuel assemblies for LWRs, the working of Zircaloy, the fabrication of fuel assemblies, the quality assurance of nuclear fuel, the behavior of UO 2 fuel, the grading-up of LWRs and nuclear fuel, and the nuclear fuel business in Japan are reported. The reprocessing of spent fuel and plutonium fuel are described. (Kako, I.)

UO2 - Zr chemical interaction of PHWR fuel pins under high temperature

International Nuclear Information System (INIS)

Majumdar, P.; Mukhopadhyay, D.; Gupta, S.K.

2001-01-01

At high temperature Zircaloy clad interacts with the UO 2 fuel as well as with the steam to produce oxide layer of a-Zr(O) and ZrO 2 . This layer formation significantly reduces the structural strength of the clad. A computer code SFDCPA/MOD1 has been developed to simulate the interaction and predict the oxide layer thickness for any accidental transient condition. It is well validated with published experimental data on the isothermal and transient temperature condition. The program is applied to Indian Pressurized Heavy Water Reactor (PHWR) fuel pin under certain severe transient condition where it experiences temperature above 1000 C. The study gives an idea of the un-oxidized thickness of Zircaloy, which is an important criterion for fuel integrity. (author)

Role of ion chromatograph in nuclear fuel fabrication process at Nuclear Fuel Complex

International Nuclear Information System (INIS)

Balaji Rao, Y.; Prasada Rao, G.; Prahlad, B.; Saibaba, N.

2012-01-01

The present paper discusses the different applications of ion chromatography followed in nuclear fuel fabrication process at Nuclear Fuel Complex. Some more applications of IC for characterization of nuclear materials and which are at different stages of method development at Control Laboratory, Nuclear Fuel Complex are also highlighted

Fuel assembly for FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki.

1995-01-01

Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

Nuclear reactor fuel element assembly spacer grid and method of making

International Nuclear Information System (INIS)

Chetter, J.

1975-01-01

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

Nuclear fuel cycle system analysis

International Nuclear Information System (INIS)

Ko, W. I.; Kwon, E. H.; Kim, S. G.; Park, B. H.; Song, K. C.; Song, D. Y.; Lee, H. H.; Chang, H. L.; Jeong, C. J.

2012-04-01

The nuclear fuel cycle system analysis method has been designed and established for an integrated nuclear fuel cycle system assessment by analyzing various methodologies. The economics, PR(Proliferation Resistance) and environmental impact evaluation of the fuel cycle system were performed using improved DB, and finally the best fuel cycle option which is applicable in Korea was derived. In addition, this research is helped to increase the national credibility and transparency for PR with developing and fulfilling PR enhancement program. The detailed contents of the work are as follows: 1)Establish and improve the DB for nuclear fuel cycle system analysis 2)Development of the analysis model for nuclear fuel cycle 3)Preliminary study for nuclear fuel cycle analysis 4)Development of overall evaluation model of nuclear fuel cycle system 5)Overall evaluation of nuclear fuel cycle system 6)Evaluate the PR for nuclear fuel cycle system and derive the enhancement method 7)Derive and fulfill of nuclear transparency enhancement method The optimum fuel cycle option which is economical and applicable to domestic situation was derived in this research. It would be a basis for establishment of the long-term strategy for nuclear fuel cycle. This work contributes for guaranteeing the technical, economical validity of the optimal fuel cycle option. Deriving and fulfillment of the method for enhancing nuclear transparency will also contribute to renewing the ROK-U.S Atomic Energy Agreement in 2014

Nuclear fuel lease accounting

International Nuclear Information System (INIS)

Danielson, A.H.

1986-01-01

The subject of nuclear fuel lease accounting is a controversial one that has received much attention over the years. This has occurred during a period when increasing numbers of utilities, seeking alternatives to traditional financing methods, have turned to leasing their nuclear fuel inventories. The purpose of this paper is to examine the current accounting treatment of nuclear fuel leases as prescribed by the Financial Accounting Standards Board (FASB) and the Federal Energy Regulatory Commission's (FERC's) Uniform System of Accounts. Cost accounting for leased nuclear fuel during the fuel cycle is also discussed

Boosting nuclear fuels

International Nuclear Information System (INIS)

Demarthon, F.; Donnars, O.; Dupuy-Maury, F.

2002-01-01

This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UO x and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)

Nuclear fuel pellet loading apparatus

International Nuclear Information System (INIS)

Gerkey, K.S.

1979-01-01

An automatic apparatus for loading a predetermined amount of nuclear fuel pellets into a nuclear fuel element to be used in a nuclear reactor is described. The apparatus consists of a vibratory bed capable of supporting corrugated trays containing rows of nuclear fuel pellets and arranged in alignment with the open ends of several nuclear fuel elements. A sweep mechanism is arranged above the trays and serves to sweep the rows of fuel pellets onto the vibratory bed and into the fuel element. A length detecting system, in conjunction with a pellet stopping mechanism, is also provided to assure that a predetermined amount of nuclear fuel pellets are loaded into each fuel element

Nuclear fuel

International Nuclear Information System (INIS)

Azevedo, J.B.L. de.

1980-01-01

All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.) [pt

Nuclear power and the nuclear fuel cycle

International Nuclear Information System (INIS)

Hardy, C.J.; Silver, J.M.

1985-09-01

The report provides data and assessments of the status and prospects of nuclear power and the nuclear fuel cycle. The report discusses the economic competitiveness of nuclear electricity generation, the extent of world uranium resources, production and requirements, uranium conversion and enrichment, fuel fabrication, spent fuel treatment and radioactive waste management. A review is given of the status of nuclear fusion research

Study on velocity field in a wire wrapped fuel pin bundle of sodium cooled reactor. Detailed velocity distribution in a subchannel

International Nuclear Information System (INIS)

Sato, Hiroyuki; Kobayashi, Jun; Miyakoshi, Hiroyuki; Kamide, Hideki

2009-01-01

A sodium cooled fast reactor is designed to attain a high burn-up core in a feasibility study on commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is of importance to obtain the flow velocity distribution in a wire wrapped pin bundle. A 2.5 times enlarged 7-pin bundle water model was applied to investigate the detailed velocity distribution in an inner subchannel surrounded by 3 pins with wrapping wire. The test section consisted of a hexagonal acrylic duct tube and fluorinated resin pins which had nearly the same refractive index with that of water and a high light transmission rate. The velocity distribution in an inner subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through the front and lateral sides of the duct tube. In the vertical velocity distribution in a narrow space between the pins, the wrapping wire decreased the velocity downstream of the wire and asymmetric flow distribution was formed between the pin and wire. In the horizontal velocity distribution, swirl flow around the wrapping wire was obviously observed. The measured velocity data are useful for code validation of pin bundle thermalhydraulics. (author)