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Sample records for lmfbr fuel pins

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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.

  6. Evaluation of integrally finned cladding for LMFBR fuel pins

    International Nuclear Information System (INIS)

    Cantley, D.A.; Sutherland, W.H.

    1975-01-01

    An integral fin design effectively reduces the coolant temperature gradients within an LMFBR subassembly by redistributing coolant flow so as to reduce the maximum cladding temperature and increase the duct wall temperature. The reduced cladding temperatures are offset by strain concentrations resulting from the fin geometry, so there is little net effect on predicted fuel pin performance. The increased duct wall temperatures, however, significantly reduce the duct design lifetime so that the final conclusion is that the integral fin design is inferior to the standard wire wrap design. This result, however, is dependent upon the material correlations used. Advanced alloys with improved irradiation properties could alter this conclusion

  7. 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.)

  8. 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

  9. 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

  10. 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

  11. 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

  12. Performance of LMFBR fuel pins with (Pu,Th)O/sub 2-x/ and UO2

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1983-09-01

    The irradiation performance of (Pu,Th)O/sub 2-x/ and UO 2 fueled pins for breeder reactor application were compared to the extensive performance data base for the (U,Pu)O/sub 2-x/ fuel system. Th-Pu and 238 U- 233 U based fuel systems were candidate fuel fertile/fissile isotopic combinations for development of alternatives to the current LMFBR fuel cycle. Initial screening tests were conducted in the EBR-II to obtain comparative performance data because of the limited experience with these fuel systems. In some cases, 235 U was used as a substitute for 233 U because of the difficulties in fabrication of available 233 U due to its high gamma ray emission rate

  13. 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

  14. 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.

  15. 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

  16. 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.

  17. 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

  18. A model for pressure in an LMFBR duct due to discharge of gas from a failed fuel pin

    International Nuclear Information System (INIS)

    Srinivas, S.; Chopra, P.S.

    1977-01-01

    In this paper an analytical model for the calculation of pressure pulses in hexagonal ducts due to discharge of gas from a failed fuel pin is developed. The analysis yields the time history of the pressure pulse which can be used in the calculation of permanent deformation of the duct or in the assessment of the susceptibility of the duct to fracture. The real physical situation of gas discharging through a pin in a duct filled with liquid is complex to model. Here the phenomenon is modeled based on some reasonable assumptions. In this model the analysis is divided into two stages. In the first stage the gas expands as a spherical bubble, but the influence of the duct wall is taken into account. At the end of the first stage the spherical shape of the bubble is assumed to be lost and the gas is assumed to expand axially as a column. The analysis involves solving the continuity and momentum equations for the liquid along with the energy balance equation for the gas

  19. LMFBR fuel analysis. Task A: oxide fuel dynamics. Final report, October 1977--September 1978

    International Nuclear Information System (INIS)

    Dhir, V.K.; Frank, M.; Kastenberg, W.E.; McKone, T.E.

    1979-03-01

    Three aspects of LMFBR safety are discussed. The first concerns the potential reactivity effects of whole core fuel motion prior to pin failure in low ramp rate transient overpower accidents. The second concerns the effects of flow blockages following pin failure on the coolability of a core following an unprotected overpower transient. The third aspect concerns the safety related implications of using thorium based fuels in LMFBR's

  20. 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

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

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

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

  2. 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.

  3. Internal fuel pin oxidizer

    International Nuclear Information System (INIS)

    Andrews, M.G.

    1978-01-01

    A nuclear fuel pin has positioned within it material which will decompose to release an oxidizing agent which will react with the cladding of the pin and form a protective oxide film on the internal surface of the cladding

  4. 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

  5. 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

  6. 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)

  7. Strategies in development of advanced fuels for LMFBR

    International Nuclear Information System (INIS)

    Handa, Muneo

    1976-12-01

    Overseas strategies in development of advanced fuels for LMFBR are reviewed. Recent irradiation experiment and out-of-pile test data of the fuels are given in detail. The present status of development of oxide fueled LMFBR is also treated. (auth.)

  8. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. (Auth.)

  9. Fueling method in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Inoue, Kotaro.

    1985-01-01

    Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

  10. Nuclear fuel pin

    International Nuclear Information System (INIS)

    Hartley, Kenneth; Moulding, T.L.J.; Rostron, Norman.

    1979-01-01

    Fuel pin for use in fast breeder nuclear reactors containing fissile and fertile areas of which the fissile and fertile materials do not mix. The fissile material takes the shape of large and small diameter microspheres (the small diameter microspheres can pass through the interstices between the large microspheres). The barrier layers being composed of microspheres with a diameter situated between those of the large and small microspheres ensure that the materials do not mix [fr

  11. 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

  12. Synthesis Report on the understanding of failed LMFBR fuel element performance

    International Nuclear Information System (INIS)

    Plitz, H.; Bagley, K.; Harbourne, B.

    1990-07-01

    In the coarse of LMFBR operation fuel element failures cannot entirely be avoided as experienced during the operation of PFR, PHENIX and KNK II, where 44 failed fuel elements have been registered between 1978 and 1989. In earlier irradiations, post irradiation examinations showed mixed oxide pin diameter increases up to pin pitch distance, urging to stress reactor safety questions on the potential of fuel pin failure propagation within pin bundles. The chemical interaction of sodium with mixed oxide fuel is regarded to be the key for the understanding of failed fuel behavior. Valuable results on the failed fuel pin behavior during operation were obtained from the SILOE sodium loop test. Based on the bulk of experience with the detection of fuel pin failures, with the continued operation and with the handling of failed pins respectively elements, one can state: 1. All fuel pin failures have been detected securely in time and have been located. 2. Small defects are developing slowly. 3. Even large defects at end-of-life pins resulted in limited fuel loss. 4. Clad failures behave benign in main aspects. 5. The chemical interaction of sodium with mixed oxide is an important factor in the behavior of failed fuel pins, especially at high burnup. 6. Despite different pin designs and different operation conditions, on the basis of 44 failed elements in PFR, PHENIX and KNK II no pin-to-pin propagation was observed and fuel release was rather low, often not detectable. 7. In no case hazard conditions affecting reactor safety have been experienced

  13. Work plan: transient release from LMFBR fuel

    International Nuclear Information System (INIS)

    Kress, T.S.; Parker, G.W.; Fontana, M.H.

    1975-09-01

    The proposed LMFBR Transient Release Program at ORNL is designed to investigate, by means of ex-reactor experiments and analytical modeling, the release and transport of fuel, fission products, and transuranic elements from fast reactor cores in the event of certain hypothetical accidents. It is desired to experimentally produce energy depositions that are characteristic of severe hypothetical reactor transients by the application of direct electrical current to mixed-oxide fuels under sodium. The experimental program includes tests with and without sodium, investigations of alternative methods of generating fuel and sodium aerosols, the use of UO 2 as a fuel simulant, additions of tracers as fission product simulants, effects of radiation, and under-water and under-sodium efforts to study the behavior of the vapor bubble itself. Analytical modeling will accompany all phases of the program, and the data will be correlated with models developed. 21 references. (auth)

  14. Automated fuel pin loading system

    Science.gov (United States)

    Christiansen, D.W.; Brown, W.F.; Steffen, J.M.

    An automated loading system for nuclear reactor fuel elements utilizes a gravity feed conveyor which permits individual fuel pins to roll along a constrained path perpendicular to their respective lengths. The individual lengths of fuel cladding are directed onto movable transports, where they are aligned coaxially with the axes of associated handling equipment at appropriate production stations. Each fuel pin can be be reciprocated axially and/or rotated about its axis as required during handling steps. The fuel pins are inerted as a batch prior to welding of end caps by one of two disclosed welding systems.

  15. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. First, a thermal creep damage index is set up through a sufficiently sophisticated clad physical analysis including arbitrary time dependence of power and neutron flux as well as effects of sodium temperature, burnup and steel mechanical behavior. Although this strain limit approach implies a more general but time consuming model., on the counterpart the net output is improved and e.g. clad temperature, stress and strain maxima may be easily assessed. A full spectrum of variables are statistically treated to account for their probability distributions. Creep damage probability may be obtained and can contribute to a quantitative fuel probability estimation

  16. Advanced methods for fabrication of PHWR and LMFBR fuels

    International Nuclear Information System (INIS)

    Ganguly, C.

    1988-01-01

    For self-reliance in nuclear power, the Department of Atomic Energy (DAE), India is pursuing two specific reactor systems, namely the pressurised heavy water reactors (PHWR) and the liquid metal cooled fast breeder reactors (LMFBR). The reference fuel for PHWR is zircaloy-4 clad high density (≤ 96 per cent T.D.) natural UO 2 pellet-pins. The advanced PHWR fuels are UO 2 -PuO 2 (≤ 2 per cent), ThO 2 -PuO 2 (≤ 4 per cent) and ThO 2 -U 233 O 2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O 2 pellets clad in SS 316 or D9 is the reference fuel for the first generation of prototype and commercial LMFBRs all over the world. However, (UPu)C and (UPu)N are considered as advanced fuels for LMFBRs mainly because of their shorter doubling time. The conventional method of fabrication of both high and low density oxide, carbide and nitride fuel pellets starting from UO 2 , PuO 2 and ThO 2 powders is 'powder metallurgy (P/M)'. The P/M route has, however, the disadvantage of generation and handling of fine powder particles of the fuel and the associated problem of 'radiotoxic dust hazard'. The present paper summarises the state-of-the-art of advanced methods of fabrication of oxide, carbide and nitride fuels and highlights the author's experience on sol-gel-microsphere-pelletisation (SGMP) route for preparation of these materials. The SGMP process uses sol gel derived, dust-free and free-flowing microspheres of oxides, carbide or nitride for direct pelletisation and sintering. Fuel pellets of both low and high density, excellent microhomogeneity and controlled 'open' or 'closed' porosity could be fabricated via the SGMP route. (author). 5 tables, 14 figs., 15 refs

  17. Feasibility study for adapting ITREC plant to reprocessing LMFBR fuels

    International Nuclear Information System (INIS)

    Moccia, A.; Rolandi, G.

    1976-05-01

    The report evaluates the feasibility of adapting ITREC plant to the reprocessing LMFBR fuels, with the double purpose of: 1) recovering valuable Pu contained in these fuels and recycling it to the fabrication plant; 2) trying, on a pilot scale, the chemical process technology to be applied in a future industrial plant for reprocessing the fuel elements discharged from fast breeder power reactors

  18. Suspension scheme for fuel pin

    International Nuclear Information System (INIS)

    Butts, C.E.; Gray, H.C.

    1975-01-01

    A description is presented of a nuclear fuel pin suspension arrangement comprising, in combination, a rod; a first beam member connected to said rod at one end; a plurality of parallel-spaced slidable fuel support plates attached to said first beam member, the longitudinal axis of first beam member being perpendicular to the longitudinal axis of each of said fuel support plates, a first coupling means disposed along the length of the first beam member for permitting slidable fuel support plates parallel movement with respect to the longitudinal axis of said first beam member, a second coupling means located at one end of each of slidable fuel plates for slidably engaging first coupling means of first beam member, a second beam member connected to the other end of each of parallel-spaced slidable fuel support plates and providing an extension, second beam member being provided with a third coupling means disposed along the length of second beam member at one end thereof; and a plurality of fuel pins provided with a fourth coupling means located at one end of each fuel pin for slidably engaging third coupling means of second beam member to permit each fuel pin parallel movement with respect to the longitudinal axis of second beam member. (U.S.)

  19. Blockages in LMFBR fuel assemblies: a review

    International Nuclear Information System (INIS)

    Han, J.T.; Fontana, M.H.

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions

  20. Blockages in LMFBR fuel assemblies: a review

    Energy Technology Data Exchange (ETDEWEB)

    Han, J T; Fontana, M H

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions.

  1. Airborne effluent control for LMFBR fuel reprocessing plants

    International Nuclear Information System (INIS)

    Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-01-01

    A significant part of the LMFBR fuel reprocessing development program has been devoted to the development of efficient removal systems for the volatile fission products, including 131 I, krypton, tritium, 129 I, and most recently 14 C. Flowsheet studies have indicated that very significant reductions of radioactive effluents can be achieved by integrating advanced effluent control systems with new concepts of containment and ventilation; however, the feasibility of such has not yet been established, nor have the economics been examined. This paper presents a flowsheet for the application of advanced containment systems to the processing of LMFBR fuels and summarizes the status and applicability of specific fission product removal systems

  2. 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.)

  3. Biological behavior of mixed LMFBR-fuel-sodium aerosols

    International Nuclear Information System (INIS)

    Mahlum, D.D.; Hackett, P.L.; Hess, J.O.; Allen, M.D.

    1979-01-01

    Immediately after exposure of rats to mixed aerosols of sodium-LMFBR fuel, about 80 to 90% of the body burden of 239 Pu is in the gastrointestinal tract; 1.5 to 4% is in the lungs. With fuel-only aerosols, less of the body burden was in the GI tract and more in the lung and the head. Blood and urine values suggest an increased absorption of 239 Pu from sodium-fuel than from fuel-only aerosols

  4. Automated system for loading nuclear fuel pins

    International Nuclear Information System (INIS)

    Marshall, J.L.

    1983-10-01

    A completely automatic and remotely controlled fuel pin fabrication system is being designed by the Westinghouse Hanford Company. The Pin Operations System will produce fuel pins for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor Plant (CRBRP). The system will assemble fuel pin components into cladding tubes in a controlled environment. After fuel loading, the pins are filled with helium, the tag gas capsules are inserted, and the top end cap welded. Following welding, the pins are surveyed to assure they are free of contamination and then the pins are helium leak tested

  5. 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

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

    International Nuclear Information System (INIS)

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

    1985-04-01

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

  7. Detailed design consideration on wire-spaced LMFBR fuel subassemblies under the effects of uncertainties and non-nominal geometries

    International Nuclear Information System (INIS)

    Hishida, H.

    1979-01-01

    This paper explains some analytical methods for evaluating the effects of deviation in subchannel coolant flow rate from the nominal value due to fuel pin bundle deflection and manufacturing tolerances and of inter-sub-channel coolant mixing and local temperature rise due to a wire-spacer on the hot spot temperature. Numerical results are given in each chapter with respect to a prototype LMFBR core. (author)

  8. 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

  9. LMFBR fuel cycle studies progress report, August 1972, No. 42

    International Nuclear Information System (INIS)

    Unger, W.E.; Blanco, R.E.; Crouse, D.J.; Irvine, A.R.; Watson, C.D.

    1972-10-01

    This report continues a series outlining progress in the development of methods for reprocessing of LMFBR fuels. Development work is reported on problems of irradiated fuel transport to the processing facility, the dissolution of the fuel and the chemical recovery of PuO 2 --UO 2 values, the containment of volatile fission products, product purification, conversion of fuel processing plant product nitrate solutions to solids suitable for shipping and for subsequent fuel fabrication. Pertinent experimental results are presented for the information of those immediately concerned with the field. Detailed description of experimental work and data are included in the topical reports and in the Chemical Technology Division Annual Reports

  10. Retention of gaseous fission products in reprocessing LMFBR fuels

    International Nuclear Information System (INIS)

    Burch, W.D.; Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-05-01

    The report is devoted to status of the development programme at the Oak Ridge National Laboratory on methods for retaining iodine-131 and 129, Krypton-85, Tritium and Carbon-14 in reprocessing LMFBR fuels. The Iodox process, Fluorocarbon absorption process and Voloxidation process are described for retention of iodine, Krypton-85 and Tritium, respectively. Flowsheets for the different processes are given and results of experimental runs in small engineering-scale equipment are reported

  11. Stress relaxation of thermally bowed fuel pins

    International Nuclear Information System (INIS)

    Crossland, I.G.; Speight, M.V.

    1983-01-01

    The presence of cross-pin temperature gradients in nuclear reactor fuel pins produces differential thermal expansion which, in turn, causes the fuel pin to bow elastically. If the pin is restrained in any way, such thermal bowing causes the pin to be stressed. At high temperatures these stresses can relax by creep and it is shown here that this causes the pin to suffer an additional permanent deflection, so that when the cross-pin temperature difference is removed the pin remains bowed. By representing the cylindrical pin by an equivalent I-beam, the present work examines this effect when it takes place by secondary creep. Two restraint systems are considered, and it is demonstrated that the rate of relaxation depends mainly upon the creep equation, and hence the temperature, and also the magnitude of the initial stresses. (author)

  12. 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.)

  13. 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

  14. CAPRICORN subchannel code for sodium boiling in LMFBR fuel bundles

    International Nuclear Information System (INIS)

    Padilla, A. Jr.; Smith, D.E.; O'Dell, L.D.

    1983-01-01

    The CAPRICORN computer code analyzes steady-state and transient, single-phase and boiling problems in LMFBR fuel bundles. CAPRICORN uses the same type of subchannel geometry as the COBRA family of codes and solves a similar system of conservation equations for mass, momentum, and energy. However, CAPRICORN uses a different numerical solution method which allows it to handle the full liquid-to-vapor density change for sodium boiling. Results of the initial comparison with data (the W-1 SLSF pipe rupture experiment) are very promising and provide an optimistic basis for proceeding with further development

  15. 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

  16. Dissolution of LMFBR fuel-sodium aerosols

    International Nuclear Information System (INIS)

    Allen, M.D.; Moss, O.R.

    1979-01-01

    Plutonium dioxide, normally insoluble in biological fluids, becomes much more soluble when mixed with sodium as the aerosol is formed. Sodium-fuel aerosols are approximately 20 times less soluble in simulated lung fluid than in distilled water. Solubility of sodium-fuel aerosols increases when Na 2 CO 3 are added to the distilled-water dissolution fluid. Mixed-oxide fuel aerosols without sodium present are relatively insoluble in distilled water, simulated lung fluid, and distilled water with Na 2 CO 3 and NaHCO 3 added

  17. Failed fuel detection and location of LMFBR

    International Nuclear Information System (INIS)

    Mimoto, Yasuhide; Hukuda, Tooru; Nakamoto, Koichiro

    1974-01-01

    This is a summary report on Failed Fuel Detection and Location Methods of liquid metal cooled fast breeder reactors, and describes an outline of related research and development conducted by PNC. (auth.)

  18. Possible application of nonredundant pinhole arrays to fuel pin imaging

    International Nuclear Information System (INIS)

    Berzins, G.J.; Han, K.S.

    1975-11-01

    LMFBR Safety Test Facility imaging experiments rely on emission of radiation by the fuel pins and thus appear to strongly complement radiographic techniques in that they are most employable during peak excursion--a time of least favorable radiographic signal--to--noise ratio. Radiography, on the other hand, can provide information long before or after the excursion--times of below threshold signal for direct imaging techniques. An underlying premise of any imaging experiment is that, in addition to sufficient brightness, sufficient contrast exists in the scene. A further restriction is imposed by intervening materials, such as the wall of a containment vessel, that not only absorb but also scatter the radiation. These questions are approached by examining the properties of potential recording instrumentation, of pinhole apertures, and of the necessary radiation sources

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

    International Nuclear Information System (INIS)

    Betten, P.R.; Tow, D.M.

    1984-04-01

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

  20. 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

  1. Nuclear fuel pin controlled failure device

    International Nuclear Information System (INIS)

    Schlenker, L.D.

    1975-01-01

    Each fuel pin of a fuel assembly for a water-cooled nuclear reactor is provided with means for rupturing the cladding tube at a predetermined location if an abnormal increase in pressure of the gases present occurs due to a loss-of-coolant accident. Preferably all such rupture means are oriented to minimize the hydraulic resistance to the flow of emergency core coolant such as all rupture means pointing in the same direction. Rupture means may be disposed at different elevations in adjacent fuel pins and, further, fuel pins may be provided with two or more rupture means, one of which is in the upper portion of the fuel pin. Rupture means are mechanical as by providing a locally weakened condition of a controlled nature in the cladding. (U.S.)

  2. Ultrasonic inspections of fuel alignment pins

    International Nuclear Information System (INIS)

    Rathgeb, W.; Schmid, R.

    1994-01-01

    As a remedy to the practical problem of defects in fuel alignment pins made of Inconel X750, an inspection technique has been developed which fully meets the requirements of detecting defects. The newly used fuel alignment pins made of austenite are easy to test and therefore satisfy the necessity of further inspections.For the fuel alignment pins of the upper core structure a safe and fast inspection technique was made available. The inspection sensitivity is high and it is possible to give quantitative directions concerning defect orientation and depth. After the required inspections had been concluded in 1989, a total of 18 inspections were carried out in various national and international nuclear power plants in the following years. During this time more than 6000 fuel alignment pines were examined.For the fuel alignment pins the inspection technique provided could increase the understanding of the defect process. This technique contributed to the development of an adaptive and economical repair strategy. ((orig.))

  3. Internal fuel motion as an inherent shutdown mechanism for LMFBR accidents: PINEX-3, PINEX-2, and HUT 5-2A experiments

    International Nuclear Information System (INIS)

    Ferrell, P.C.; Porten, D.R.; Martin, F.J.

    1981-01-01

    The PINEX-2 experiment verified the concept of axial internal molten fuel motion within annular fuel, representing an inherent shutdown mechanism for hypothetical transient overpower excursions on the order of 5$/s. The PINEX-3 experiment, simulating a 50 cents/s transient overpower, showed that limitations on the effectiveness of fuel motion may arise from freezing of the fuel and blockage of the internal movement. Analysis of these experiments was performed to assess the physical processes that dominate fuel relocation potential and to apply them to prototypic LMFBR pin conditions. Results indicate that internal fuel motion should be reliable as a shutdown mechanism in LMFBR's for a range of reactivity insertion rates beyond presently available experimental data

  4. Peripheral pin alignment system for fuel assemblies

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1981-01-01

    An alignment system is provided for nuclear fuel assemblies in a nuclear core. The core support structure of the nuclear reactor includes upwardly pointing alignment pins arranged in a square grid and engage peripheral depressions formed in the lateral periphery of the lower ends of each of the fuel assemblies of the core. In a preferred embodiment, the depressions are located at the corners of the fuel assemblies so that each depression includes one-quarter of a cylindrical void. Accordingly, each fuel assembly is positioned and aligned by one-quarter of four separate alignment pins which engage the fuel assemblies at their lower exterior corners. (author)

  5. Natural circulation in simulated LMFBR fuel assemblies

    International Nuclear Information System (INIS)

    Levin, A.E.; Carbajo, J.J.; Lloyd, D.B.; Montgomery, B.H.; Rose, S.D.; Wantland, J.L.

    1985-01-01

    Natural circulation experiments have been performed using simulated liquid metal fast breeder reactor fuel assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale sodium loop. Objective of these tests has been to provide experimental data under conditions that might be encountered during a partial or total loss of the shutdown heat removal system (SHRS) in a reactor. The experiments have included single- and two-phase tests under quasi-steady and transient conditions, at both nominal and non-nominal system conditions. Results from these test indicate that the potential for reactor damage during degraded SHRS operation is extremely slight, and that natural circulation can be a major contributor to safe operation of the system in both single- and two-phase flow during such operation

  6. 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

  7. Thermohydraulic and thermal stress aspects of a porous blockage in an LMFBR fuel assembly

    International Nuclear Information System (INIS)

    Kuzay, T.M.; Marr, W.W.; Helenberg, H.W.; Ariman, T.; Wilson, R.E.; Pedersen, D.R.

    1979-01-01

    The current safety scenarios of Liquid Metal Fast Breeder Reactors (LMFBR) under local fault propagation include the study of a hypothetical accident initiated by the formation of an external debris porous blockage in a fuel subassembly. In this preliminary experimental and analytical investigation, a non-heat-generating porous blockage was postulated to cover 18 flow channels of a 37 pin Fast Test Reactor (FTR) type fuel subassembly. The axial extent of the blockage is 50 mm. The blockage material is stainless steel (SS 316) with 30 percent average porosity (percent void volume). The blockage and the pins were modeled with a finite element technique and the thermal field in the blockage was predicted. This thermal field was utilized to do a planar thermal stress analysis of the postulated blockage. To verify the analytical model and also to better understand the thermal-hydraulics of such a porous blockage out-of-pile tests were conducted in a sodium loop. Data from the out-of-pile tests was utilized to calibrate and improve the analytical model

  8. Overview of the fast reactors fuels program. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

  9. Fuel pin bowing in CAGR

    International Nuclear Information System (INIS)

    Crossland, I.G.

    1982-01-01

    Some of the more important mechanisms by which pin bowing can occur in Advanced Gas Cooled Reactors are examined. These include creep relaxation of the stresses which occur when thermal bowing is restrained and asymmetric axial clad creep. The clad temperature changes which accompany such bowing are also investigated and the theoretical results briefly compared with the empirical behaviour. (author)

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

    International Nuclear Information System (INIS)

    Thurnay, K.

    1984-02-01

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

  11. Fuel pins irradiation: experimental devices and analytical behaviour

    International Nuclear Information System (INIS)

    Lemaignan, C.

    1996-01-01

    In this text we present the general characteristics of adapted irradiation loops in research reactors and the main results that we can expected with these loops in the behaviour field of PWR and LMFBR fuels( fuel densification, fuel cladding interactions, fission products release, reactor accidents)

  12. A probabilistic design method for LMFBR fuel rods

    International Nuclear Information System (INIS)

    Peck, S.O.; Lovejoy, W.S.

    1977-01-01

    Fuel rod performance analyses for design purposes are dependent upon material properties, dimensions, and loads that are statistical in nature. Conventional design practice accounts for the uncertainties in relevant parameters by designing to a 'safety factor', set so as to assure safe operation. Arbitrary assignment of these safety factors, based upon a number of 'worst case' assumptions, may result in costly over-design. Probabilistic design methods provide a systematic way to reflect the uncertainties in design parameters. PECS-III is a computer code which employs Monte Carlo techniques to generate the probability density and distribution functions for time-to-failure and cumulative damage for sealed plenum LMFBR fuel rods on a single rod or whole core basis. In Monte Carlo analyses, a deterministic model (that maps single-valued inputs into single-valued outputs) is coupled to a statistical 'driver'. Uncertainties in the input are reflected by assigning probability densities to the input parameters. Dependent input variables are considered multivariate normal. Independent input variables may be arbitrarily distributed. Sample values are drawn from these input densities, and a complete analysis is done by the deterministic model to generate a sample point in the output distribution. This process is repeated many times, and the number of times each output value occurs is accumulated. The probability that some measure of rod performance will fall within given limits is estimated by the relative frequency with which the Monte Carlo samples fall within tho

  13. 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

    considerations: many problems concerning fuel pin behaviour which have been encountered when developing LMFBR , have been mostly solved by a judicious extrapolation of the previously obtained experimental results. In order to perform this type of extrapolation, we need, first of all, a correct analysis of the technological key-phenomena (e.g. clad deformation) as a function of the known irradiation parameters, and then the inventory of the other phenomena which may be implied in the above analysis in order to unknot the inter-linkage and cross effects. When no such effect exists, it is of no need for the desired extrapolation general and complex behaviour code. A simple calculation dealing with the few implied parameters is quite sufficient.

  14. 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

  15. Thermal evaluation facility for LMFBR spent fuel transport

    International Nuclear Information System (INIS)

    Wesley, D.A.

    1980-04-01

    A full-scale mock-up of a 217 pin breeder reactor fuel assembly in a cylindrical pipe was initially designed and constructed by Oak Ridge National Laboratory (ORNL). It was transferred to Sandia where it was extensively redesigned and modified. The 217 pin hexagonal core assembly was installed in a smaller diameter stainless steel pipe which more closely represents the diameter of a shipping canister or shipping cask basket wall. Two-hundred four of the tubes are electrically heated over an active length of 4-feet and the remaining thirteen are instrumented with multiple junction thermocouples which can be traversed axially. Thermocouples and heat-flux gauges are located on the hex core and canister perimeters at several axial locations

  16. PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

    International Nuclear Information System (INIS)

    Kirk, W.L.

    1976-08-01

    The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters

  17. PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kirk, W.L. (comp.)

    1976-08-01

    The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters.

  18. Upon local blockage formations in LMFBR fuel rod bundles with wire-wrapped spacers

    International Nuclear Information System (INIS)

    Minden, C. v.; Schultheiss, G.F.

    1982-01-01

    A theoretical and experimental study, to improve understanding of local particle depositions in a wire-wrapped LMFBR fuel bundle, has been performed. Theoretical considerations show, that a preferentially axial process of particle depositions occurs. The experiments confirm this and clarify that the blockages arise near the particle source and settle at the spatially arranged minimum gaps in the bundle. The results suggest that, considering flow reduction, cooling and DND-detection, such fuel particle blockages are less dangerous. With reference to these safety-relevant factors, wire-wrapped LMFBR fuel bundles seem to gain advantages compared to the grid design. (orig.) [de

  19. Safety consequences of local initiating events in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.; Wang, P.Y.

    1975-12-01

    The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered.

  20. Safety consequences of local initiating events in an LMFBR

    International Nuclear Information System (INIS)

    Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.; Wang, P.Y.

    1975-12-01

    The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered

  1. 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

  2. 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.)

  3. 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

  4. 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.

  5. A LMFBR for thorium utilization and for the U233/Th fuel rods specification

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.

    1982-01-01

    The use of U 233 /Th as fuel in the middle part of LMFBR core and the Pu/U in the external part of the core, are proposed. The basic neutronic and safety characteristics and the specifications of fuel rods to be used in the internal core, are presented. (E.G.) [pt

  6. 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

  7. Nuclear fuel assemblies and fuel pins usable in such assemblies

    International Nuclear Information System (INIS)

    Jolly, R.

    1982-01-01

    A novel end cap for a nuclear fuel assembly is described in detail. It consists of a trisection arrangement which is received within a cell of a cellular grid. The cell contains abutment means with which the trisection comes into abutment. The grid also contains an abutment means for preventing the trisections from being inserted into the cell in an incorrect orientation. The present design allows fuel pins to be securely held in a hold-down grid of a sub-assembly. The design also allows easier dis-assembly of the swollen and embrittled fuel pins prior to reprocessing. (U.K.)

  8. 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

  9. 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.)

  10. LMFBR fuel analysis. Task A: Oxide fuel dynamics. Final report, October 1, 1976--September 30, 1977

    International Nuclear Information System (INIS)

    Dhir, V.K.; Doshi, J.; Frank, M.; Hauss, B.; Kastenberg, W.E.; Wong, K.

    1977-10-01

    The study presented deals with several areas of uncertainty in the analysis of the unprotected overpower transient for the Clinch River Breeder Reactor. These areas of uncertainty include the time, place, and mode of fuel pin failure; pre-failure fuel motion; fuel freezing, plugging, and plate-out following pin failure; and the potential for re-criticality. Internal molten fuel motion prior to pin failure was found to be sensitive to ramp rate and burnup. The strain-limit fuel failure criterion was found to be inappropriate for analysis based on existing data. The coupling of pre-transient- and transient-induced stresses tended to force the failure location towards the core midplane

  11. 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

  12. The state of art of the methods for thermohydraulics design of LMFBR fuel elements

    International Nuclear Information System (INIS)

    Fernandez y Fernandez, E.; Carajilescov, P.

    1981-09-01

    The present (experimental and analytical) state of art of the methods for thermohydraulics design of LMFBR fuel elements is analyzed. A development program is suggested, in order to obtain a computer code for modelling the distribution of coolant enthalpy in reactor core. This computer code is in development. (Author) [pt

  13. Positioning and locking device for fuel pin to grid attachment

    International Nuclear Information System (INIS)

    Frick, T.M.; Wineman, A.L.

    1976-01-01

    A positioning and locking device for fuel pin to grid attachment provides an inexpensive means of positively positioning and locking the individual fuel pins which make up the driver fuel assemblies used in nuclear reactors. The device can be adapted for use with a currently used attachment grid assembly design and insures that the pins remain in their proper position throughout the in-reactor life of the assembly. This device also simplifies fuel bundle assembly in that a complete row of fuel pins can be added to the bundle during each step of assembly. 8 claims, 8 drawing figures

  14. LMFBR operational and experimental local-fault experience, primarily with oxide fuel elements

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS- and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation

  15. LMFBR operational and experimental in-core local-fault experience, primarily with oxide fuel elements

    International Nuclear Information System (INIS)

    Warinner, D.K.

    Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS-and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation

  16. Route survey for LMFBR spent fuel transportation analysis

    International Nuclear Information System (INIS)

    Foley, J.T.

    1977-05-01

    Descriptions are given of surveys that were made along segments of interstate highways to obtain information on objects near the right-of-ways and on highway features that constitute hazards in the event of transportation accidents. Data collected during the surveys are summarized. The work was done in support of the LMFBR Hazards Analysis which was being performed for the Division of Reactor Development and Demonstration of the U.S. Energy Research and Development Administration

  17. 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

  18. 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

  19. Deposition of inhaled LMFBR-fuel-sodium aerosols in beagle dogs

    International Nuclear Information System (INIS)

    Hackett, P.L.; Mahlum, D.D.; Briant, J.K.; Catt, D.L.; Peters, L.R.; Clary, A.J.

    1980-01-01

    Initial alveolar deposition of LMFBR-fuel aerosols in beagle dogs amounted to 30% of the inhaled activity, but only 5% of the total inhaled activity was deposited in dogs exposed to sodium-fuel aerosols. Aerosol deposition in the gastrointestinal tract amounted to 4% of the initial body burden of fuel-aerosol exposed dogs and 24% of the burden of animals receiving sodium-fuel aerosols. Preliminary analytical data for the dog exposures appear to agree with rodent data for deposition and distribution patterns of aerosols of similar sodium: fuel ratios

  20. 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

  1. 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

  2. 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.

  3. Benchmark physics experiment of metallic-fueled LMFBR at FCA. 2

    International Nuclear Information System (INIS)

    Iijima, Susumu; Oigawa, Hiroyuki; Ohno, Akio; Sakurai, Takeshi; Nemoto, Tatsuo; Osugi, Toshitaka; Satoh, Kunio; Hayasaka, Katsuhisa; Bando, Masaru.

    1993-10-01

    An availability of data and method for a design of metallic-fueled LMFBR is examined by using the experiment results of FCA assembly XVI-1. Experiment included criticality and reactivity coefficients such as Doppler, sodium void, fuel shifting and fuel expansion. Reaction rate ratios, sample worth and control rod worth were also measured. Analysis was made by using three-dimensional diffusion calculations and JENDL-2 cross sections. Predictions of assembly XVI-1 reactor physics parameters agree reasonably well with the measured values, but for some reactivity coefficients such as Doppler, large zone sodium void and fuel shifting further improvement of calculation method was need. (author)

  4. TREAT experimental data base regarding fuel dispersals in LMFBR loss-of-flow accidents

    International Nuclear Information System (INIS)

    Simms, R.; Fink, C.L.; Stanford, G.S.; Regis, J.P.

    1981-01-01

    The reactivity feedback from fuel relocation is a central issue in the analysis of loss-of-flow (LOF) accidents in LMFBRs. Fuel relocation has been studied in a number of LOF simulations in the TREAT reactor. In this paper the results of these tests are analyzed, using, as the principal figure of merit, the changes in equivalent fuel worth associated with the fuel motion. The equivalent fuel worth was calculated from the measured axial fuel distributions by weighting the data with a typical LMFBR fuel-worth function. At nominal power, the initial fuel relocation resulted in increases in equivalent fuel worth. Above nominal power the fuel motion was dispersive, but the dispersive driving forces could not unequivocally be identified from the experimental data

  5. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    International Nuclear Information System (INIS)

    Riley, D.R.; Dickson, P.W.

    1981-01-01

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle

  6. Report of the IAEA advisory group meeting on LMFBR fuel reprocessing

    International Nuclear Information System (INIS)

    1976-05-01

    A summary of the papers and discussions of the meeting is presented, reviewing the status of development in LMFBR fuel reprocessing and focusing attention on important problem areas. The following topics are discussed: Transport, storage and removal of sodium; decladding and shearing; dissolution; Purex process; fluoride volatility method; off-gas purification; waste disposal. Status reports of national programmes of Belgium, France, Federal Republic of Germany, Italy, Japan, United Kingdom, USSR and USA are included

  7. Reactor transients tests for SNR fuel elements in HFR reactor

    International Nuclear Information System (INIS)

    Plitz, H.

    1989-01-01

    In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

  8. Problems of heat transfer within the containing vessel of high performance LMFBR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Pope, R.B.; Gartling, D.K.; Schimmel, W.P. Jr.; Larson, D.W.

    1976-01-01

    A preliminary assessment of heat transfer problems internal to a LMFBR spent fuel shipping cask is reported. The assessment is based upon previous results obtained in full-scale, electrically heated mockups of an LMFBR assembly located in a containing pipe, and also upon analytical and empirical studies presented in this paper. It is shown that a liquid coolant will be required to adequately distribute the decay heat of short-cooled assemblies from the fuel region to the containing cask structure. Liquid sodium apparently provides the best heat transfer, and sufficient data are available to adequately model the heat transfer processes involved. Dowtherm A is the most efficient organic evaluated to date and presented in the open literature. Since the organic materials have high Prandtl and usually high Rayleigh numbers, natural convection is the predominant mode of heat transfer. It is shown that a more comprehensive understanding of the convective processes will be required before heat transfer with an organic coolant can be adequately modeled. However, in view of systems considerations, Dowtherm A should be further considered as an alternative to sodium for use as a LMFBR spent fuel shipping cask coolant

  9. Safety research needs for carbide and nitride fueled LMFBR's. Final report

    International Nuclear Information System (INIS)

    Kastenberg, W.E.

    1975-01-01

    The results of a study initiated at UCLA during the academic year 1974--1975 to evaluate and review the potential safety related research needs for carbide and nitride fueled LMFBR's are presented. The tasks included the following: (1) Review Core and primary system designs for any significant differences from oxide fueled reactors, (2) Review carbide (and nitride) fuel element irradiation behavior, (3) Review reactor behavior in postulated accidents, (4) Examine analytical methods of accident analysis to identify major gaps in models and data, and (5) Examine post accident heat removal. (TSS)

  10. ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wright, A.L.; Kress, T.S.; Smith, A.M.

    1980-01-01

    This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

  11. 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)

  12. UK irradiation experience relevant to advanced carbide fuel concepts for LMFBR's

    International Nuclear Information System (INIS)

    Bagley, K.Q.; Batey, W.; Paris, R.; Sloss, W.M.; Snape, G.P.

    1977-01-01

    Despite discouraging prognoses of fabrication and reprocessing problems, it is recognized that the quest for a carbide fuel pin design which fully exploits the favourable density and thermal conductivity of (U,Pu) monocarbide must be maintained. Studies in aid of carbide fuel development have, therefore, continued in the UK in parallel with those on oxide, albeit at a substantially lower level of effort, and a sufficient body of irradiation experience has been accumulated to allow discrimination of realistic fuel pin designs

  13. 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

  14. Advanced LMFBR fuel cladding susceptability to stress corrosion due to reprocessing impurities

    International Nuclear Information System (INIS)

    Henslee, S.P.

    1987-03-01

    The potential degradation of LMFBR fuel cladding alloys by chlorides, when used in metallic fuel systems, was evaluated. The alloys tested were D-9 and HT-9 stainless steels, austenitic and ferritic alloys respectively. These two alloys were tested in parallel with and their performance compared to the austenitic stainless steel Type 316. All alloys were tested for 7400 hours in a stress rupture environment with chloride exposure at either 550/degree/C 650/degree/C. None of the alloys tested were found to exhibit any degradation in time-to-rupture by the presence of chlorides under the conditions imposed during testing. 8 refs., 4 figs., 2 tabs

  15. Optimal pin enrichment distributions in nuclear reactor fuel bundles

    International Nuclear Information System (INIS)

    Lim, E.Y.

    1976-01-01

    A methodology has been developed to determine the fuel pin enrichment distribution that yields the best approximation to a prescribed power distribution in nuclear reactor fuel bundles. The problem is formulated as an optimization problem in which the optimal pin enrichments minimize the sum of squared deviations between the actual and prescribed fuel pin powers. A constant average enrichment constraint is imposed to ensure that a suitable value of reactivity is present in the bundle. When constraints are added that limit the fuel pins to a few enrichment types, one must determine not only the optimal values of the enrichment types but also the optimal distribution of the enrichment types amongst the pins. A matrix of boolean variables is used to describe the assignment of enrichment types to the pins. This nonlinear mixed integer programming problem may be rigorously solved with either exhaustive enumeration or branch and bound methods using a modification of the algorithm from the continuous problem as a suboptimization. Unfortunately these methods are extremely cumbersome and computationally overwhelming. Solutions which require only a moderate computational effort are obtained by assuming that the fuel pin enrichments in this problem are ordered as in the solution to the continuous problem. Under this assumption search schemes using either exhaustive enumeration or branch and bound become computationally attractive. An adaptation of the Hooke--Jeeves pattern search technique is shown to be especially efficient

  16. 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.

  17. LMFBR fuel analysis. Task A: oxide fuel dynamics. Final report, July 1, 1975--September 30, 1976

    International Nuclear Information System (INIS)

    Dhir, V.K.; Hauss, B.; Kastenberg, W.E.; Saqui, R.; Sun, Y.H.; Wong, K.

    1976-11-01

    The report summarizes the results of studies conducted in support of the U.S. Nuclear Regulatory Commission's review of the Preliminary Safety Analysis Report for the Clinch River Breeder Reactor. In particular it deals with three aspects of the unprotected transient overpower accident. The first aspect is the response of the Clinch River Breeder Reactor to low reactivity insertion rates. Second, the investigation of a new method for computing the time, place and mode of fuel pin failure is studied. Lastly, the question of post-failure, fuel freezing, and plate-out is addressed. Several areas of uncertainty in the analysis of these accidents is also discussed

  18. 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

  19. 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

  20. Present status of uranium-plutonium mixed carbide fuel development for LMFBR

    International Nuclear Information System (INIS)

    Handa, Muneo; Suzuki, Yasufumi.

    One Oarai characteristic of a carbide fuel is that its doubling time is about 13 years which is only about half as long as that of an oxide fuel. The development of carbide fuels in the past ten years has been truly remarkable. Especially, through the new fuel development program initiated in 1974 in the United States, success has been achieved with respect to He- and Na-bond fuels in obtaining a 16 a/o burning rate without damage to cladding tubes. In 1984 at FFTF, a radiation of a fuel assembly consisting 91 fuel pins is contemplated. On the other hand, in Japan, in 1974, a Fuel Research Wing specializing in the study of carbide fuels was constructed in the Oarai Laboratory of the Atomic Energy Research Institute and in the fall of 1982, was successful in fabricating two carbide fuel pins having different chemical compositions

  1. Role of fuel bubble phenomenology in assessment of LMFBR source term

    International Nuclear Information System (INIS)

    Cho, D.H.; Condiff, D.W.; Chan, S.H.

    1985-01-01

    Phenomenological aspects of a fuel vapor bubble formed in the sodium pool in a hypothetical severe accident are considered. The potential for fuel bubble collapse in the sodium pool is analyzed. It appears that for a wide range of hypothetical LMFBR accidents involving core vaporization, the fuel vapor bubble would likely be quenched and collapse prior to migration to the cover gas region. Such rapid quenching is due mainly to radiative heat transfer from the fuel bubble, coupled with the inherent capability of the sodium pool (large subcooling and high thermal conductivity) to dissipate thermal energy. Major uncertainty in the analysis concerns fuel vapor condensation phenomena at the sodium interface and its effect on the sodium surface radiation absorptivity. This is discussed in detail

  2. 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

  3. 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

  4. Conceptual design study of LMFBR core with carbide fuel

    International Nuclear Information System (INIS)

    Tezuka, H.; Hojuyama, T.; Osada, H.; Ishii, T.; Hattori, S.; Nishimura, T.

    1987-01-01

    Carbide fuel is a hopeful candidate for demonstration FBR(DFBR) fuel from the plant cost reduction point of view. High thermal conductivity and high heavy metal content of carbide fuel lead to high linear heat rate and high breeding ratio. We have analyzed carbide fuel core characteristics and have clarified the concept of carbide fuel core. By survey calculation, we have obtained a correlation map between core parameters and core characteristics. From the map, we have selected a high efficiency core whose features are better than those of an oxide core, and have obtained reactivity coefficients. The core volume and the reactor fuel inventory are approximately 20% smaller, and the burn-up reactivity loss is 50% smaller compared with the oxide fuel core. These results will reduce the capital cost. The core reactivity coefficients are similar to the conventional oxide DFBR's. Therefore the carbide fuel core is regarded as safe as the oxide core. Except neutron fluence, the carbide fuel core has better nuclear features than the oxide core

  5. 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)

  6. 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.

  7. 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)

  8. Experience related to the safety of advanced LMFBR fuel elements

    International Nuclear Information System (INIS)

    Kerrisk, J.F.

    1975-07-01

    Experiments and experience relative to the safety of advanced fuel elements for the liquid metal fast breeder reactor are reviewed. The design and operating parameters and some of the unique features of advanced fuel elements are discussed breifly. Transient and steady state overpower operation and loss of sodium bond tests and experience are discussed in detail. Areas where information is lacking are also mentioned

  9. Cumulative damage fraction design approach for LMFBR metallic fuel elements

    International Nuclear Information System (INIS)

    Johnson, D.L.; Einziger, R.E.; Huchman, G.D.

    1979-01-01

    The cumulative damage fraction (CDF) analytical technique is currently being used to analyze the performance of metallic fuel elements for proliferation-resistant LMFBRs. In this technique, the fraction of the total time to rupture of the cladding is calculated as a function of the thermal, stress, and neutronic history. Cladding breach or rupture is implied by CDF = 1. Cladding wastage, caused by interactions with both the fuel and sodium coolant, is assumed to uniformly thin the cladding wall. The irradiation experience of the EBR-II Mark-II driver fuel with solution-annealed Type 316 stainless steel cladding provides an excellent data base for testing the applicability of the CDF technique to metallic fuel. The advanced metal fuels being considered for use in LMFBRs are U-15-Pu-10Zr, Th-20Pu and Th-2OU (compositions are given in weight percent). The two cladding alloys being considered are Type 316 stainless steel and a titanium-stabilized Type 316 stainless steel. Both are in the cold-worked condition. The CDF technique was applied to these fuels and claddings under the assumed steady-state operating conditions

  10. 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

  11. 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

  12. 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.)

  13. Review and evaluation of cladding attack of LMFBR fuel

    International Nuclear Information System (INIS)

    Koizumi, M.; Nagai, S.; Furuya, H.; Muto, T.

    1977-01-01

    The behavior of cladding inner wall corrosion during irradiation was evaluated in terms of fuel density, fuel form, O/M ratio, plutonium concentration, cladding composition, cladding pretreatment, cladding inner diameter, burnup and cladding inner wall temperature. Factors which influence the corrosion are O/M ratio (oxygen to metal ratio), burn up, cladding inner diameter and cladding inner wall temperature. Maximum cladding inner wall corrosion depth was formulated as a function of O/M ratio, burn up and cladding inner wall temperature

  14. On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H.

    1976-05-01

    The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding

  15. 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

  16. Effects of entrained gas on the acoustic detection of sodium boiling in a simulated LMFBR fuel bundle

    International Nuclear Information System (INIS)

    Leavell, W.H.; Sides, W.H.

    1975-01-01

    The relationship between acoustic intensity of nucleate boiling and void fraction was studied in a simulated LMFBR fuel bundle. Results indicate that as the void fraction increases the detected intensity of nucleate boiling decreased until it was indistinguishable from background noise. (JWR)

  17. 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

  18. A Swiss contribution to a secure LMFBR fuel cycle

    International Nuclear Information System (INIS)

    Nicolet, M.; Bischoff, K.; Hausmann, W.; Stofer, B.

    1978-12-01

    Since 1967, EIR has been using the sphere-pac fuel concept, which takes advantage of the wet route fabrication of (U,Pu) carbide-microspheres using an internal gelation method, followed by carbothermic reduction of the precipitated metal-oxides. Some of the promises of the wet process are a shorter fabrication route than for pellet manufacture, no dust problems, reduced fire hazard for carbides, and last but not least the improvement of Pu safeguards. The method is particularly suitable for direct coupling to a reprocessing plant, where coprocessing of both U and Pu and spiked solutions will be possible. (Auth.)

  19. 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

  20. 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

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

    International Nuclear Information System (INIS)

    Betten, P.R.; Tow, D.M.

    1984-01-01

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

  2. Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

    2008-09-01

    The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

  3. 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

  4. 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

  5. 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)

  6. Post-accident fuel relocation and heat removal in the LMFBR

    International Nuclear Information System (INIS)

    Kazimi, M.S.; Tsai, S.S.; Gasser, R.D.

    1976-08-01

    Assessment of the dynamics of post-accident fuel relocation and heat removal is an important aspect of the evaluation of the consequences of a hypothetical accident in an LMFBR. Such an assessment is of particular importance in the evaluation of the post-accident radiological doses around the reactor site. In the present evaluation particular attention is given to the design features of the Clinch River Breeder Reactor Plant (CRBR). Fuel relocation and heat removal, assuming certain conditions have resulted in core disruption, are discussed. The discussion of events and phenomena involved in the relocation processes is centered around the resulting patterns of heat source distribution. The factors influencing fuel relocation and distribution in the inlet and outlet plena of the reactor vessel are discussed. The current technology of in-vessel heat removal is applied to the design of the CRBR reactor. Both fuel debris cooling limits and overall coolant flow in the reactor under natural convection conditions are explored. Some of the uncertainties in ex-vessel fuel behavior are addressed. In particular, the effect of melting the cavity bed on the rate of growth of a molten fuel pool is investigated

  7. 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

  8. 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

  9. 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)

  10. 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)

  11. Finite element analysis of irradiation-induced dilation of the fuel subassembly duct in LMFBR

    International Nuclear Information System (INIS)

    Gao Fuhai; Fu Hao; Li Nan; Yang Kongli; Wang Mingzhen

    2013-01-01

    Background: The calculation of irradiation-induced dilation of the fuel subassembly duct in LMFBR is important for fast reactor core design.. Purpose: To investigate how to calculate the dilation by using finite element method (FEM). Methods: First, irradiation-induced creep and swelling material models are introduced. Then, a theoretical solution based on a simplified bending plate model is briefly given. Finally, a stress update scheme for the adopted material models is presented and furthermore embedded into ABAQUS user interface UMAT to conduct finite element analysis. Both solutions are compared and discussed. Results: FEM successfully predicts the duct dilation and its solution agrees well with theoretical one in small deformation. Conclusions: The proposed stress update scheme is effective, The accuracy of the theory solution declines when dilation becomes larger. The maximum stress occurs at the duct corner point, and the location has stress relaxation effect. (authors)

  12. 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

  13. 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

  14. 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)

  15. 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)

  16. Design and economic implications of heterogeneity in an LMFBR core

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1983-01-01

    Much emphasis is currently being placed in LMFBR design on reducing both the capital cost and the fuel cycle cost of an LMFBR to insure its economic competativeness without a rapid increase in the uranium prices. In this study the relationship between two core design options, their neutronic consequences, and their effect on fuel cycle cost are analyzed. The two design options are the selection of pin diameter and the degree of heterogeneity. In the case of a heterogeneous core, with a low sodium void reactivity worth this ratio of fertile internal blanket to driver assemblies is generally about 0.40. However, some advantages of cores with heterogeneity of 0.08 to 0.2 for a fixed pin diameter have been reported

  17. 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.)

  18. 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.)

  19. 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)

  20. 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

  1. The radiological significance of transuranium radioisotopes released to the environment during operation of the LMFBR fuel cycle

    International Nuclear Information System (INIS)

    Barr, N.F.

    1976-01-01

    Estimates based on current knowledge and conservative assumptions indicate that release of transuranium elements from the Liquid Metal Fast Breeder Reactor (LMFBR) fuel cycle are likely to proaduce population dose commitments small compared to those produced by naturally occurring alpha emitters and globally dispersed transuranium radioisotopes from tests of nuclear weapons in the atmosphere. Potential health consequences of these releases to current and future generations are estimated to be very small compared to risks associated with the production of energy by fossil fuels. The estimates are subject to a number of uncertainties imposed by lack of knowledge. Some of the uncertainties are not likely to be greatly reduced until LMFBR facilities are designed and operated. Others may be significantly reduced prior to facility design and operation. The paper discusses the sensitivity of the estimates to uncertainties and approches to reducing those uncertainties that strongly influence the estimates. (author)

  2. In-pile observations of fuel and clad relocation during LMFBR initiation phase accident experiments - the STAR experiments

    International Nuclear Information System (INIS)

    Wright, S.A.; Schumacher, G.; Henkel, P.R.; Royl, P.

    1987-01-01

    A series of seven in-pile experiments (the STAR experiments) were performed in which clad motion and fuel dispersal were observed in small pin bundles with high-speed cinematography. The experimental heating conditions reproduced a range of Loss of Flow (LOF) accident scenarios for the lead subassemblies in LMFBRs. The experiments show strong tendencies for limited clad motion in multiple pin bundles, early fuel disruption and dispersal (prior to fuel melting) in moderate power transients having simultaneous clad melting and fuel disruption. The more recent experiments indicate a possibility of steel vapor driven fuel dispersal after fuel breakup and intimate fuel/steel mixing. (author)

  3. 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

  4. Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

    International Nuclear Information System (INIS)

    Edelmann, M.

    1977-01-01

    Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the sub-assemblies with high precision. In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift. The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow. Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power. (author)

  5. Evaluation of organic coolants for the transportation of LMFBR spent fuel rods

    International Nuclear Information System (INIS)

    Arnold, C. Jr.

    1978-05-01

    The physical and chemical processes that are likely to occur when sodium coated LMFBR spent fuel rods are submerged in various aromatic organic coolants was defined by means of immersion experiments carried out with sodium coated 304 stainless steel coupons. Upon immersion of sodium coated coupons at 220 0 C in hydrocarbon type coolants such as Therminol 88, a mixture of terphenyls, not only was the metallic sodium retained on the coupon, but a carbonaceous coating formed on the surface of the sodium. In contrast, coolants that contained aromatic ether bonds, such as Dowtherm A, reacted with sodium at 220 0 C to form phenolate and other salts, which precipitated from the coolant in the form of a dark sludge. With Dowtherm A, removal of metallic sodium from the coupon was essentially complete in a matter of hours at temperatures of 160--220 0 C. Data on the rate and efficiency of sodium removal upon immersion in Dowtherm A at elevated temperatures were obtained. In addition the kinetics and chemistry of the sodium/Dowtherm A reaction were defined. Because sodium sludges are potentially incompatible with the containing structural materials and the fuel elements, it is recommended that sodium be removed prior to immersion in the coolant via reaction with benzoic acid; this method should be adaptable to the facilities at reactor sites. In aging studies Dowtherm A was found to be thermally stable up to 400 0 C and radiatively stable at ambient conditions. The combined effect of heat and radiation was not defined

  6. Fuel-coolant interaction-phenomena under prompt burst conditions. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Jacobs, H.; Young, M.F.; Reil, K.O.

    1979-01-01

    The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2 (190 MPa peak).

  7. Development of the delyed-neutron triangulation technique for locating failed fuel in LMFBR

    International Nuclear Information System (INIS)

    Kryter, R.C.

    1975-01-01

    Two major accomplishments of the ORNL delayed neutron triangulation program are (1) an analysis of anticipated detector counting rates and sensitivities to unclad fuel and erosion types of pin failure, and (2) an experimental assessment of the accuracy with which the position of failed fuel can be determined in the FFTF (this was performed in a quarter-scale water mockup of realistic outlet plenum geometry using electrolyte injections and conductivity cells to simulate delayed-neutron precursor releases and detections, respectively). The major results and conclusions from these studies are presented, along with plans for further DNT development work at ORNL for the FFTF and CRBR. (author)

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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

  13. Experimental program on fuel rod behaviour under off-normal conditions

    International Nuclear Information System (INIS)

    Languille, A.; Cecchi, P.

    1985-01-01

    During LMFBR plant operation, fuel developments are primarily concerned with the fuel pin irradiation behaviour under steady-state conditions up to high burn-up levels. But additional studies under off-normal conditions are necessary in order to assess fuel pin performance and to define operational limits. (author)

  14. Absorption process for removing krypton from the off-gas of an LMFBR fuel reprocessing plant

    International Nuclear Information System (INIS)

    Stephenson, M.J.; Dunthorn, D.I.; Reed, W.D.; Pashley, J.H.

    1975-01-01

    The Oak Ridge Gaseous Diffusion Plant selective absorption process for the collection and recovery of krypton and xenon is being further developed to demonstrate, on a pilot scale, a fluorocarbon-based process for removing krypton from the off-gas of an LMFBR fuel reprocessing plant. The new ORGDP selective absorption pilot plant consists of a primary absorption-stripping operation and all peripheral equipment required for feed gas preparation, process solvent recovery, process solvent purification, and krypton product purification. The new plant is designed to achieve krypton decontamination factors in excess of 10 3 with product concentration factors greater than 10 4 while processing a feed gas containing typical quantities of common reprocessing plant off-gas impurities, including oxygen, carbon dioxide, nitrogen oxides, water, xenon, iodine, and methyl iodide. Installation and shakedown of the facility were completed and some short-term tests were conducted early this year. The first operating campaign using a simulated reprocessing plant off-gas feed is now underway. The current program objective is to demonstrate continuous process operability and performance for extended periods of time while processing the simulated ''dirty'' feed. This year's activity will be devoted to routine off-gas processing with little or no deliberate system perturbations. Future work will involve the study of the system behavior under feed perturbations and various plant disturbances. (U.S.)

  15. LMFBR source term experiments in the Fuel Aerosol Simulant Test (FAST) facility

    International Nuclear Information System (INIS)

    Petrykowski, J.C.; Longest, A.W.

    1985-01-01

    The transport of uranium dioxide (UO 2 ) aerosol through liquid sodium was studied in a series of ten experiments in the Fuel Aerosol Simulant Test (FAST) facility at Oak Ridge National Laboratory (ORNL). The experiments were designed to provide a mechanistic basis for evaluating the radiological source term associated with a postulated, energetic core disruptive accident (CDA) in a liquid metal fast breeder reactor (LMFBR). Aerosol was generated by capacitor discharge vaporization of UO 2 pellets which were submerged in a sodium pool under an argon cover gas. Measurements of the pool and cover gas pressures were used to study the transport of aerosol contained by vapor bubbles within the pool. Samples of cover gas were filtered to determine the quantity of aerosol released from the pool. The depth at which the aerosol was generated was found to be the most critical parameter affecting release. The largest release was observed in the baseline experiment where the sample was vaporized above the sodium pool. In the nine ''undersodium'' experiments aerosol was generated beneath the surface of the pool at depths varying from 30 to 1060 mm. The mass of aerosol released from the pool was found to be a very small fraction of the original specimen. It appears that the bulk of aerosol was contained by bubbles which collapsed within the pool. 18 refs., 11 figs., 4 tabs

  16. Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

    International Nuclear Information System (INIS)

    Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S.

    2003-01-01

    During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- ε turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential

  17. 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

  18. Improved Retrieval Technique of pin-wise composition for spent fuel recycling

    Energy Technology Data Exchange (ETDEWEB)

    Park, YunSeo; Kim, Myung Hyun [Kyung Hee University , Yongin (Korea, Republic of)

    2016-10-15

    New reutilization method which does not require fabrication processing was suggested and showed feasibility by Dr. Aung Tharn Daing. This new reutilization method is predict spent nuclear fuel pin composition, reconstruct new fuel assembly by spent nuclear pin, and directly reutilize in same PWR core. There are some limitation to predict spent nuclear fuel pin composition on his methodology such as spatial effect was not considered enough. This research suggests improving Dr. Aung Tharn Daing's retrieval technique of pin-wise composition. This new method classify fuel pin groups by its location effect in fuel assembly. Most of fuel pin composition along to burnup in fuel assembly is not highly dependent on location. However, compositions of few fuel pins where near water hole and corner of fuel assembly are quite different in same burnup. Required number of nuclide table is slightly increased from 3 to 6 for one fuel assembly with this new method. Despite of this little change, prediction of the pin-wise composition became more accurate. This new method guarantees two advantages than previous retrieving technique. First, accurate pin-wise isotope prediction is possible by considering location effect in a fuel assembly. Second, it requires much less nuclide tables than using full single assembly database. Retrieving technique of pin-wise composition can be applied on spent fuel management field useful. This technique can be used on direct use of spent fuel such as Dr. Aung Tharn Daing showed or applied on pin-wise waste management instead of conventional assembly-wise waste management.

  19. 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

  20. 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)

  1. 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

  2. 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)

  3. 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

  4. 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.

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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)

  12. 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

  13. 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.)

  14. Manipulator for fuel assemblies in a spent fuel pool, especially for a LMFBR

    International Nuclear Information System (INIS)

    Dalmas, R.

    1988-01-01

    The spent fuel manipulator has - a travelling crane moving longitudinally: - a carriage moving on the travelling crane in a direction perpendicular to its motion so that the carriage is positioned over each assembly, - a telescopic rod carried by the carriage and terminating in a vertically mobile grapple, - a tubular shielded hood on the carriage extending downwards to house the rod, grapple and fuel assembly and maintaining a biologically acceptable level of radiation above the surface of the pool [fr

  15. Uncertainty analysis of light water reactor unit fuel pin cells

    Energy Technology Data Exchange (ETDEWEB)

    Kamerow, S.; Ivanov, K., E-mail: sln107@PSU.EDU, E-mail: kni1@PSU.EDU [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, PA (United States); Moreno, C. Arenas, E-mail: cristina.arenas@UPC.EDU [Department of Physics and Nuclear Engineering, Technical University of Catalonia, Barcelona (Spain)

    2011-07-01

    The study explored the calculation of uncertainty based on available covariance data and computational tools. Uncertainty due to temperature changes and different fuel compositions are the main focus of this analysis. Selected unit fuel pin cells were analyzed according to the OECD LWR UAM benchmark specifications. Criticality and uncertainty analyses were performed using TSUNAMI-1D sequence in SCALE 6.0. It was found that uncertainties increase with increasing temperature while k{sub eff} decreases. This increase in the uncertainty is due to the increase in sensitivity of the largest contributor of uncertainty, namely nuclide reaction {sup 238}U (n, gamma). The sensitivity grew larger as the capture cross-section of {sup 238}U expanded due to Doppler broadening. In addition, three different compositions (UOx, MOx, and UOxGd{sub 2}O{sub 3}) of fuel cells were analyzed. It showed a remarkable increase in uncertainty in k{sub eff} for the case of the MOx fuel cell and UOxGd{sub 2}O{sub 3} fuel cell. The increase in the uncertainty of k{sub eff} in UOxGd{sub 2}O{sub 3} fuel was nearly twice of that in MOx fuel and almost four times the amount in UOx fuel. The components of the uncertainties in k{sub eff} in each case were examined and it was found that the neutron-nuclide reaction of {sup 238}U, mainly (n,n'), contributed the most to the uncertainties in the cases of MOx and UOxGd{sub 2}O{sub 3}. At higher energy, the covariance coefficient matrix of {sup 238}U (n,n') to {sup 238}U (n,n') and {sup 238}U (n,n') cross-section showed very large values. Further, examination of the UOxGd{sub 2}O{sub 3} case found that the {sup 238}U (n,n') became the dominant contributor to the uncertainty because most of the thermal neutrons in the cell were absorbed by Gadolinium in UOxGd{sub 2}O{sub 3} case and thus shifting the neutron spectrum to higher energy. For the MOx case on other hand, {sup 239}Pu has a very strong absorption cross-section at low energy

  16. 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)

  17. 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

  18. 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

  19. 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

  20. Development of LIFE4-CN: a combined code for steady-state and transient analyses of advanced LMFBR fuels

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Zawadzki, S.; Billone, M.C.; Nayak, U.P.; Roth, T.

    1979-01-01

    The methodology used to develop the LMFBR carbide/nitride fuels code, LIFE4-CN, is described in detail along with some subtleties encountered in code development. Fuel primary and steady-state thermal creep have been used as an example to illustrate the need for physical modeling and the need to recognize the importance of the materials characteristics. A self-consistent strategy for LIFE4-CN verification against irradiation data has been outlined with emphasis on the establishment of the gross uncertainty bands. These gross uncertainty bands can be used as an objective measure to gauge the overall success of the code predictions. Preliminary code predictions for sample steady-state and transient cases are given

  1. 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

  2. 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)

  3. 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)

  4. 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

  5. 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)

  6. 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)

  7. 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

  8. 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.

  9. 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.

  10. 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)

  11. COXPRO-II: a computer program for calculating radiation and conduction heat transfer in irradiated fuel assemblies

    International Nuclear Information System (INIS)

    Rhodes, C.A.

    1984-12-01

    This report describes the computer program COXPRO-II, which was written for performing thermal analyses of irradiated fuel assemblies in a gaseous environment with no forced cooling. The heat transfer modes within the fuel pin bundle are radiation exchange among fuel pin surfaces and conduction by the stagnant gas. The array of parallel cylindrical fuel pins may be enclosed by a metal wrapper or shroud. Heat is dissipated from the outer surface of the fuel pin assembly by radiation and convection. Both equilateral triangle and square fuel pin arrays can be analyzed. Steady-state and unsteady-state conditions are included. Temperatures predicted by the COXPRO-II code have been validated by comparing them with experimental measurements. Temperature predictions compare favorably to temperature measurements in pressurized water reactor (PWR) and liquid-metal fast breeder reactor (LMFBR) simulated, electrically heated fuel assemblies. Also, temperature comparisons are made on an actual irradiated Fast-Flux Test Facility (FFTF) LMFBR fuel assembly

  12. 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

  13. 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

  14. 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

  15. Measuring the linear heat generation rate of a nuclear reactor fuel pin

    International Nuclear Information System (INIS)

    Smith, R.D.

    1981-01-01

    A miniature gamma thermometer is described which is capable of travelling through bores distributed in an array through a nuclear reactor core and measure the linear heat generation rate of the fuel pins. (U.K.)

  16. 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

  17. 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)

  18. Nuclear reactor fuel element with a cluster of parallel fuel pins

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  19. 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

  20. 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)

  1. 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

  2. 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)

  3. Investigation into fuel pin reshuffling options in PWR in-core fuel management for enhancement of efficient use of nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Daing, Aung Tharn, E-mail: atdaing@khu.ac.kr; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

    2014-07-01

    Highlights: • This paper discusses an alternative option, fuel pin reshuffling for maximization of cycle energy production. • The prediction results of isotopic compositions of each burnt pin are verified. • The operating performance is analyzed at equilibrium core with fuel pin reshuffling. • The possibility of reuse of spent fuel pins for reduction of fresh fuel assemblies is investigated. - Abstract: An alternative way to enhance efficient use of nuclear fuel is investigated through fuel pin reshuffling options within PWR fuel assembly (FA). In modeling FA with reshuffled pins, as prerequisite, the single pin calculation method is proposed to estimate the isotopic compositions of each pin of burnt FA in the core-wide environment. Subsequently, such estimation has been verified by comparing with the neutronic performance of the reference design. Two scenarios are concerned, i.e., first scenario was targeted on the improvement of the uniform flux spatial distribution and on the enhancement of neutron economy by simply reshuffling the existing fuel pins in once-burnt fuel assemblies, and second one was focused on reduction of fresh fuel loading and discharged fuel assemblies with more economic incentives by reusing some available spent fuel pins still carrying enough reactivity that are mechanically sound ascertained. In scenario-1, the operating time was merely somewhat increased for few minutes when treating eight FAs by keeping enough safety margins. The scenario-2 was proved to reduce four fresh FAs loading without largely losing any targeted parameters from the safety aspect despite loss of 14 effective full power days for operation at reference plant full rated power.

  4. 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

  5. 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

  6. 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)

  7. Performance analysis of LMFBR control rods

    International Nuclear Information System (INIS)

    Pitner, A.L.; Birney, K.R.

    1975-01-01

    Control rods in the FFTF and LMFBR's will consist of pin bundles of stainless steel-clad boron carbide pellets. In the FFTF reference design, sixty-one pins of 0.474-inch diameter each containing a 36-inch stack of 0.362-inch diameter boron carbide pellets comprise a control rod. Reactivity control is provided by the 10 B (n,α) 7 Li reaction in the boron carbide. This reaction is accompanied by an energy release of 2.8 MeV, and heating from this reaction typically approaches 100 watts/cm 3 for natural boron carbide pellets in an LMFBR flux. Performance analysis of LMFBR control rods must include an assessment of the thermal performance of control pins. In addition, irradiation performance with regard to helium release, pellet swelling, and reactivity worth depletion as a function of service time must be evaluated

  8. 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 %

  9. 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

  10. 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

  11. 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...

  12. 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

  13. 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)

  14. Accident considerations in LMFBR design

    International Nuclear Information System (INIS)

    Simpson, D.E.; Alter, H.; Fauske, H.K.; Hikido, K.; Keaten, R.W.; Stevenson, M.G.; Strawbridge, L.

    1975-12-01

    LMFBR safety design criteria are discussed from the standpoints of accident severity classification and damage criteria, and the following design events are considered: fuel failure propagation, reactivity addition faults, heat transport system events, steam generator faults, sodium spills, fuel handling and storage faults, and external events

  15. 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

  16. 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%)

  17. 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.

  18. Accuracy of dimension measurements from neutron radiographs of nuclear fuel pins

    International Nuclear Information System (INIS)

    Domanus, J.C.

    1976-01-01

    A comparison is given of accuracies obtained with measuring the dimensions (pellet diameter and fuel-clad gap) from neutron and X-ray radiographs of a calibrated nuclear fuel pin performed with a projection microscope, microdensitometers and a video micrometer

  19. Hydrodynamic analysis of the LMFBR prompt burst excursion (PBE) experiment

    International Nuclear Information System (INIS)

    Young, M.F.

    1977-01-01

    A series of in-pile experiments has been conducted at Sandia Laboratories to provide information on pressure levels and conversion of thermal energy into mechanical work in LMFBR cores during hypothetical, superprompt-critical excursions. Pressures generated in these experiments are recorded by a pressure transducer located at the top and bottom of a sodium channel surrounding a single, fresh UO 2 fuel pin. Work energy conversion is measured by a linear motion transducer connected to a piston at the top of the sodium column. Since the pressure transducers are located fairly far from the location of pin failure, it becomes necessary to determine the effect of channel geometry and piston motion on the observed pressure data. A two-dimensional, hydrodynamic analysis of pressure pulse propagation in the fuel pin-coolant channel geometry was therefore performed using the CSQII computer code. The initial series of PBE experiments consists of single, fresh UO 2 pins surrounded by a sodium-filled or dry-coolant channel contained in a closed test capsule. The capsule is subjected to a maximum pulse in the Annular Core Pulse Reactor (ACPR) resulting in an energy deposition of from 2350 to 2900 J/g (14 and 20 percent enriched pins). The pulse width at half maximum (PWHM) is about 5 ms

  20. 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

  1. 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

  2. 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

  3. The behaviour of Phenix fuel pin bundle under irradiation

    International Nuclear Information System (INIS)

    Marbach, G.; Millet, P.; Blanchard, P.; Huillery, R.

    1979-07-01

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

  4. Comparative analysis of a hypothetical loss-of-flow accident in an irradiated LMFBR core using different computer models for a common benchmark problem

    International Nuclear Information System (INIS)

    Wider, H.U.; Devos, J.; Nguyen, H.; Goethem, G. Van.; Miles, K.J.; Tentner, A.M.; Pizzica, P.

    1989-01-01

    This report summarizes the results of an international exercise to compare whole-core accident calculations of the initiation phase of an unprotected LOF accident in a large irradiated LMFBR. The results for the accident phase before pin failure are in rather good agreement except for the fuel pin mechanics predictions. There are also some differences in the sodium boiling calculations but the voiding rates which are of key importance are very similar. The post - failure fuel motion and sodium voiding predictions show significant differences. However, the majority of these calculations agree that temporary fuel accumulations occur which increase the power beyond that caused by sodium voiding alone

  5. 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

  6. 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

  7. SASSYS LMFBR systems code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

    1983-01-01

    The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

  8. 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

  9. 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.

  10. Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

    International Nuclear Information System (INIS)

    Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

    1976-01-01

    Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

  11. On the hazard accumulation of actinide waste in a Pu-fueled LMFBR power economy with and without by-product actinide recycling

    International Nuclear Information System (INIS)

    Anselmi, L.; Caruso, K.; Hage, W.; Schmidt, E.

    1979-01-01

    The actinide waste arisings in terms of hazard potential for ingestion and inhalation are given for a Pu-fueled LMFBR Power Economy as function of decay time. The data were assessed for two simplified fuel cycles, one considering the recycling of by-product actinides and the other their complete discharge to the high-level waste. Two durations of nuclear power and several loss fractions of actinides to the waste were considered. The major contributors in form of chemical elements or isotopes to the actinide waste hazard built up during the nuclear power duration were identified for various decay intervals

  12. 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.

  13. 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

  14. Visual observations of fuel disruption in in-pile LMFBR accident experiments

    International Nuclear Information System (INIS)

    Wright, S.A.; Mast, P.K.

    1982-01-01

    Sandia National Laboratories has been investigating initiation phase phenomena in a series of Fuel Disruption (FD) experiments since 1977. In this program high speed cinematography is used to observe fuel disruption in in-pile experiments that simulate loss of flow accidents. Thus, these experiments provide high resolution measurements of initial fuel and clad motion with prototypic materials and prototypic heating conditions. The main objective of the FD experiment is to determine the timing (relative to fuel temperature) and the mode of fuel disruption under LOF heating conditions. Observed modes of disruption include fuel swelling, solid state breakup, cracking, ejection of a molten fuel jet, slumping, and rapid expansion of small particles. Because the temperature and character of the fuel at disruption are known, disruption can be correlated with the mechanisms driving the disruption such as fuel vapor pressure, molten fuel expansion, fission gases, and impurity gases

  15. In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

    1983-01-01

    Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

  16. 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.

  17. Theoretical interpretation of SCARABEE single pin in-pile boiling experiments

    International Nuclear Information System (INIS)

    Struwe, D.; Bottoni, M.; Fries, W.; Elbel, H.; Angerer, G.

    1977-01-01

    In the framework of LMFBR safety analysis a theoretical interpretation of some of the most representative of the single pin experiments of the in-pile SCARABEE project has been performed from both viewpoints of thermohydraulic and fuel behaviour using the computer codes CAPRI-2 and SATURN-1. The analysis is aimed at investigating the pin behavior from the preirradiation history, through the observed sequence of events following a coolant mass flow reduction from boiling inception up to pin breakdown. A comparison of theoretical results with experimentally recorded data has allowed a deeper insight into the peculiar features of the experiments and enabled a valuable code verification. (Auth.)

  18. 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

  19. 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)

  20. Wire-wrapped rod-bundle heat-transfer analysis for LMFBR

    International Nuclear Information System (INIS)

    Wong, C.N.C.; Todreas, N.E.

    1982-07-01

    Helical wire wraps are widely used in the LMFBR fuel and blanket assemblies to provide coolant mixing and maintain proper spacing between fuel pins. The presence of the helical wire, however, may possibly induce heat transfer problems, such as the uncertainty of the maximum clad temperature as a result of the contact between the wires and the pins. In this study, the detailed transient three dimensional velocity and temperature distributions for the coolant around the pin will be determined by solving the governing momentum and energy equation numerically. A computer code HEATRAN has been developed to perform this calculation. Before the computer code HEATRAN is applied to the wire wrapped rod bundle problem, it is used to analyze a wide range of fluid and heat transfer problem to verify its capabilities

  1. 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

  2. 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.)

  3. 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)

  4. 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.)

  5. 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

  6. 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

  7. Environmental control aspects for fabrication, reprocessing and waste disposal of alternative LWR and LMFBR fuels

    International Nuclear Information System (INIS)

    Nolan, A.M.; Lewallen, M.A.; McNair, G.W.

    1979-11-01

    Environmental control aspects of alternative fuel cycles have been analyzed by evaluating fabrication, reprocessing, and waste disposal operations. Various indices have been used to assess potential environmental control requirements. For the fabrication and reprocessing operations, 50-year dose commitments were used. Waste disposal was evaluated by comparing projected nuclide concentrations in ground water at various time periods with maximum permissible concentrations (MPCs). Three different fabrication plants were analyzed: a fuel fabrication plant (FFP) to produce low-activity uranium and uranium-thorium fuel rods; a plutonium fuel refabrication plant (PFRFP) to produce plutonium-uranium and plutonium-thorium fuel rods; and a uranium fuel refabrication plant (UFRFP) to produce fuel rods containing the high-activity isotopes 232 U and 233 U. Each plant's dose commitments are discussed separately. Source terms for the analysis of effluents from the fuel reprocessing plant (FRP) were calculated using the fuel burnup codes LEOPARD, CINDER and ORIGEN. Effluent quantities are estimated for each fuel type. Bedded salt was chosen for the waste repository analysis. The repository site is modeled on the Waste Isolation Pilot Program site in New Mexico. Wastes assumed to be stored in the repository include high-level vitrified waste from the FRP, packaged fuel residue from the FRP, and transuranic (TRU) contaminated wastes from the FFP, PFRFP, and UFRFP. The potential environmental significance was determined by estimating the ground-water concentrations of the various nuclides over a time span of a million years. The MPC for each nuclide was used along with the estimated ground-water concentration to generate a biohazard index for the comparison among fuel compositions

  8. 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

  9. 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)

  10. 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

  11. 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

  12. 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.

  13. 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

  14. 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

  15. 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)

  16. Fuel and coolant motions following pin failure: EPIC models and the PBE-5S experiment

    International Nuclear Information System (INIS)

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

    1979-01-01

    The EPIC computer code has been used to analyze the post-fuel-pin-failure behavior in the PBE-5S experiment performed at Sandia Laboratories. The effects of modeling uncertainties on the calculation are examined. The calculations indicate that the majority of the piston motion observed in the test is due to the initial pressurization of the coolant channel by fuel vapor at cladding failure. A more definitive analysis requires improvements in calculational capabilities and experiment diagnostics

  17. 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

  18. 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.)

  19. Studies on modeling to failed fuel detection system response in LMFBR

    International Nuclear Information System (INIS)

    Miyazawa, T.; Saji, G.; Mitsuzuku, N.; Hikichi, T.; Odo, T.; Rindo, H.

    1981-05-01

    Failed Fuel Detection (FFD) system with Fission Products (FP) detection is considered to be the most promissing method, since FP provides direct information against fuel element failure. For designing FFD system and for evaluating FFD signals, some adequate FFD signal response to fuel failure have been required. But few models are available in nowadays. Thus Power Reactor and Nuclear Fuel Development Corporation (PNC) had developed FFD response model with computer codes, based on several fundamental investigations on FP release and FP behavior, and referred to foreign country experiences on fuel failure. In developing the model, noble gas and halogen FP release and behavior were considered, since FFD system would be composed of both cover gas monitoring and delayed neutron monitoring. The developed model can provide typical fuel failure response and detection limit which depends on various background signals at cover gas monitoring and delayed neutron monitoring. According to the FFD response model, we tried to assume fuel failure response and detection limit at Japan experimental fast reactor ''JOYO''. The detection limit of JOYO FFD system was estimated by measuring the background signals. Followed on the studies, a complete computer code has been now made with some improvement. On the paper, the details of the model, out line of developed computer code, status of JOYO FFD system, and trial assumption of JOYO FFD response and detection limit. (author)

  20. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  1. Emission computer tomography on a Dodewaard mixed oxide fuel pin

    International Nuclear Information System (INIS)

    Buurveld, H.A.; Dassel, G.

    1993-12-01

    A nondestructive technique as well as a destructive PIE technique have been used to verify the results obtained with a newly 8-e computer tomography (GECT) system. Multi isotope Scanning (MIS), electron probe micro analysis (EPMA) and GECT were used on a mixed oxide (MOX) fuel rod from the Dodewaard reactor with an average burnup of 24 MWd/kg fuel. GECT shows migration of Cs to the periphery of fuel pellets and to radial cracks and pores in the fuel, whereas MIS shows Cs migration to pellet interfaces. The EPMA technique appeared not to be useful to show migration of Cs but, it shows the distribution of fission products from Pu. EPMA clearly shows the distribution of fission products from Pu, but did not reveal the Cs-migration. (orig./HP)

  2. Research on nuclear energy in the fields of fuel cycle, PWR reactors and LMFBR reactors

    International Nuclear Information System (INIS)

    Barre, B.; Camarcat, N.

    1995-01-01

    In this article we present the CEA research programs to improve the safety of the next generation of reactors, to manage the Plutonium and the wastes of the fuel cycle end and to ameliorate the competitiveness. 6 refs

  3. 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

  4. 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)

  5. Compatibility of niobium, titanium, and vanadium metals with LMFBR cladding

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1975-10-01

    A series of laboratory capsule annealing experiments were conducted to assess the compatibility of niobium, vanadium, and titanium with 316 stainless steel cladding in the temperature range of 700 to 800 0 C. Niobium, vanadium, and titanium are cantidate oxygen absorber materials for control of oxygen chemistry in LMFBR fuel pins. Capsule examination indicated good compatibility between niobium and 316 stainless steel at 800 0 C. Potential compatibility problems between cladding and vanadium or titanium were indicated at 800 0 C under reducing conditions. In the presence of Pu/sub 0.25/U/sub 0.75/O/sub 1.98/ fuel (Δanti G 02 congruent to -160 kcal/mole) no reaction was observed between vanadium or titanium and cladding at 800 0 C

  6. 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)

  7. 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

  8. 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)

  9. Thermal analysis of the IDENT 1578 fuel pin shipping container

    International Nuclear Information System (INIS)

    Ingham, J.G.

    1980-01-01

    The IDENT 1578 container, which is a 110-in. long 5.5-in. OD tube, is designed for shipping FFTF fuel elements in T-3 casks between HEDL, HFEF, and other laboratories. The thermal analysis was conducted to evaluate whether or not the container satisfies its thermal design criteria

  10. 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

  11. 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.

  12. 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

  13. 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

  14. Development of 3-Pin Fuel Test Loop and Utilization Technology

    International Nuclear Information System (INIS)

    Lee, Chung Young; Sim, B. S.; Lee, C. Y.

    2007-06-01

    The principal contents of this project are to design, fabricate and install the steady-state fuel test loop in HANARO for nuclear technology development. Procurement and, fabrication of main equipment, licensing and installation for fuel test loop have been performed. Following contents are described in the report. 1. Design - Design of the In-pile system and Out pile system 2. Fabrication and procurement of the equipment - Fabrication of the In-pile system and In-pool piping - Fabrication and procurement of the equipment of the out-pile system 3. Acquisition of the license - Preparation of the safety analysis report and acquisition of the license - Pre-service inspection of the facility 4. Installation and commissioning - Installation of the FTL - Development of the commissioning procedure

  15. Criticality experiments with fast flux test facility fuel pins

    International Nuclear Information System (INIS)

    Bierman, S.R.

    1990-11-01

    A United States Department of Energy program was initiated during the early seventies at the Hanford Critical Mass Laboratory to obtain experimental criticality data in support of the Liquid Metal Fast Breeder Reactor Program. The criticality experiments program was to provide basic physics data for clean well defined conditions expected to be encountered in the handling of plutonium-uranium fuel mixtures outside reactors. One task of this criticality experiments program was concerned with obtaining data on PuO 2 -UO 2 fuel rods containing 20--30 wt % plutonium. To obtain this data a series of experiments were performed over a period of about twelve years. The experimental data obtained during this time are summarized and the associated experimental assemblies are described. 8 refs., 7 figs

  16. Fabrication details for wire wrapped fuel assembly components

    International Nuclear Information System (INIS)

    Bosy, B.J.

    1978-09-01

    Extensive hydraulic testing of simulated LMFBR blanket and fuel assemblies is being carried out under this MIT program. The fabrication of these test assemblies has involved development of manufacturing procedures involving the wire wrapped pins and the flow housing. The procedures are described in detail in the report

  17. Safety problems related to microheterogeneities in physically mixed oxide fuels

    International Nuclear Information System (INIS)

    Renard, A.; Evrard, G.; Vanhellemont, G.

    1976-01-01

    The safety aspects of microheterogeneities in LMFBR mixed oxide fuel are reviewed from the point of view of the pin behaviour dynamic study, the fabrication and the quality control. The paper emphasizes some significant parameters in transient conditions, the prevention means in the fabrication process and the analysis methods for the control

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

    International Nuclear Information System (INIS)

    Govindha Rasu, N.; Velusamy, K.; Sundararajan, T.; Chellapandi, P.

    2013-01-01

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

  19. Two dimensional, two fluid model for sodium boiling in LMFBR fuel assemblies

    International Nuclear Information System (INIS)

    Granziera, M.R.; Kazimi, M.S.

    1980-05-01

    A two dimensional numerical model for the simulation of sodium boiling transient was developed using the two fluid set of conservation equations. A semiimplicit numerical differencing scheme capable of handling the problems associated with the ill-posedness implied by the complex characteristic roots of the two fluid problems was used, which took advantage of the dumping effect of the exchange terms. Of particular interest in the development of the model was the identification of the numerical problems caused by the strong disparity between the axial and radial dimensions of fuel assemblies. A solution to this problem was found which uses the particular geometry of fuel assemblies to accelerate the convergence of the iterative technique used in the model. Three sodium boiling experiments were simulated with the model, with good agreement between the experimental results and the model predictions

  20. 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

  1. Plant-scale anodic dissolution of unirradiated IFR fuel pins

    International Nuclear Information System (INIS)

    Gay, E.C.; Tomczuk, Z.; Miller, W.E.

    1993-01-01

    This report discusses anodic dissolution which is a major operation in the pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor (IFR), an advanced reactor design developed at Argonne National Laboratory. This process involves electrorefining the heavy metals (uranium and plutonium) from chopped, steel-clad fuel segments. The heavy metals are electrotransported from anodic dissolution baskets to solid and liquid cathodes in a molten salt electrolyte (LiCl-KCI) at 500 degrees C. Uranium is recovered on a solid cathode mandrel, while a uranium-plutonium mixture is recovered in a liquid cadmium cathode. The anode configuration consists of four baskets mounted on an anode shaft. These baskets provide parallel circuits in the electrolyte and salt flow through the chopped fuelbed as the baskets are rotated. The baskets for the engineering-scale tests were sized to contain up to 2.5 kg of heavy metal. Anodic dissolution of 10 kg batches of chopped, steel-clad simulated tuel (U-10% Zr and U-Zr-Fs alloy) was demonstrated

  2. Basic analysis and a comparison of the characteristics GCFRs and the LMFBR with the thorium cycle in one-group diffusion theory

    International Nuclear Information System (INIS)

    Sabundjian, G.; Ishiguro, Y.

    1991-09-01

    A preliminary study of neutronics of thorium cycle fast breeder reactor has been done using simplified reactor models and analyses methods with the aim of finding a type of breeder reactor suitable for an efficient utilization of thorium that is abundant in Brazil. Basic methods of cross section processing and reactor calculation are studied and applied to analyse breeding characteristics of GCFRs and LMFBRs. The GCFR is fueled with oxide pins and cooled with helium. The LMFBR is fueled with thin metallic pins to achieve high power densities. Neutronics characteristics are determined as functions of the average power density and the fuel volume fraction. Results show that a high power density and a high fuel volume fraction are desirable to achieve short doubling times, that the GCFR is inferior to the LMFBR in regard to the doubling time and that the LMFBR can achieve reactor doubling times ten years with an average power density of ∼ 600MW/m 3 and fuel volume fraction of 40%. (author)

  3. 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

  4. Neutron coincidence counter for MOX fuel pins in storage trays: users' manual

    International Nuclear Information System (INIS)

    Cowder, L.; Menlove, H.

    1982-08-01

    The neutron coincidence counter for measurement of mixed-oxide fuel pins in storage trays is described. The special detector head has been designed so that the detectors, high-voltage junction boxes, and electronics are interchangeable with those of the high-level neutron coincidence counter system. This manual describes the system components and the operation and maintenance of the counter. The counter was developed at Los Alamos National Laboratory for in-plant inspection applications by the International Atomic Energy Agency

  5. 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.

  6. 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

  7. 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

  8. Modelling of the thermomechanical and physical processes in FR fuel pins using the GERMINAL code

    International Nuclear Information System (INIS)

    Roche, L.; Pelletier, M.

    2000-01-01

    In the frame of the R and D on Fast Reactor mixed oxide fuels, CEA/DEC has developed the computer code GERMINAL for studying fuel pin thermal and mechanical behaviour, both during steady-state and incidental conditions, up to high burn-up (25 at%). The first part of this paper is devoted to the description of the main models: fuel evolution (central hole and porosity evolution, Plutonium redistribution, O/M radial profile, transient gas swelling, melting fuel behaviour, minor actinides production), high burn-up models (fission gas, volatile fission products and JOG formation), fuel-cladding heat transfer, fuel-cladding mechanical interaction. The second part gives some examples of calculation results taken from the GERMINAL validation data base (more than 40 experiments from PHENIX, PFR, CABRI reactors), with special emphasis on: local fission gas retention and global release, fuel geometry evolution, radial redistribution of plutonium for high burn-up fuels, solid and annular fuel behaviour during power ramps including fuel melting, helium formation from MA (Am and Np) doped homogeneous fuels. (author)

  9. 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.

  10. 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.)

  11. 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.

  12. 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

  13. The development of automated fuel dismantling equipment for a future head-end plant

    International Nuclear Information System (INIS)

    Haberlin, M.M.

    1987-01-01

    For future reprocessing plants, practicable methods for dismantling fuel elements are being examined at Springfields Nuclear Power Development Laboratories which would meet the requirements of a high throughput facility. This paper contains the initial results of an experimental programme undertaken to develop and evaluate an automated high speed single/multiple pin extraction unit. Concomitant parts of the programme include the design and evaluation of single and multi-pin extraction chucks. Dummy fuel elements, a 325 pin gridded LMFBR assembly and a 17 x 17 pin gridded LWR assembly were used to assess process efficacy

  14. TRANSENERGY S: computer codes for coolant temperature prediction in LMFBR cores during transient events

    International Nuclear Information System (INIS)

    Glazer, S.; Todreas, N.; Rohsenow, W.; Sonin, A.

    1981-02-01

    This document is intended as a user/programmer manual for the TRANSENERGY-S computer code. The code represents an extension of the steady state ENERGY model, originally developed by E. Khan, to predict coolant and fuel pin temperatures in a single LMFBR core assembly during transient events. Effects which may be modelled in the analysis include temporal variation in gamma heating in the coolant and duct wall, rod power production, coolant inlet temperature, coolant flow rate, and thermal boundary conditions around the single assembly. Numerical formulations of energy equations in the fuel and coolant are presented, and the solution schemes and stability criteria are discussed. A detailed description of the input deck preparation is presented, as well as code logic flowcharts, and a complete program listing. TRANSENERGY-S code predictions are compared with those of two different versions of COBRA, and partial results of a 61 pin bundle test case are presented

  15. 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)

  16. 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

  17. The investigation of fast reactor fuel pin start up behaviour in the irradiation experiment DUELL II

    International Nuclear Information System (INIS)

    Freund, D.; Geithoff, D.

    1988-04-01

    The irradiation experiments DUELL-II within the SNR-300 operational Transient Experimental Program deal with the investigation of fresh mixed oxide fuel behaviour at start-up. The irradiation has been carried out in the HFR Petten in four so-called DUELL capsules with two fuel pin samples each. The fuel pins with a total length of 453 mm contained a fuel column of 150 mm length, consisting of high dense (U,Pu)O 2-x fuel with an initial porosity of 4%, a Pu-content of 20.9%, and an O/Me ratio of 1.96. The fuel pellet diameter was 6.37 mm, the outer diameter of the SS cladding, material No. 1.4970, was 7.6 mm. The irradiation included four phases, consisting of preconditioning at 85% nominal power (corresponds to 550 W/cm), a following increase to full power, and two following full power periods of 1 and 10 days, respectively. Post irradiation examination showed incomplete fuel restructuring in the first capsules with central void diameters of 800 μm in the hot plane, complete restructuring in the last capsule, leading to central voids of approximately 1 mm diameter. The residual gaps between fuel and clad varied between 25 and 44 μm. The clad inner surface did not show any corrosion attack. The analysis of fuel restructuring has been carried out with the computer code SATURN-S showing good agreement with the PIE results. The analysis led to a series of model improvements, especially for crack volume and relocation modelling. (orig./GL) [de

  18. 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.)

  19. 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

  20. 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.)

  1. 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)

  2. 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

  3. 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)

  4. 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

  5. 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

  6. 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)

  7. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Design and performance of reference cores. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely of tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  8. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1977--May 31, 1977

    International Nuclear Information System (INIS)

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

    1977-01-01

    Progress is summarized in the following tasks: (1) bundle flow studies (wrapped and bare rods); (2) subchannel flow studies (bare rods); (3) LMFBR outlet plenum flow mixing; and (4) theoretical determination of local temperature fields in LMFBR fuel rod bundles

  9. Performance evaluation of CPF shredder type mechanical crusher with simulated core fuel pin

    International Nuclear Information System (INIS)

    Nakahara, Masaumi; Sano, Yuichi; Aose, Shin-ichi

    2006-12-01

    In the advanced aqueous reprocessing system, powder fuel dissolution has been investigated, which is quite effective on the dissolution for highly concentrated solution. As one of the effective means that powder the irradiated MOX fuel, we have been developing shredder type mechanical crusher. This apparatus can automatically crush the sheared fuel pieces by twin-shaft disk blades, powder the crushed fragments by disk blades and screen blade, and recover the powdered fuel. The shredder type mechanical crusher was developed for using in a hot cell in Chemical Processing Facility, and the first crush experiment with this crusher was carried out at July 2004 using the simulated core fuel pin. This experiment showed that the crushed fragments could not be grinded efficiency because screen blade vibrated up and down during the operation. Additionally, the strength of screen blade block was insufficient to crush the sheared fuel pieces stably. Therefore, about 70% of fuel was recovered in maximum. Based on the results of the first experiment, screen blade was fixed up mainly and the second experiment was carried out with improved apparatus at September 2005. In this experiment, about 96% of fuel could be recovered in maximum because screen blade was stable during the operation. (J.P.N.)

  10. Prediction of the Long Term Cooling Performance for 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

    2005-12-15

    In the long term cooling phase that the emergency cooling water injection ends, the performance of the residual heat removal for the 3-pin fuel test loop has been predicted by a simplified heat transfer model. In the long term cooling phase the residual heat is 1323W for PWR fuel test mode and 1449W for CANDU fuel test mode. The each residual heat is assumed as 2% of the fission power of the test fuel used in the anticipated operational occurrence and design basis accident analyses. The each fission power used for the analyses is 105% of the rated fission power in the normal operation. In the long term cooling phase the residual heat is removed to the HANARO pool through the double pressure vessels of the in-pile test section. Saturate pooling boiling is assumed on the test fuel and condensation heat transfer is expected on the inner wall of the fuel carrier and the flow divider. Natural convection heat transfer on a heated vertical wall is also assumed on the outer wall of the outer pressure vessel. The conduction heat transfer is only considered in the gap between the double pressure vessels charged with neon gas and in the downcomer filled with coolant. The heat transfer rate between the coolant temperature of 152 .deg. C in the in-pile test section and the water temperature of 45 .deg. C in the HANARO pool is predicted as about 1666W. The 152 .deg. C is the saturate temperature of the coolant pressure predicted from the MARS code. The cooling capacity of 1666W is greater than the residual heats of 1323W and 1449W. Consequently the long term cooling performance of the 3-pin fuel test loop is sufficient for the anticipated operational occurrences and design basis accidents.

  11. 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)

  12. 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)

  13. Fuel pin failure root causes and power distribution gradients in WWER cores

    International Nuclear Information System (INIS)

    Mikus, J.

    2008-01-01

    The purpose of this work is to investigate the influence of some core heterogeneities and reactor construction materials on space power distribution in WWER type cores, especially from viewpoint of the values and gradient occurrence that could result in static loads with some consequences, e.g., fuel pin (FP) or fuel assembly (FA) bowing and possible contribution to the FP failure root causes. Presented information were obtained by means of experiments on research reactor LR-0 concerning the: 1) Power distribution estimation on pellet surface of the FPs neighbouring a FP containing gadolinium (Gd 2 O 3 ) burnable absorber integrated into fuel in WWER-440 and -1000 type cores; 2) Power distribution measurement in periphery FAs neighbouring the baffle in WWER-1000 type cores and 3) Power distribution in FAs neighbouring the control rod absorbing part in a WWER-440 type core. (author)

  14. Replaceable LMFBR core components

    International Nuclear Information System (INIS)

    Evans, E.A.; Cunningham, G.W.

    1976-01-01

    Much progress has been made in understanding material and component performance in the high temperature, fast neutron environment of the LMFBR. Current data have provided strong assurance that the initial core component lifetime objectives of FFTF and CRBR can be met. At the same time, this knowledge translates directly into the need for improved core designs that utilize improved materials and advanced fuels required to meet objectives of low doubling times and extended core component lifetimes. An industrial base for the manufacture of quality core components has been developed in the US, and all procurements for the first two core equivalents for FFTF will be completed this year. However, the problem of fabricating recycled plutonium while dramatically reducing fabrication costs, minimizing personnel exposure, and protecting public health and safety must be addressed

  15. 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)

  16. 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)

  17. Analysis of th SBLOCAs in the room 1 for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

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

    2004-10-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss of Coolant Accidents (SBLOCAs) in the Room 1 for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the downstream of the main cooling water pump and the upstream of the main cooler in the room 1. The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperature (PCT) is predicted to be about 931.4 .deg. C for the cold leg break accident in PWR fuel test mode and 931.6 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 8% of the cross section area of the pipe for PWR fuel test mode and the 10% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

  18. Analysis of the SBLOCAs in HANARO pool for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

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

    2004-09-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss Of Coolant Accidents (SBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperatures (PCT) are predicted to be about 906.9 .deg. C for the cold leg break accident in PWR fuel test mode and 971.9 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 6% of the cross section area of the pipe for PWR fuel test mode and the 8% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

  19. Features of the Numerical Solution of Thermal Destruction Fuel Pins Problems in the Fast Reactor

    Science.gov (United States)

    Usov, E. V.; Butov, A. A.; Klimonov, I. A.; Chuhno, V. I.; Nikolaenko, A. V.; Zhdanov, V. S.; Pribaturin, N. A.; Strizhov, V. F.

    2017-11-01

    In this paper the description of the basic equations which can be used for calculation of melting of fuel and cladding of the fast reactor, moving of the melt on a fuel pin surface and its solidification is presented. The special attention is given speed of calculation algorithms and fidelity of the phenomena which are observed at a stage of severe accidents in fast reactors. For check of working capacity of initial models, numerical calculations of Stefan-type problems on front movement of melting/solidification in cylindrical geometry are presented. Comparison with the solutions received by known analytical methods is executed. For validation of the numerical realization of calculation algorithms the analysis is carried out and experiments in which melting of the model fuel pins of fast reactors was studied are chosen. On the basis of the chosen experiments calculation schemes taking into account initial and boundary conditions are prepared and modeling is performed. Modeling results are shown in the present paper. Estimation of calculation error of the basic physical parameters is done by results of the modeling and conclusions are drawn on a correctness of algorithms operation.

  20. 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.)

  1. 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)

  2. COVE-1: a finite difference creep collapse code for oval fuel pin cladding material

    International Nuclear Information System (INIS)

    Mohr, C.L.

    1975-03-01

    COVE-1 is a time-dependent incremental creep collapse code that estimates the change in ovality of a fuel pin cladding tube. It uses a finite difference method of solving the differential equations which describe the deflection of the tube walls as a function of time. The physical problem is nonlinear, both with respect to geometry and material properties, which requires the use of an incremental, analytical, path-dependent solution. The application of this code is intended primarily for tubes manufactured from Zircaloy. Therefore, provision has been made to include some of the effects of anisotropy in the flow equations for inelastic incremental deformations. 10 references. (U.S.)

  3. Mechanical behavior of irradiated fuel-pin cladding evaluated under transient heating and pressure conditions

    International Nuclear Information System (INIS)

    Hamilton, M.L.; Johnson, G.D.; Hunter, C.W.; Duncan, D.R.

    1982-11-01

    Fast breeder fuel-pin cladding has been tested under experimental conditions simulating the temperature and pressure history characteristic of anticipated transient events. Irradiation induces severe reductions in both strength and ductility. Ductility losses are independent of the rate of temperature increase and saturate by a fluence of approx. 2 x 10 22 n/cm 2 (E > 0.1 MeV). Losses in strength are dependent on the rate of temperature increase but saturate at a fluence of approx.5 x 10 22 n/cm 2 . Evidence is presented to show that fission products are probably responsible for the degradation in mechanical properties

  4. 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)

  5. Status of LMFBR development project in Japan

    International Nuclear Information System (INIS)

    Nagane, G.; Akebi, M.; Matsuno, Y.

    1987-01-01

    Initiation of the LMFBR development project in Japan was decided by the Atomic Energy Commission of Japan in 1966. In 1967, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established to realize the project as a part of its tasks of a wide scope covering all the reseatch and development activities concerning fuel cycle. In the present paper the status of experimental fast reactor (Joyo), which is the first milestone of the LMFBR project, prototype fast reactor (Monju) and R and D activities supporting the project including that for larger LMFBRs in the future is described. (author)

  6. 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.

  7. Nondestructive characterization of mixed oxide pellets in welded nuclear fuel pins by neutron radiography and gamma-autoradiography

    International Nuclear Information System (INIS)

    Panakkal, J.P.; Ghosh, J.K.; Roy, P.R.

    1989-01-01

    Nondestructive evaluation of nuclear fuel pellets after the welding of fuel pins plays a vital role in assuring a safe and reliable operation of reactors. Some of the important characteristics to be monitored in low plutonium enriched mixed oxide fuel pellets are plutonium enrichment, size of plutonium dioxide agglomerates, incorrect loading and geometric shape. Experiments were carried out at Bhabha Atomic Research Centre, Bombay on experimental fuel pins containing mixed oxide pellets of different geometry (solid and annular), of different plutonium enrichment (0-6 w% of plutonium dioxide) and containing PuO 2 agglomerates of size 125-2000 microns to evaluate these characteristics nondestructively. Neutron radiography of these fuel pins was carried out using a swimming pool type reactor 'APSARA'. Results of quantitative evaluation of the neutron radiographs and a simple model correlating neutron interaction probability and the optical density are presented. Gamma autoradiography of these fuel pins showed that these parameters could be evaluated with a few limitations. This paper presents the experimental details, quantitative analysis of the radiographs by microdensitometry and merits and demerits of neutron radiography and gamma autoradiography for nondestructive charcterisation of nuclear fuel pellets. (orig.)

  8. 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

  9. 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

  10. 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.)

  11. 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

  12. Subchannel analysis of sodium-cooled reactor fuel assemblies with annular fuel pins

    International Nuclear Information System (INIS)

    Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

    2009-01-01

    Using a RELAP5-3D subchannel analysis model, the thermal-hydraulic behavior of sodium-cooled fuel assemblies with internally and externally cooled annular fuel rods was investigated, in an effort to enhance the economic performance of sodium-fast reactors by increasing the core power density, decreasing the core pressure drop, and extending the fuel discharge burnup. Both metal and oxide fuels at high and low conversion ratios (CR=0.25 and CR=1.00) were investigated. The externally and internally cooled annular fuel design is most beneficial when applied to the low CR core, as clad temperatures are reduced by up to 62.3degC for the oxide fuel, and up to 18.5degC for the metal fuel. This could result in a power uprates of up to ∼44% for the oxide fuel, and up to ∼43% for the metal fuel. The use of duct ribs was explored to flatten the temperature distribution at the core outlet. Subchannel analyses revealed that no fuel melting would occur in the case of complete blockage of the hot interior-annular channel for both metal and oxide fuels. Also, clad damage would not occur for the metal fuel if the power uprate is 38% or less, but would indeed occur for the oxide fuel. (author)

  13. Characterization of fuel swelling in helium-bonded carbide fuel pins

    International Nuclear Information System (INIS)

    Louie, D.L.Y.

    1987-08-01

    This work is not only the first attempt at characterizing the swelling of (U,Pu)C fuel pellets, but it also represents the only detailed examinations on carbide fuel swelling at high fuel burnups (4 to 16 at. %). This characterization includes the contributions of fission gases, cracks and solid fission products to fuel swelling. Significantly, the contributions of fission gases and cracks were determined by using the image analysis technique (IAT) which allows researchers to take areal measurements of the irradiated fuel porosity and cracks from the photographs of metallographic fuel samples. However, because areal measurements for varying depths in the fuel pellet could not be obtained, the crack areal measurements could not be converted into volumetric quantities. Consequently, in this situation, an areal fuel swelling analysis was used. The macroscopic fission-gas induced fuel swelling (MAS) caused by fission-gas bubbles and pores > 1 μm was determined using the measured irradiated fuel porosity because the measuring range of IAT is limited to bubbles and pores >1 μm. Conversely, for fuel swelling induced by fission-gas bubbles < 1 μm, the microscopic fission-gas induced fuel swelling (MIS) was estimated using an areal fuel swelling model

  14. Analysis of the LBLOCAs in the HANARO pool for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

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

    2004-12-01

    The Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Large Break Loss of Coolant Accidents (LBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the LBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). Double ended guillotine break is assumed for the large break loss of coolant accidents. The discharge coefficients of 0.1, 0.33, 0.67, 1.0 are investigated for the LBLOCAs. The test fuels for PWR and CANDU test modes are not heated up for the LBLOCAs caused by the double ended guillotine break in the HANARO pool. The reason is that the sufficient emergency cooling water to cool down the test fuels is supplied continuously to the in-pile test section. Therefore the PCTs for the LBLOCAs in the HANARO pool meet the design criterion of commercial PWR fuel that maximum PCT is lower than 1204 .deg. C

  15. Analysis of the LBLOCAs in the room 1 for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

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

    2004-12-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Large Break Loss of Coolant Accidents (LBLOCAs) in the Room 1 for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the LBLOCAs. The location of the pipe break is assumed at the downstream of the main cooling water pump and the upstream of the main cooler in the room 1. Double ended guillotine break is assumed for the large break loss of coolant accidents. The discharge coefficients of 0.1, 0.33, 0.67, 1.0 are investigated for the LBLOCAs. The maximum Peak Cladding Temperature (PCT) is predicted to be about 734.7 .deg. C for the PWR fuel test mode and 850.4 .deg. C for the CANDU fuel test mode respectively. The maximum peak cladding temperatures meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

  16. LMFBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1988-01-01

    Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

  17. Simulated fuel melt movement and relocation in two seven-pin-bundle geometries (SIMBATH out-of-pile experiments Vsub(Th) 74; Vsub(Th) 79)

    International Nuclear Information System (INIS)

    Peppler, W.; Menzenhauer, P.; Will, H.

    1983-11-01

    The SIMBATH program was initiated to investigate the physical phenomena of transient material movement and relocation during transient overpower (TOP) and loss of flow (LOF) driven TOP accidents in LMFBR's. The energy release during the accident is simulated out of pile by the reaction of a thermite mixture. Two 7-pin experiments have been performed using flowing sodium. In both tests the transient material movement has been recorded by X-ray-cinematography at about 1000 frames per second. The tests have been analysed. The dominant physical phenomena governing the material movement have been estimated. (orig.) [de

  18. FRESS pin failure model and its application to E-8 TREAT test

    International Nuclear Information System (INIS)

    Kalimullah.

    1979-01-01

    FRESS is a cladding rupture prediction model for an irradiated mixed-oxide LMFBR fuel pin during transient heating based only on the internal pressurization of the cladding by the fission gas released from fuel grains during the transient. The model is applied to the analysis of the hottest PNL-10-53 pin in the 7-pin E-8 TREAT test which simulates a $3/sec transient overpower. Although the uncertainties of the inputs to the temperature calculation done with the COBRA code have not been included, the uncertain input parameters to FRESS have been varied over their estimated uncertainties. The cladding rupture predictions are a few tens of milliseconds late compared to the most probable failure time detected in the test. However, these calculations seem to indicate that fisson gas pressure is a significant mechanism for causing clad rupture in this test

  19. 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

  20. 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.

  1. 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

  2. Cause of defect in the end plug welding of the JOYO fuel pin

    International Nuclear Information System (INIS)

    Ouchi, Masaru; Otani, Seiji; Onisi, Koichi; Tateisi, Yoshinori; Ikawa, Yukio.

    1976-01-01

    About twelve thousand fuel pins for the JOYO core fuel were fabricated, and their end plug welding was inspected by X-ray radiography. The defect fractions were 0.2 percent for the lower end plugs and 1.8 percent for the upper, respectively. It had been known that the defect was due to ''line porosity''. In this study, the cause of the ''line porosity defect'' was investigated by the welding experiment performed on some dummy specimens of three different types; open end; closed end; and closed end with dummy pellets and a spring. The position of electrodes was varied for changing the arc gap from 0.3 mm to 1.2 mm. The experimental results are summarized in tables. The results showed that no defect was found in the open end type specimens even with the arc gap of 1.2 mm. Whereas in the other two types of specimens, the defect fraction of 60 to 75 percent was observed with the same arc gap. As for the effect of the arc gap, it was shown that 0.3 mm is the best among 0.3 mm, 0.5 mm and 1.2 mm. No defect was observed in the third type of specimens with the arc gap of 0.3 mm. In summary, it was found that the line porosity defect did not depend on the shape of the end plugs. It is considered to be dependent on both the structure of dummy fuel pins and the position of electrodes. (Aoki, K.)

  3. THE APPLICATION OF MAMMOTH FOR A DETAILED TIGHTLY COUPLED FUEL PIN SIMULATION WITH A STATION BLACKOUT

    Energy Technology Data Exchange (ETDEWEB)

    Gleicher, Frederick; Ortensi, Javier; DeHart, Mark; Wang, Yaqi; Schunert, Sebastian; Novascone, Stephen; Hales, Jason; Williamson, Rich; Slaughter, Andrew; Permann, Cody; Andrs, David; Martineau, Richard

    2016-09-01

    Accurate calculation of desired quantities to predict fuel behavior requires the solution of interlinked equations representing different physics. Traditional fuels performance codes often rely on internal empirical models for the pin power density and a simplified boundary condition on the cladding edge. These simplifications are performed because of the difficulty of coupling applications or codes on differing domains and mapping the required data. To demonstrate an approach closer to first principles, the neutronics application Rattlesnake and the thermal hydraulics application RELAP-7 were coupled to the fuels performance application BISON under the master application MAMMOTH. A single fuel pin was modeled based on the dimensions of a Westinghouse 17x17 fuel rod. The simulation consisted of a depletion period of 1343 days, roughly equal to three full operating cycles, followed by a station blackout (SBO) event. The fuel rod was depleted for 1343 days for a near constant total power loading of 65.81 kW. After 1343 days the fission power was reduced to zero (simulating a reactor shut-down). Decay heat calculations provided the time-varying energy source after this time. For this problem, Rattlesnake, BISON, and RELAP-7 are coupled under MAMMOTH in a split operator approach. Each system solves its physics on a separate mesh and, for RELAP-7 and BISON, on only a subset of the full problem domain. Rattlesnake solves the neutronics over the whole domain that includes the fuel, cladding, gaps, water, and top and bottom rod holders. Here BISON is applied to the fuel and cladding with a 2D axi-symmetric domain, and RELAP-7 is applied to the flow of the circular outer water channel with a set of 1D flow equations. The mesh on the Rattlesnake side can either be 3D (for low order transport) or 2D (for diffusion). BISON has a matching ring structure mesh for the fuel so both the power density and local burn up are copied accurately from Rattlesnake. At each depletion time

  4. 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

  5. Calculation of heat rating and burn-up for test fuel pins irradiated in DR 3

    International Nuclear Information System (INIS)

    Bagger, C.; Carlsen, H.; Hansen, K.

    1980-01-01

    A summary of the DR 3 reactor and HP1 rig design is given followed by a detailed description of the calculation procedure for obtaining linear heat rating and burn-up values of fuel pins irradiated in HP1 rigs. The calculations are carried out rather detailed, especially regarding features like end pellet contribution to power as a function of burn-up, gamma heat contributions, and evaluation of local values of heat rating and burn-up. Included in the report is also a description of the fast flux- and cladding temperature calculation techniques currently used. A good agreement between measured and calculated local burn-up values is found. This gives confidence to the detailed treatment of the data. (author)

  6. Microcontroller based instrumentation for the fuel pin preparation facility by sol-gel method

    International Nuclear Information System (INIS)

    Suhasini, B.; Prabhakar Rao, J.; Srinivas, K.C.

    2009-01-01

    The fuel pin preparation facility by Sol-Gel route has been set up at Chemistry Group at Indira Gandhi Centre for Atomic Research, Kalpakkam. Sol-Gel, a solution-gelation process involves conversion of solutions of nitrates of uranium-plutonium (at 0 deg C) into gel microspheres. To measure the exact quantities of the above solutions and to ensure their temperatures, a variety of sensors have been used at various stages in the plant. To monitor and acquire the data of process parameters used in the production and for an automated operation of the plant, a PC (master)-microcontroller (slave) based instrumentation has been developed along with acquisition software and a GU interface developed in Visual Basic. (author)

  7. A mechanical deformation model of metallic fuel pin under steady state conditions

    International Nuclear Information System (INIS)

    Lee, D. W.; Lee, B. W.; Kim, Y. I.; Han, D. H.

    2004-01-01

    As a mechanical deformation model of the MACSIS code predicts the cladding deformation due to the simple thin shell theory, it is impossible to predict the FCMI(Fuel-Cladding Mechanical Interaction). Therefore, a mechanical deformation model used the generalized plane strain is developed. The DEFORM is a mechanical deformation routine which is used to analyze the stresses and strains in the fuel and cladding of a metallic fuel pin of LMRs. The accuracy of the program is demonstrated by comparison of the DEFORM predictions with the result of another code calculations or experimental results in literature. The stress/strain distributions of elastic part under free thermal expansion condition are completely matched with the results of ANSYS code. The swelling and creep solutions are reasonably well agreed with the simulations of ALFUS and LIFE-M codes, respectively. The predicted cladding strains are under estimated than experimental data at the range of high burnup. Therefore, it is recommended that the fine tuning of the DEFORM based on various range of experimental data

  8. 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.

  9. Development of unfolding method to obtain pin-wise source strength distribution from PWR spent fuel assembly measurement

    International Nuclear Information System (INIS)

    Sitompul, Yos Panagaman; Shin, Hee-Sung; Park, Se-Hwan; Oh, Jong Myeong; Seo, Hee; Kim, Ho Dong

    2013-01-01

    An unfolding method has been developed to obtain a pin-wise source strength distribution of a 14 × 14 pressurized water reactor (PWR) spent fuel assembly. Sixteen measured gamma dose rates at 16 control rod guide tubes of an assembly are unfolded to 179 pin-wise source strengths of the assembly. The method calculates and optimizes five coefficients of the quadratic fitting function for X-Y source strength distribution, iteratively. The pin-wise source strengths are obtained at the sixth iteration, with a maximum difference between two sequential iterations of about 0.2%. The relative distribution of pin-wise source strength from the unfolding is checked using a comparison with the design code (Westinghouse APA code). The result shows that the relative distribution from the unfolding and design code is consistent within a 5% difference. The absolute value of the pin-wise source strength is also checked by reproducing the dose rates at the measurement points. The result shows that the pin-wise source strengths from the unfolding reproduce the dose rates within a 2% difference. (author)

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

    International Nuclear Information System (INIS)

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

    1976-09-01

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

  11. 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

  12. 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)

  13. 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)

  14. 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

  15. 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}.

  16. CEC activities in the field of LMFBR safety

    International Nuclear Information System (INIS)

    Balz, W.; Finzi, S.; Klersy, R.

    1976-01-01

    The aim of the ECC is to reach a common LMFBR Safety strategy in Europe. To this end the Commission promotes collaboration between the different fast reactor projects in the Community through working groups and collaborative arrangements and contributes with a research activity executed in its Joint Research Centre Ispra. A short description is given of the activity in the working groups and of the Ispra programme on LMFBR Safety. This programme covers: LMFBR thermohydraulics, fuel coolant interactions, dynamic structure loading and response, safety related material properties and whole core accident code development

  17. 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)

  18. 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)

  19. The coupling algorithm between fuel pin and coolant channel in the European Accident Code EAC-2

    International Nuclear Information System (INIS)

    Goethem, G. van; Lassmann, K.

    1989-01-01

    In the field of fast breeder reactors the Commission of the European Communities (CEC) is conducting coordination and harmonisation activities as well as its own research at the CEC's Joint Research Centre (JRC). The development of the modular European Accident Code (EAC) is a typical example of concerted action between EC Member States performed under the leadership of the JRC. This computer code analyzes the initiation phase of low-probability whole-core accidents in LMFBRs with the aim of predicting the rapidity of sodium voiding, the mode of pin failure, the subsequent fuel redistribution and the associated energy release. This paper gives a short overview on the development of the EAC-2 code with emphasis on the coupling mechanism between the fuel behaviour module TRANSURANUS and the thermohydraulics modules which can be either CFEM or BLOW3A. These modules are also briefly described. In conclusion some numerical results of EAC-2 are given: they are recalculations of an unprotected LOF accident for the fictitious EUROPE fast breeder reactor which was earlier analysed in the frame of a comparative exercise performed in the early 80s and organised by the CEC. (orig.)

  20. Large Eddy Simulation of turbulent flow in wire wrapped fuel pin bundles cooled by sodium

    International Nuclear Information System (INIS)

    Saxena, Aakanksha; Cadiou, Thierry; Bieder, Ulrich; Viazzo, Stephane

    2013-06-01

    The objective of the study is to understand the thermal hydraulics in a core sub-assembly with liquid sodium as coolant by performing detailed numerical simulations. The passage for the coolant flow between the fuel rods is maintained by thin wires wrapped around the rods. The contact point between the fuel pin and the spacer wire is the region of creation of hot spots and a cyclic variation of temperature in hot spots can adversely affect the mechanical properties of the clad due to the phenomena like thermal stripping. The current status quo provides two different models to perform the numerical simulations, namely Reynolds Averaged Navier-Stokes (RANS) and Large Eddy Simulation (LES). The two models differ in the extent of modelling used to close the Navier-Stokes equations. LES is a filtered approach where the large scale of motions are explicitly resolved while the small scale motions are modelled whereas RANS is a time averaging approach where all scale of motions are modelled. Thus LES involves less modelling as compared to RANS and so the results are comparatively more accurate. An attempt has been made to use the LES model. The simulations have been performed using the code Trio-U (developed by CEA). The turbulent statistics of the flow and thermal quantities are calculated. Finally the goal is to obtain the frequency of temperature oscillations at the region of hot spots near the spacer wire. (authors)

  1. 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)

  2. LMFBR plant parameters

    International Nuclear Information System (INIS)

    1979-03-01

    This document contains up-to-date data on existing or firmly decided prototype or demonstration LMFBR reactors (Table I), on planned commercial size LMFBR according to the present status of design (Table II) and on experimental fast reactors such as BOR-60, DFR, EBR-II, FERMI, FFTF, JOYO, KNK-II, PEC, RAPSODIE-FORTISSIMO (Table III). Only corrected and revised parameters submitted by the countries participating in the IWGFR are included in this document

  3. 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

  4. VERA Pin and Fuel Assembly Depletion Benchmark Calculations by McCARD and DeCART

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin; Cho, Jin Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Monte Carlo (MC) codes have been developed and used to simulate a neutron transport since MC method was devised in the Manhattan project. Solving the neutron transport problem with the MC method is simple and straightforward to understand. Because there are few essential approximations for the 6- dimension phase of a neutron such as the location, energy, and direction in MC calculations, highly accurate solutions can be obtained through such calculations. In this work, the VERA pin and fuel assembly (FA) depletion benchmark calculations are performed to examine the depletion capability of the newly generated DeCART multi-group cross section library. To obtain the reference solutions, MC depletion calculations are conducted using McCARD. Moreover, to scrutinize the effect by stochastic uncertainty propagation, uncertainty propagation analyses are performed using a sensitivity and uncertainty (S/U) analysis method and stochastic sampling (S.S) method. It is still expensive and challenging to perform a depletion analysis by a MC code. Nevertheless, many studies and works for a MC depletion analysis have been conducted to utilize the benefits of the MC method. In this study, McCARD MC and DeCART MOC transport calculations are performed for the VERA pin and FA depletion benchmarks. The DeCART depletion calculations are conducted to examine the depletion capability of the newly generated multi-group cross section library. The DeCART depletion calculations give excellent agreement with the McCARD reference one. From the McCARD results, it is observed that the MC depletion results depend on how to split the burnup interval. First, only to quantify the effect of the stochastic uncertainty propagation at 40 DTS, the uncertainty propagation analyses are performed using the S/U and S.S. method.

  5. 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)

  6. DEFORM-4: fuel pin characterization and transient response in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Miles, K.J.; Hill, D.J.

    1986-01-01

    The DEFORM-4 module is the segment of the SAS4A Accident Analysis Code System that calculates the fuel pin characterization in response to a steady state irradiation history, thereby providing the initial conditions for the transient calculation. The various phenomena considered include fuel porosity migration, fission gas bubble induced swelling, fuel cracking and healing, fission gas release, cladding swelling, and the thermal-mechanical state of the fuel and cladding. In the transient state, the module continues the thermal-mechanical response calculation, including fuel melting and central cavity pressurization, until cladding failure is predicted and one of the failed fuel modules is initiated. Comparisons with experimental data have demonstrated the validity of the modeling approach

  7. 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.

  8. 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

  9. Verification of the FBR fuel bundle-duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

    Science.gov (United States)

    Uwaba, Tomoyuki; Ito, Masahiro; Nemoto, Junichi; Ichikawa, Shoichi; Katsuyama, Kozo

    2014-09-01

    The BAMBOO computer code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle-duct interaction (BDI) condition. The pin diameters of the examined test bundles were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype fast breeder reactor (FBR) and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT) images and local parameters of bundle deformation such as pin-to-duct and pin-to-pin clearances were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms, the same as in the case of small diameter pin bundles. In addition, the BAMBOO analysis results confirmed that cladding oval distortion effectively suppresses BDI in large diameter pin bundles as well as in small diameter pin bundles.

  10. Verification of the FBR fuel bundle–duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

    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; Nemoto, Junichi [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ichikawa, Shoichi [Japan Atomic Energy Agency, 2-1, Shiraki, Tsuruga-shi, Fukui 919-1279 (Japan); Katsuyama, Kozo [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

    2014-09-15

    The BAMBOO computer code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle–duct interaction (BDI) condition. The pin diameters of the examined test bundles were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype fast breeder reactor (FBR) and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT) images and local parameters of bundle deformation such as pin-to-duct and pin-to-pin clearances were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms, the same as in the case of small diameter pin bundles. In addition, the BAMBOO analysis results confirmed that cladding oval distortion effectively suppresses BDI in large diameter pin bundles as well as in small diameter pin bundles.

  11. Equation of state for L.M.F.B.R. fuel (measurement of fission gas release during transients)

    International Nuclear Information System (INIS)

    Combette, P.; Barthelemy, P.

    1979-01-01

    A sample of fuel (UO 2 or UPuO 2 ) can be heated by fission in a heating transient up to energy deposition 4000 j/g, in the Silene reactor. The Kistler type capsule, the calorimeter device and the radiochemical analysis of fission products enable the pressure pulse and the fuel energy deposition to be measured. So, the relationship between the fuel vapour pressure and the fuel specific energy can be deduced. Peaks pressure (about 1 MPa) coming from fresh UO 2 vaporization, have been measured on a 7 milliseconds time scale. There is a good agreement with the E.O.S. for fresh UO 2 , which is well known for low pressure (1 MPa). Numerous tests have been done with 93% enriched UO 2 and a first test with highly active fuel containing plutonium (15 at %) has been performed. The capsule allows the released gas coming from the irradiated fuel to be retained for measurements and analysis. To investigate the mode of fuel disruption, in-pile fission-heated fuel pellets has been recorded by high speed cinematography

  12. Improvement of the computing speed of the FBR fuel pin bundle deformation analysis code 'BAMBOO'

    International Nuclear Information System (INIS)

    Ito, Masahiro; Uwaba, Tomoyuki

    2005-04-01

    JNC has developed a coupled analysis system of a fuel pin bundle deformation analysis code 'BAMBOO' and a thermal hydraulics analysis code ASFRE-IV' for the purpose of evaluating the integrity of a subassembly under the BDI condition. This coupled analysis took much computation time because it needs convergent calculations to obtain numerically stationary solutions for thermal and mechanical behaviors. We improved the computation time of the BAMBOO code analysis to make the coupled analysis practicable. 'BAMBOO' is a FEM code and as such its matrix calculations consume large memory area to temporarily stores intermediate results in the solution of simultaneous linear equations. The code used the Hard Disk Drive (HDD) for the virtual memory area to save Random Access Memory (RAM) of the computer. However, the use of the HDD increased the computation time because Input/Output (I/O) processing with the HDD took much time in data accesses. We improved the code in order that it could conduct I/O processing only with the RAM in matrix calculations and run with in high-performance computers. This improvement considerably increased the CPU occupation rate during the simulation and reduced the total simulation time of the BAMBOO code to about one-seventh of that before the improvement. (author)

  13. Assessment of LMFBR spent fuel shipping cask concepts for the CRBRP and the US conceptual design study

    International Nuclear Information System (INIS)

    Pope, R.B.; Ortman, J.M.; Eakes, R.G.; Leisher, W.B.; Dupree, S.A.

    1980-01-01

    Study of conceptual shipping systems for CRBRP and CDS spent fuel has shown that systems significantly different from those used for LWR spent fuel will be required. In the conceptual design, liquid sodium was assumed to be the coolant in canisters containing the spent fuel assemblies, and multiple levels of containment were provided by canisters, an inner cask lid and an outer cask lid. Cask cooling at the reactor site during loading, and cooldown at the receiving site prior to unloading are significant but tractable problems

  14. Implementation into a CFD code of neutron kinetics and fuel pin models for nuclear reactor transient analyses

    International Nuclear Information System (INIS)

    Chen Zhao; Chen, Xue-Nong; Rineiski, Andrei; Zhao Pengcheng; Chen Hongli

    2014-01-01

    Safety analysis is an important tool for justifying the safety of nuclear reactors. The traditional method for nuclear reactor safety analysis is performed by means of system codes, which use one-dimensional lumped-parameter method to model real reactor systems. However, there are many multi-dimensional thermal-hydraulic phenomena cannot be predicated using traditional one-dimensional system codes. This problem is extremely important for pool-type nuclear systems. Computational fluid dynamics (CFD) codes are powerful numerical simulation tools to solve multi-dimensional thermal-hydraulics problems, which are widely used in industrial applications for single phase flows. In order to use general CFD codes to solve nuclear reactor transient problems, some additional models beyond general ones are required. Neutron kinetics model for power calculation and fuel pin model for fuel pin temperature calculation are two important models of these additional models. The motivation of this work is to develop an advance numerical simulation method for nuclear reactor safety analysis by implementing neutron kinetics model and fuel pin model into general CFD codes. In this paper, the Point Kinetics Model (PKM) and Fuel Pin Model (FPM) are implemented into a general CFD code FLUENT. The improved FLUENT was called as FLUENT/PK. The mathematical models and implementary method of FLUENT/PK are descripted and two demonstration application cases, e.g. the unprotected transient overpower (UTOP) accident of a Liquid Metal cooled Fast Reactor (LMFR) and the unprotected beam overpower (UBOP) accident of an Accelerator Driven System (ADS), are presented. (author)

  15. On the problems relating to the accuracy of the measurement of fuel pin diameters by neutron radiography

    International Nuclear Information System (INIS)

    Matfield, R.

    1983-01-01

    The paper identifies the sources of error in the neutron radiographic system and attempts to estimate some of these errors. The sources of error are in the fuel pin materials, the radiographic set-up, the radiographic equipment, image formation, the microdensitometer, the edge criteria and the dimensional measurement from the microdensitometer trace. However, the critical problem area is that of determining a representative edge criteria and upon this will depend the ability of the method to achieve the required measurement accuracy. (Auth.)

  16. 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)

  17. Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1990-01-01

    Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel

  18. Transactions of the second technical exchange meeting on fuel- and clad-motion diagnostics for LMFBR safety test facilities

    Energy Technology Data Exchange (ETDEWEB)

    DeVolpi, A. (comp.)

    1976-01-01

    Papers are presented which deal with diagnostic requirements and fuel motion monitoring capabilities of hodoscopes, coded aperture systems, x-ray radiography, and in-core detectors. Separate abstracts and indexing were prepared for each paper. (DG)

  19. 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)

  20. SIEX: a correlated code for the prediction of liquid metal fast breeder reactor (LMFBR) fuel thermal performance

    International Nuclear Information System (INIS)

    Dutt, D.S.; Baker, R.B.

    1975-06-01

    The SIEX computer program is a steady state heat transfer code developed to provide thermal performance calculations for a mixed-oxide fuel element in a fast neutron environment. Fuel restructuring, fuel-cladding heat conduction and fission gas release are modeled to provide assessment of the temperature. Modeling emphasis has been placed on correlations to measurable quantities from EBR-II irradiation tests and the inclusion of these correlations in a physically based computational scheme. SIEX is completely modular in construction allowing the user options for material properties and correlated models. Required code input is limited to geometric and environmental parameters, with a ''consistent'' set of material properties and correlated models provided by the code. 24 references. (U.S.)

  1. Component design for LMFBR's

    International Nuclear Information System (INIS)

    Fillnow, R.H.; France, L.L.; Zerinvary, M.C.; Fox, R.O.

    1975-01-01

    Just as FFTF has prototype components to confirm their design, FFTF is serving as a prototype for the design of the commercial LMFBR's. Design and manufacture of critical components for the FFTF system have been accomplished primarily using vendors with little or no previous experience in supplying components for high temperature sodium systems. The exposure of these suppliers, and through them a multitude of subcontractors, to the requirements of this program has been a necessary and significant step in preparing American industry for the task of supplying the large mechanical components required for commercial LMFBR's

  2. 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)

  3. Fission-gas bubble modeling for LMFBR accidents

    International Nuclear Information System (INIS)

    Ostensen, R.W.

    1977-01-01

    The behavior of fission-gas bubbles in unrestructured oxide fuel can have a dominant effect on the course of a core disruptive accident in an LMFBR. The paper describes a simplified model of bubble behavior and presents results of that model in analyzing the relevant physical assumptions and predicting gas behavior in molten fuel

  4. Assessment of accident energetics in LMFBR core-disruptive accidents

    International Nuclear Information System (INIS)

    Fauske, H.K.

    1977-01-01

    An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated core meltdown event

  5. 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)

  6. Thermal analysis methods for LMFBR wire wrapped bundles

    International Nuclear Information System (INIS)

    Todreas, N.E.

    1976-11-01

    A note is presented which was written to stimulate an awareness and discussion of the fundamental differences in the formulation of certain existing analysis codes for LMFBR wire wrap bundles. The contention of the note is that for those array types where data exists (one wire per pin, equal start angles), the ENERGY method results for coolant temperature under forced convection conditions provide benchmarks of reliability equal to the results of codes COBRA and TH1-3D

  7. Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

    Science.gov (United States)

    Bowles, K. J.; Gluyas, R. E.

    1975-01-01

    The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

  8. 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

  9. 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

  10. Measurements of dynamic shape factors of LMFBR aggregate aerosols

    International Nuclear Information System (INIS)

    Allen, M.D.; Moss, O.R.; Briant, J.K.

    1980-01-01

    Dynamic shape factors for branched, chain-like aggregates of LMFBR mixed-oxide fuels have been measured with a LAPS spiral-duct centrifuge. The aerosol was generated by repeatedly pulsing a focused laser beam onto the surface of a typical LMFBR fuel pellet. The measured values of the dynamic shape factor, corrected for slip, vary between kappa = 3.60 at D/sub ae/ = 0.5 μm, and kappa = 2.23 at D/sub ae/ = 1.5 μm

  11. 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)

  12. Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

    International Nuclear Information System (INIS)

    Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

    1982-01-01

    The document reports in detail the fuel pin fabrication data and describes the irradiation conditions and history. All the relevant results of the non-destructive and destructive post-irradiation examinations are reported. They include: visual inspection and chemical analysis of crud; length and diameter measurements; neutron radiography and gamma scanning; juncture tests and fission gas analysis (including residual gas in fuel samples); microscopy and alpha + beta/gamma autoradiography; microprobe investigations; burn-up and isotopic analysis; and hydrogen analysis in clad. The data and observations obtained are discussed in detail and conclusions are given. The irradiation and post-irradiation examinations of the R-109 pins have shown the safe, pre-calculable performance of LWR fuel pins containing mixed-oxide sphere-pac fuel with the fissile material mainly present in the large spheres

  13. 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)

  14. 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)

  15. Fluid-mixing studies in a hexagonal 217-pin wire-wrapped rod bundle

    International Nuclear Information System (INIS)

    Symolon, P.D.; Todreas, N.E.

    1981-02-01

    Mixing, pressure drop, and flow split experiments were performed on a 217 pin LMFBR fuel bundle with a pitch to diameter ratio of 1.25 and a lead length of 12 inches. It was found that the turbulent flow data could best be characterized by the energy parameter C/sub 1L/=.106, which is 9% higher than the value from the correlation of Chiu et al. Chiu's correlation was developed on a data base of 61 and 91 pins. The spread of existing data about the correlation is +- 25%, but the error band on our data is expected to be less (approx. +- 10% since injection depth effects were not previously considered). This result is consistent with the concept of increased swirl flow in larger bundles

  16. LMFBR: safety aspects

    International Nuclear Information System (INIS)

    Natta, M.

    1990-01-01

    This presentation of LMFBR safety is limited at Super Phenix reactor. After a brief description of the reactor, some details on safety systems, in normal or accidental conditions, are given. The main functions studied are: chain reaction trip, residual power evacuation, reactor containment. In heavy accident the behaviour of Super Phenix is studied which its particular characteristics and the possibilities of operators reactions. The probability of appearance and the maximum consequences of heavy accidents are given [fr

  17. Sensitivity analysis of power depression and axial power factor effect on fuel pin to temperature and related properties distribution

    International Nuclear Information System (INIS)

    Suwardi, S.

    2001-01-01

    The presented paper is a preliminary step to evaluate the effect of radial and axial distribution of power generation on thermal analysis of whole fuel pin model with large L/D ratio. The model takes into account both radial and axial distribution of power generation due to power depression and core geometry, temperature and microstructure dependent on thermal conductivity. The microstructure distribution and the gap conductance for typical steady-state situation are given for the sensitivity analysis. The temperature and thermal conductivity distribution along the radial and axial directions obtained by different power distribution is used to indicate the sensitivity of power depression and power factor on thermal aspect. The evaluation is made for one step of incremental time and steady state approach is used. The analysis has been performed using a finite element-finite difference model. The result for typical reactor fuel shows that the sensitivity is too important to be omitted in thermal model

  18. An Optimization Study of LWR Fuel Assembly Design for TRU Burning using FCM and UO{sub 2}-ThO{sub 2} Fuel Pins

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Daehee; Hong, Ser Gi [Kyung Hee Univ., Yongin (Korea, Republic of)

    2014-05-15

    The objective of this work is to design optimized LWR fuel assemblies for the transmutation of TRU (transuranic) nuclides by using FCM (Fully Ceramic Micro-encapsulated) and UO{sub 2}-ThO{sub 2} fuel pins without degradation of safety-related parameters. In our study, the pin pitch (equivalently to P/D (Pitch-to-Diameter) ratio with a fixed fuel rod diameter) is used as a design parameter. The motivation is to make MTC (Moderator Temperature Coefficient) less negative at EOC because it was found that the small LWR core design in our previous work has a very strong MTC at EOC (∼-80pcm/K) which can lead to a large positive reactivity insertion under MSLB (Main Steam Line Break) accident and to a reduction of shutdown margin of the control rods. The basic idea is to increase moderator-to-fuel ratio such that the fuel assemblies have less negative MTC due to increase the moderation. The results show that a small increase of P/D ratio by 3.8% can give a considerably less negative MTC and an increase of TRU destruction rate without an increase of pin power peaking. In our study, a special emphasis is given on the effects of the increased P/D ratio for MTC. From the results, it was found that an increase of P/D ratio (we considered up to P/D=1.38) leads to a less negative MTC and a less negative FTC, an increase of TRU destruction rate, and a decrease of {sup 233}U production in UO{sub 2}-ThO{sub 2} pins. In particular, a small change of P/D ratio from 1.33 to 1.38 led to a change of MTC from - 75 pcm/.deg. C to -67 pcm/.deg. C at EOC, and a small increase of net TRU destruction rate from 26.4% to 28.3%. As conclusion, a small increase of P/D ratio is effective in obtaining the less negative MTC at EOC with a small increase of TRU destruction rate and without a significant degradation of FTC.

  19. 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)

  20. Heat transfer on HLM cooled wire-spaced fuel pin bundle simulator in the NACIE-UP facility

    Energy Technology Data Exchange (ETDEWEB)

    Di Piazza, Ivan, E-mail: ivan.dipiazza@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Angelucci, Morena; Marinari, Ranieri [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy); Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Forgione, Nicola [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy)

    2016-04-15

    Highlights: • Experiments with a wire-wrapped 19-pin fuel bundle cooled by LBE. • Wall and bulk temperature measurements at three axial positions. • Heat transfer and error analysis in the range of low mass flow rates and Péclet number. • Comparison of local and section-averaged Nusselt number with correlations. - Abstract: The NACIE-UP experimental facility at the ENEA Brasimone Research Centre (Italy) allowed to evaluate the heat transfer coefficient of a wire-spaced fuel bundle cooled by lead-bismuth eutectic (LBE). Lead or lead-bismuth eutectic are very attractive as coolants for the GEN-IV fast reactors due to the good thermo-physical properties and the capability to fulfil the GEN-IV goals. Nevertheless, few experimental data on heat transfer with heavy liquid metals (HLM) are available in literature. Furthermore, just a few data can be identified on the specific topic of wire-spaced fuel bundle cooled by HLM. Additional analysis on thermo-fluid dynamic behaviour of the HLM inside the subchannels of a rod bundle is necessary to support the design and safety assessment of GEN. IV/ADS reactors. In this context, a wire-spaced 19-pin fuel bundle was installed inside the NACIE-UP facility. The pin bundle is equipped with 67 thermocouples to monitor temperatures and analyse the heat transfer behaviour in different sub-channels and axial positions. The experimental campaign was part of the SEARCH FP7 EU project to support the development of the MYRRHA irradiation facility (SCK-CEN). Natural and mixed circulation flow regimes were investigated, with subchannel Reynolds number in the range Re = 1000–10,000 and heat flux in the range q″ = 50–500 kW/m{sup 2}. Local Nusselt numbers were calculated for five sub-channels in different ranks at three axial positions. Section-averaged Nusselt number was also defined and calculated. Local Nusselt data showed good consistency with some of the correlation existing in literature for heat transfer in liquid metals

  1. Heat transfer on HLM cooled wire-spaced fuel pin bundle simulator in the NACIE-UP facility

    International Nuclear Information System (INIS)

    Di Piazza, Ivan; Angelucci, Morena; Marinari, Ranieri; Tarantino, Mariano; Forgione, Nicola

    2016-01-01

    Highlights: • Experiments with a wire-wrapped 19-pin fuel bundle cooled by LBE. • Wall and bulk temperature measurements at three axial positions. • Heat transfer and error analysis in the range of low mass flow rates and Péclet number. • Comparison of local and section-averaged Nusselt number with correlations. - Abstract: The NACIE-UP experimental facility at the ENEA Brasimone Research Centre (Italy) allowed to evaluate the heat transfer coefficient of a wire-spaced fuel bundle cooled by lead-bismuth eutectic (LBE). Lead or lead-bismuth eutectic are very attractive as coolants for the GEN-IV fast reactors due to the good thermo-physical properties and the capability to fulfil the GEN-IV goals. Nevertheless, few experimental data on heat transfer with heavy liquid metals (HLM) are available in literature. Furthermore, just a few data can be identified on the specific topic of wire-spaced fuel bundle cooled by HLM. Additional analysis on thermo-fluid dynamic behaviour of the HLM inside the subchannels of a rod bundle is necessary to support the design and safety assessment of GEN. IV/ADS reactors. In this context, a wire-spaced 19-pin fuel bundle was installed inside the NACIE-UP facility. The pin bundle is equipped with 67 thermocouples to monitor temperatures and analyse the heat transfer behaviour in different sub-channels and axial positions. The experimental campaign was part of the SEARCH FP7 EU project to support the development of the MYRRHA irradiation facility (SCK-CEN). Natural and mixed circulation flow regimes were investigated, with subchannel Reynolds number in the range Re = 1000–10,000 and heat flux in the range q″ = 50–500 kW/m"2. Local Nusselt numbers were calculated for five sub-channels in different ranks at three axial positions. Section-averaged Nusselt number was also defined and calculated. Local Nusselt data showed good consistency with some of the correlation existing in literature for heat transfer in liquid metals for

  2. Tracking of fuel particles after pin failure in nominal, loss-of-flow and shutdown conditions in the MYRRHA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, Sophia; Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Van Tichelen, Katrien [SCK- CEN, Boeretang 200, 2400 Mol (Belgium)

    2017-02-15

    Highlights: • Quantification of the design and safety of the MYRRHA reactor in the event of a pin failure. • Simulation of different accident scenarios in both forced and natural convection regime. • The accumulation areas at the free-surface in case of the least dense particles depend on the flow regime. • The densest particles form an important deposit at the bottom of the vessel. • Further study of the risk of core blockage requires a detailed model of the core. - Abstract: This work on fuel dispersion aims at quantifying the design and safety of the MYRRHA nuclear reactor. A number of accidents leading to the release of a secondary phase into the primary coolant loop are investigated. Among these scenarios, an incident leading to the failure of one or more of the fuel pins is simulated while the reactor is operating in nominal conditions, but also in natural convection regime either during accident transients such as loss-of-flow or during the normal shut-down of the reactor. Two single-phase CFD models of the MYRRHA reactor are constructed in ANSYS Fluent to represent the reactor in nominal and natural convection conditions. An Euler–Lagrange approach with one-way coupling is used for the flow and particle tracking. Firstly, a steady state RANS solution is obtained for each of the three conditions. Secondly, the particles are released downstream from the core outlet and particle distributions are provided over the coolant circuit. Their size and density are defined such that test cases represent potential extremes that may occur. Analysis of the results highlights different particle behaviors, depending essentially on gravity forces and kinematic effects. Statistical distributions highlight potential accumulation regions that may form at the free-surfaces, on top of the upper diaphragm plate or at the bottom of the vessel. These results help to localize regions of fuel accumulation in order to provide insight for development of strategies for

  3. Thermochemical data and its use in modeling chemical behavior in mixed-oxide fuel pins

    International Nuclear Information System (INIS)

    Gibby, R.L.; Woodley, R.E.; Adamson, M.G.; Johnson, C.E.

    1979-01-01

    The status of US activities to obtain fuel chemistry data is reviewed. Analytical expressions addressing basic needs of all fuel chemistry models are presented. Fission product concentrations during irradiation, oxygen-to-metal (O/M) at beginning-of-life and at burnup, and the potential in fuel-cladding gap at burnup are described

  4. LMFBR design and its evolution. (2) Core design of LMFBR

    International Nuclear Information System (INIS)

    Uto, Nariaki; Mizuno, Tomoyasu

    2003-01-01

    Sodium-cooled core design studies are performed. MOX fuel core with axial blanket partial elimination subassembly due to safety consideration is studied. This type of core with high internal conversion ratio possesses capability of achieving 26 months of operation cycle length and 100 GWd/t of burnup averaged over core and blanket, which are superior characteristics in view of reducing cost of power generation. Metal fuel core is also studied, and its higher breeding capability reveals a potential of better core performance such as longer operation cycle length for the same level of electricity generation, though core outlet temperature is limited to lower level due to steel cladding-metal fuel compatibility concerns. Another metal fuel core concept using single Pu enrichment and two radial regions with individual fuel pin diameters achieves 550degC of core outlet temperature identical to that of MOX fuel core, keeping operation cycle length comparable with that of MOX fuel core. This series of study results show that sodium-cooled MOX and metal fuel cores have a high flexibility in satisfying various needs including fuel cycle cost and breeding capability, depending on the stage of introducing commercialized fast reactor cycle system. (author)

  5. Influence of fuel pin bowing on the temperature distribution in fuel pin cladding tubes in case of sodium cooling; experimental results

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.; Kolodziej, M.

    1978-09-01

    The influence of rod bowing on the local temperature distribution was measured with turbulent sodium flow in the cladding tubes of a 19-rod bundle mock-up of the SNR 300 Mark Ia fuel element. Such measurements have been carried out for the first time. The results presented in this report are part 1 of the experimental evaluation not yet completed. The major results are: 1. When a rod on the first ring gets deformed towards a neighbour on the second ring with a gap reduction from the nominal value of 100 % down to 20 %, the maximum azimuthal temperature difference of the outer rod increases by about 60 %. 2. The maximum azimuthal temperature difference of a rod on the first ring increases by a factor of 2, if it is approached by a neighbour on the same ring. 3. The reduction in cross section of a subchannel by rod bowing results only locally in distinct temperature rises, i.e. in the adjacent cladding tubes. Rods of the next but one row are no more subject to noticeable changes in temperature [de

  6. Design of FCI Experiments to Understand Fuel Out-Pin Phenomena in the SFR

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Park, Seong Dae [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook; Bang, In Cheol [Chungang Univ., Seoul (Korea, Republic of)

    2014-05-15

    It is important to guarantee a passive nuclear safety regarding enhanced negative reactivity by fragmenting the molten fuel. In the SFR, it has a strong point that the negative reactivity is immediately introduced when the metal fuel is melted by the UTOP or ULOP accident. These characteristics of the metal fuel can prevent from progressing in severe accidents such as core disruptive accidents (CDA). As key phenomena in the accidents, fuel-coolant interaction (FCI) phenomena have been studied over the last few decades. Especially, several previous researches focused on instability and fragmentation of a core melt jet in water. However, the studies showed too limited phenomena to fully understand. In the domestic SFR technology development, researches for severe accidents tend to lag behind ones of other countries. Or, South Korea has a very basic level of the research such as literature survey. Recently, the SAS4A code, which was developed at Argonne National Laboratory (ANL) for thermal-hydraulic and neutronic analyses of power and flow transients in liquid-metal-cooled nuclear reactors (LMRs), is still under development to consider for a metal fuel. The other countries carried out basic experiments for molten fuel and coolant interactions. However, in a high temperature condition, methods for analysis of structural interaction between molten fuel and fuel cladding are very limited. The ultimate objective of the study is to evaluate the possibility of recriticality accident induced by fuel-coolant interaction in the SFR adopting metal fuel. It is a key point to analyze the molten-fuel behavior based on the experimental results which show fuel-coolant interaction with the simulant materials. It is necessary to establish the test facility, to build database, and to develop physical models to understand the FCI phenomena in the SFR; molten fuel-coolant interaction as soon as the molten fuel is ejected to the sodium coolant channel and molten fuel-coolant interaction

  7. Results of experimental investigations for substantiation of WWER cermet fuel pin performance

    International Nuclear Information System (INIS)

    Popov, V.V.; Karpin, A.D.; Isupov, I.A.; Rumyantsev, V.N.; Troyanov, V.M.; Subonyaev, V.N.; Melnichenko, N.A.

    1997-01-01

    The out-of-pile experiment results on interaction of the cladding and matrix materials and uranium dioxide at cermet fuel temperature for normal operating conditions of the WWER-440 reactor are analyzed. Cermet fuel element behaviour under the maximum designed damage of the WWER-440 reactor is considered. In the AM reactor loop a fission product output from the unsealed cermet fuel elements have been studied. (author). 6 figs, 3 tabs

  8. IR1 flow tube and In-Pile Test Section Pressure drop test for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-15

    The in-pile Section (IPS) of 3-pin Fuel Test Loop(FTL) shall be installed in the vertical hole call IR1 of HANARO reactor core. In order to verify the pressure drop and flow rate both the inside region of IPS at the annular region between IPS and IR1 flow tube, a pressure drop was measured by varing the flow rate on both regions. The measured pressure drop in the annular region is 209kpa at 14.9kg/s which meets the limiting condition of operation of 200kpa. The measured pressure drop in side the IPS becomes 260.25kpa which is lower than the designed value of 306.65kpa. As the pressure drop is lower than the design value, it is quite conservative from the safety and operating point of view.

  9. The consequences of a sharp temperature change in the fuel pins of an accelerator-driven subcritical system

    International Nuclear Information System (INIS)

    Dagan, R.; Jianu, A.; Weisenburger, A.; Schikorr, M.; Rimpault, G.

    2013-01-01

    The effect of temperature changes and in particular those that are accompanied by strong gradients was extensively investigated for fast reactors. Subcritical systems designed for their transmutation ability are to some extent similar to critical power reactors in their subassembly structure. However, they differ in two main aspects. First, the coolant in a subcritical system is lead or lead-bismuth eutectic (LBE) and not sodium, and second, the main cause for steep temperature gradients in a fast power reactor is sudden control rod insertion, or scram, whereas in subcritical systems shutdown of the accelerator and its proton beam is the main cause for temperature gradients. Furthermore, the increased probability of operational interruptions in an accelerator driven system is largely due to the instability of the accelerator generating the proton beam. This study uses the knowledge gained from fast reactors as a preliminary reference and concentrates further on the unique features of the proposed subcritical systems. In particular, the effect of beam trips on the fuel pin integrity is evaluated as a function of the temperature gradients and the duration of the beam trips. It seems, however, that the largest hazard to the fuel pin integrity is due to the lead (or LBE) coolant. In particular, the stability of the protective oxide layer built on the clad surface with the lead coolant appears quite sensitive to sudden temperature changes. In the second part of this study, several available experimental results show that even very moderate temperature changes are sufficient to cause crack formation in the oxide layer thereby exposing the clad surface to enhanced LBE corrosion. In the worst case, complete exfoliation of the magnetite outer layer is observed. As a consequence, clad failure probability due to corrosion is considerably increased. (authors)

  10. LMFBR plant parameters

    International Nuclear Information System (INIS)

    1985-07-01

    This document has been prepared on the basis of information compiled by the members of the IAEA International Working Group on Fast Reactors (IWGFR). It contains parameters of 25 experimental, prototype and commercial size liquid metal fast breeder reactors (LMFBR). Most of the reactors are currently in operation, under construction or in an advanced planning stage. Parameters of the Clinch River Breeder Reactor (USA) are presented because its design was nearly finished and most of the components were fabricated at the time when the project was terminated. Three reactors (RAPSODIE (France), DFR (UK) and EFFBR (USA)) have been shut down. However, they are included in the report because of their important role in the development of LMFBR technology from first LMFBRs to the prototype size fast reactors. The first LMFBRs (CLEMENTINE (USA), EBR-1 (USA), BR-2 (USSR), BR-5 (USSR)) and very special reactors (LAMPRE (USA), SEFOR (USA)) were not recommended by the members of the IWGFR to be included in the report

  11. Probabilistic analysis of fuel pin behaviour during an eventual loss of coolant in PWR reactors

    International Nuclear Information System (INIS)

    1981-02-01

    Brief description of the development of the coolant loss incident in a pressurized water reactor and analysis of its significance for the behaviour of the fuel rods. Description of a probalistic method for estimating the effects of the accident on the fuel rods and results obtained [fr

  12. Characterization of velocity and temperature fields in a 217 pin wire wrapped fuel bundle of sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Naveen Raj, M.; Velusamy, K.

    2016-01-01

    Highlights: • We simulate flow and temperature fields in fuel subassembly of fast reactor. • We perform high fidelity computations for 217 pin bundle of 7 axial pitch lengths. • We investigate transverse and axial flows in different types of subchannels. • Correlations are proposed for transverse flow, which form input for subchannel analysis. • Periodic variations of large magnitude are observed in subchannel flow rates. - Abstract: RANS based computational fluid dynamic (CFD) simulation of flow and temperature fields in a fast reactor fuel subassembly has been carried out. The sodium cooled prototype subassembly consists of 217 pins with helical wire spacers. An axial length of seven helical wire pitches has been considered for the study adopting a structured mesh having 36 million points and 84 processors in parallel. The computational model has been validated against in-house and published experimental data for friction factor and Nusselt number. Also, the transverse flow in the central subchannel and swirl flow in the peripheral subchannel are compared against reported experimental data and those computed by subchannel models. The focus of the study is investigation of transverse and axial flows in different types of subchannels. Based on the 3-dimensional CFD study, correlations have been proposed for calculation of transverse flow, which forms an important input for development of subchannel analysis codes. Periodic variations have been observed in the subchannel axial flow rates. For the subchannels located in the central region, the peak to peak variation in the axial flow rate is ∼21% and it is found to be contributed by the changes in the flow area and hydraulic resistance due to frequent passage of helical wires through the subchannel. For the subchannels located in the periphery, this variation is as high as 50%. The transverse flow in the central subchannels follows a cosine profile, for all the faces. However, there is a phase lag of 120

  13. Fuel and fuel pin behaviour in a high burnup fast breeder fuel subassembly: Results of destructive post-irradiation examinations of the KNK II/1 fuel subassembly NY-205

    International Nuclear Information System (INIS)

    Patzer, G.

    1991-05-01

    The report gives a summarizing overview of the design characteristics, of the irradiation history and of the results of the destructive post-irradiation examinations of the fuel pins of the high-burnup fuel subassembly NY-205 of the KNK II first core. This element was operated for about 10 years and reached a maximum local burnup of 175 MWd/kg(HM) and a maximum neutron dose of 67 dpa-NRT. The main design data of this subassembly agree with those of the SNR 300 Mark-Ia, and it reached more than twice of the burnup and a similar neutron dose as foreseen for the SNR 300 fuel subassemblies [de

  14. Out-of-reactor experimental study of fuel-pin failure phenomena

    International Nuclear Information System (INIS)

    Wrona, B.J.; Galvin, T.M.; Stahl, D.

    1976-01-01

    Fundamental experiments have been performed with a direct-electrical-heating apparatus, on both unclad and quartz-clad UO 2 pellet stacks, to study the effect of a radial constraint on solid and molten-fuel motion during power transients. Results of simulated transient over-power experiments show that molten UO 2 can be quite mobile when the fuel centerline temperature exceeds the boiling point, i.e., fuel vapor pressures become a significant driving force for relocating molten fuel. For radially constrained pellet stacks, when an escape path was provided around the top pellet, significant upward axial fuel motion occurred prior to cladding rupture. Thus, the time sequence of events shows that potential exists for providing a negative reactivity-feedback effect, which would promote nuclear reactor safety. The data tend to support the existence of a ''pressurized-bottle'' effect, which was observed in high-speed movies

  15. Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

    International Nuclear Information System (INIS)

    Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

    1981-04-01

    Three fuel pin bundles, R-109/1, 2 and 3, were irradiated in a PWR loop in the HFR at Petten during respectively 131, 57 and 57 effective full power days at average powers of approximately 39 kW.m -1 and at peak powers of approximately 60 kW.m -1 . The results of the post-irradiation examinations of these fuel bundles are presented. (Auth.)

  16. Tank type LMFBR type reactors

    International Nuclear Information System (INIS)

    Shimizu, Hiroshi

    1985-01-01

    Purpose: To detect the abnormality in the suspended body or reactor core supporting structures thereby improve the safety and reliability of tank type LMFBR reactors. Constitution: Upon inspection during reactor operation period, the top end of the gripper sensing rod of a fuel exchanger is abutted against a supporting bed and the position of the reactor core supporting structures from the roof slab is measured by a stroke measuring device. Then, the sensing rod is pulled upwardly to abut against the arm portion and the position is measured by the stroke measuring device. The measuring procedures are carried out for all of the sensing rods and the measured values are compared with a previously determined value at the initial stage of the reactor operation. As a result, it is possible to detect excess distortions and abnormal deformation in the suspended body or reactor core supporting structures. Furthermore, integrity of the suspended body against thermal stresses can be secured by always measuring the coolant liquid level by the level measuring sensor. (Kamimura, M.)

  17. 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

  18. 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)

  19. Effect of operating temperature on LMFBR core performance

    International Nuclear Information System (INIS)

    Noyes, R.C.; Bergeron, R.J.; di Lauro, G.F.; Kulwich, M.R.; Stuteville, D.W.

    1977-01-01

    The purpose of the study is to provide an engineering evaluation of high and low temperature LMFBR core designs. The study was conducted by C-E supported by HEDL expertise in the areas of materials behavior, fuel performance and fabrication/fuel cycle cost. The evaluation is based primarily on designs and analyses prepared by AI, GE and WARD during Phase I of the PLBR studies

  20. 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.)

  1. Application of different failure criteria in fuel pin modelling and consequences for overpower transients in LMFBRs

    International Nuclear Information System (INIS)

    Kuczera, B.; Royl, P.

    1975-01-01

    The CAPRI-2 code system for analysis of hypothetical core disruptive accidents in LMFBRs has recently been coupled with the transient deformation model BREDA-2. The new code system determines thermal and mechanical loads under transient conditions for both, fresh and irradiated fuel and cladding, taking into account fuel restructuring as well as effects from fission gas and fuel and clad swelling. The system has been used for analysis of mild uncontrolled overpower transients in the SNR-300 to predict failure, and to initialize and calculate subsequent fuel coolant interaction (FCI). Thirteen channels have been coupled by point kinetics for the whole core analysis. Three different failure mechanisms and their influence on accident sequence have been investigated: clad melt-through; clad burst caused by internal pressure build-up; clad straining due to differential thermal expansion between fuel and clad cylinders. The results of these analyses show that each failure mechanism will lead to rather different failure and accident sequences. There is still a lack of experimental data from which failure thresholds can be derived. To get better predictions from the applied models an improved understanding of fission release and its relation to fuel porosity also some better experimental data on fluence and temperature dependent rupture strains of the cladding material should be available

  2. Gas release from a failed fuel pin after reactor shut-down

    International Nuclear Information System (INIS)

    Pshenichnikov, B.V.

    1975-01-01

    A mathematical model of gassing from a hypothetical core fuel element in the active zone of a stopped water-moderated reactor was analysed to investigate the process of liberation of gaseous fission products from an unpressurized fuel element. A one-dimensional problem was obtained as a result of the accepted hypotheses. A fault was assumed to have occured during reactor operation; at the same time, a vapour-gas mixture was considered to be present under the envelope at reactor working pressure by the moment of stoppage. An approximative estimation was made of the retardation time of pressure balancing at the open end of the fuel element, and also of the amount of total gas remaining in the gap under the fuel element envelope after pressure drop in the reactor. Estimation of retardation time permitted to conclude that pressure in the nonhermetic fuel element envelope follows pressure fluctuation in the reactor in the course of cooling, the retardation time of pressure balancing outside and inside the fuel element lasting but a few seconds

  3. Sensitivity Study of the Peak Cladding Temperature for the Pipe Break Accidents of the 3-Pin Fuel Test Loop

    International Nuclear Information System (INIS)

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

    2005-12-01

    The effect of the thermal hydraulic operation parameters, the stroke times of safety-related valves, the node number of test fuel for MARS modeling, and the axial power distribution on the peak cladding temperature (PCT) has been investigated for the loss of coolant accident of the 3-pin fuel test loop. The thermal hydraulic operation parameters investigated are the thermal power of the fuel test loop and the flow rate, temperature, and pressure of the main cooling water. The effect of the thermal power and the coolant temperature on the peak cladding temperature is dominant as compared with that of the coolant flow rate and pressure. The maximum PCT increases up to about 34.3K for the room 1 LOCA when the thermal power increase by 5% of the normal operation power and decreases up to about 38.9K for the room 1 LOCA when the coolant temperature decrease by 2% of the normal operation temperature. The effect of the stroke time of the loop isolation valves on the PCT is also dominant. However the effect of the stroke time of the safety injection valves and depressurization vent valves are negligible. Especially the maximum PCT increases up to 25.7K with the increase of the design stroke time of the cold leg loop isolation valve by 13% and decreases up to 25.1K with the decrease of the design stroke time by 13%. The maximum PCT increases by 3.3K as the number of nodes increases from 7 to 14 for the MARS model of test fuel. Three different axial power distributions are also investigated. The maximum PCT occurs for the room 1 LOCA in case the peak power is shifted to the downstream by 20cm

  4. Analysis of a postulated accident scenario involving loss of forced flow in a LMFBR

    International Nuclear Information System (INIS)

    Moreira, M.L.

    1985-01-01

    A model to analyse a postulated accident scenario involving loss of forced flow in the reactor vessel of a LMFBR is used. Five phases of the accident are analysed: Natural Circulation, Subcooled Boiling, Nucleate Boiling, Core Dryout and Cladding melt. The heat conduction in the fuel cladding, coolant and lower and upper plenum are calculated by a lump-parameter model. Physical data of a prototype LMFBR reactor were used for the calculation. (author)

  5. 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

  6. Temperature measurement on Zircaloy-clad fuel pins during high temperature excursions

    International Nuclear Information System (INIS)

    Meservey, R.H.

    1976-04-01

    The development of a sheathed thermocouple suitable for attachment to zircaloy-clad fuel rods and for use during high temperature (2,800 0 F) excursions under loss-of-coolant accident conditions is described. Development, fabrication, and testing of the thermocouples is covered in detail. In addition, the development of a process for laser welding the thermocouples to fuel rods is discussed. The thermocouples and attachment welds have been tested for resistance to corrosion and nuclear radiation and have been subjected to fast thermal cycle, risetime, and blowdown accident tests

  7. Microcontroller based automation system for end plug welding of test fuel pins in solgel facility

    International Nuclear Information System (INIS)

    Prabhakar Rao, J.; Srinivas, K.C.; Prabhu, T.V.; Ravi, N.

    2010-01-01

    A microcontroller based stepper motor control and driver Unit for 'XY' positioning system is designed and developed to perform the 'pick-place' of fuel tube to pre-determined coordinates. This Unit provides a fine movement of the fuel tube to get perfect position for welding. The Graphical User Interface software running on PC displays the absolute position of the XY system and provides all the required control buttons to achieve the accurate positioning. The welding of clad tube with end plug is carried out in a high precision welding fixture by operating it remotely. This paper discusses about the Hardware and Software features and implementation of the instrumentation. (author)

  8. Application of a statistical methodology for the comprehension of corrosion phenomena on Phenix spent fuel pins

    International Nuclear Information System (INIS)

    Pantera, L.

    1992-11-01

    The maximum burnup of Phenix fuel elements is strongly conditioned by the internal corrosion of the steel cladding. This thesis is a part of a new study program on the corrosion phenomena. Based on the results of an experimental program during the years 1980-1990 its objective is the use of a statistical methodology for a better comprehension of the corrosion phenomena

  9. 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

  10. Summary and evaluation: fuel dynamics loss-of-flow experiments (tests L2, L3, and L4)

    International Nuclear Information System (INIS)

    Barts, E.W.; Deitrich, L.W.; Eberhart, J.G.; Fischer, A.K.; Meek, C.C.

    1975-09-01

    Three similar experiments conducted to support the analyses of hypothetical LMFBR unprotected-loss-of-flow accidents are summarized and evaluated. The tests, designated L2, L3, and L4, provided experimental data against which accident-analysis codes could be compared, so as to guide further analysis and modeling of the initiating phases of the hypothetical accident. The tests were conducted using seven-pin bundles of mixed-oxide fuel pins in Mark-II flowing-sodium loops in the TREAT reactor. Test L2 used fresh fuel. Tests L3 and L4 used irradiated fuel pins having, respectively, ''intermediate-power'' (no central void) and ''high-power'' (fully developed central void) microstructure. 12 references

  11. A model for predicting the radial power profile in a fuel pin

    International Nuclear Information System (INIS)

    Palmer, I.D.; Hesketh, K.W.; Jackson, P.A.

    1983-01-01

    A simple, fast running computer program for calculating radial power profiles, throughout life, in both standard and duplex fuel pellets for all types of thermal reactor has been developed. The code sub-divides the pellet into a number of annuli for each of which it solves for the concentrations of uranium and plutonium and hence calculates a mean inverse diffusion length. The diffusion equation is solved in terms of Bessel functions and the resulting flux profile multiplied by the concentration profiles to give a radial rating profile which is normalised to unity. The model shows good agreement with the results of detailed physics calculations for different thermal reactors over a wide burn-up range. Its incorporation into the HOTROD-4C and SLEUTH-SEER-77 fuel performance codes has led to a negligible increase in running times. (author)

  12. 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

  13. 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.

  14. Status of U.S. LMFBR programme

    International Nuclear Information System (INIS)

    Yevich, J.

    1978-01-01

    The determents of the decision for deterrence of commercial reprocessing and further demonstration of the plutonium breeder were based on two premises: time is needed to establish the programme for non-proliferating fuel cycle and there is a lessened sense of urgency for the USA to establish a commercial breeder in the near future. A strong, well funded base technology effort remains and will continue until institutional and technical solutions can be found to minimize or eliminate the proliferation risk. An LMFBR option will be maintained. The FFTF will be coming on line providing a powerful tool in breeder fuel and materials development and a baseline from which to scale up heat transfer systems and components. Sodium system hardware development and testing will continue to have high priority

  15. Development of a CVD silica coating for UK advanced gas-cooled nuclear reactor fuel pins

    International Nuclear Information System (INIS)

    Bennett, M.J.; Houlton, M.R.; Moore, D.A.; Foster, A.I.; Swidzinski, M.A.M.

    1983-04-01

    Vapour deposited silica coatings could extend the life of the 20% Cr/25% Ni niobium stabilised (20/25/Nb) stainless steel fuel cladding of the UK advanced gas cooled reactors. A CVD coating process developed originally to be undertaken at atmospheric pressure has now been adapted for operation at reduced pressure. Trials on the LP CVD process have been pursued to the production scale using commercial equipment. The effectiveness of the LP CVD silica coatings in providing protection to 20/25/Nb steel surfaces against oxidation and carbonaceous deposition has been evaluated. (author)

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

    International Nuclear Information System (INIS)

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

    1981-04-01

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

  17. A study of solute transport of radiolysis products in crud and its effects on crud growth on PWR fuel pin

    Energy Technology Data Exchange (ETDEWEB)

    Joe, Justin H. [BNF Consulting (United States); Kim, Seung Jun, E-mail: skim@lanl.gov [Mechanical and Thermal Engineering Group (AET-1), Los Alamos National Laboratory (United States); Jones, Barclay G. [Department of Nuclear Plasma Radiological Engineering, University of Illinois Urbana-Champaign (United States)

    2016-04-15

    Highlights: • We model a 3-D numerical solute transport within crud deposit on PWR fuel pin. • Source term effect from radiolysis yield and recombination is minimal. • Lower crud porosity leads substantially higher concentration of solutes. • Thicker crud deposit generates substantially higher concentration of solutes. • High concentration of radiolysis species (H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2}) can be directly related to corrosion issues on fuel cladding. - Abstract: This research examines the concentration of radiolysis species (H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2}) over the porous crud layer using a three dimensional time dependent solute transport model. A Monte Carlo random walk technique is adopted to simulate the transport behavior of the different species with various parametric studies of source term, crud thickness, and crud porosity. Particularly, this model employs a system of coupled mass transport and chemical interactions as the source term, which makes the problem non-linear. It is demonstrated that a negligible effect on radiolysis species concentrations change due to the consideration of source term. The crud thickness and porosity effect on the concentration distributions are notably observed. In general, higher concentration starts from the intersection of the heating surface with the chimney wall from the beginning and it reaches the equilibrium state within tens of seconds. The concentration profiles of the radiolysis species H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2} can be directly related to corrosion issues. The direct application of this study to nuclear engineering research is to aid in the design of reactors with higher performance without experiencing an Axial Offset Anomaly (AOA), an unexpected measured shift in axial power distribution from predicted values.

  18. A study of solute transport of radiolysis products in crud and its effects on crud growth on PWR fuel pin

    International Nuclear Information System (INIS)

    Joe, Justin H.; Kim, Seung Jun; Jones, Barclay G.

    2016-01-01

    Highlights: • We model a 3-D numerical solute transport within crud deposit on PWR fuel pin. • Source term effect from radiolysis yield and recombination is minimal. • Lower crud porosity leads substantially higher concentration of solutes. • Thicker crud deposit generates substantially higher concentration of solutes. • High concentration of radiolysis species (H 2 , O 2 , and H 2 O 2 ) can be directly related to corrosion issues on fuel cladding. - Abstract: This research examines the concentration of radiolysis species (H 2 , O 2 , and H 2 O 2 ) over the porous crud layer using a three dimensional time dependent solute transport model. A Monte Carlo random walk technique is adopted to simulate the transport behavior of the different species with various parametric studies of source term, crud thickness, and crud porosity. Particularly, this model employs a system of coupled mass transport and chemical interactions as the source term, which makes the problem non-linear. It is demonstrated that a negligible effect on radiolysis species concentrations change due to the consideration of source term. The crud thickness and porosity effect on the concentration distributions are notably observed. In general, higher concentration starts from the intersection of the heating surface with the chimney wall from the beginning and it reaches the equilibrium state within tens of seconds. The concentration profiles of the radiolysis species H 2 , O 2 , and H 2 O 2 can be directly related to corrosion issues. The direct application of this study to nuclear engineering research is to aid in the design of reactors with higher performance without experiencing an Axial Offset Anomaly (AOA), an unexpected measured shift in axial power distribution from predicted values.

  19. Comparative analysis of a hypothetical 0.1 $/SEC transient overpower accident in an irradiated LMFBR core using different computer models

    International Nuclear Information System (INIS)

    Cacciabue, P.C.; Fremont, R. de; Renard, A.

    1982-01-01

    The Report gives the results of comparative calculations performed by the Whole Core Accident Codes Group which is a subgroup of the Safety Working Group of the Fast Reactor Coordinating Committee for a hypothetical transient overpower accident in an irradiated LMFBR core. Different computer codes from members of the European Community and the United States were used. The calculations are based on a Benchmark problem, using commonly agreed input data for the most important phenomena, such as the fuel pin failure threshold, FCl parameters, etc. Beside this, results with alternative assumptions for theoretical modelling are presented with the scope to show in a parametric way the influence of more advanced modelling capabilities and/or better (so-called best estimate) input data for the most important phenomena on the accident sequences

  20. LMFBR plant parameters 1991

    International Nuclear Information System (INIS)

    1991-03-01

    The document has been prepared on the basis of information provided by the members of the IAEA International Working Group on Fast Reactors (IWGFR). It contains updated parameters of 27 experimental, prototype and commercial size liquid metal fast breeder reactors (LMFBRs). Most of the reactors are currently in operation, under construction or in an advanced planning stage. Parameters of the Clinch River Breeder Reactor (USA), PEC (Italy), RAPSODIE (France), DFR (UK) and EFFBR (USA) are included in the report because of their important role in the development of LMFBR technology from first LMFBRs to the prototype size fast reactors. Two more reactors appeared in the list: European Fast Reactor (EFR) and PRISM (USA). Parameters of these reactors included in this publication are based on the data from the papers presented at the 23rd Annual Meeting of the IWGFR. All in all more than four hundred corrections and additions have been made to update the document. The report is intended for specialists and institutions in industrialized and developing countries who are responsible for the design and operation of liquid metal fast breeder reactors

  1. Mechanical interaction between fuel pins and assemblies during LOCA in BWR

    International Nuclear Information System (INIS)

    Jonsson, T.

    1978-10-01

    The size of the rod elongation by oxidation is so large that deformation of a standard BWR fuel element with tie rods in the outer row will surely occur during a LOCA transient typical for BWRs with external pumps. Available data does not however show whether this deformation will occur early in the transient or during the cooling. Combined effects of thermal expansion of zircaloy and expansion due to oxidation and dissolution of oxygen can be expected to be large enough to cause rod bowing early in a LOCA transient. It is however not impossible that observed residual expansion of zircaloy tubes to a dominating extent are caused through expansion of zirconium oxide during cool-down. Length measurements of zircaloy tubes during a transient are desirable. (author)

  2. EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

    Energy Technology Data Exchange (ETDEWEB)

    Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billone, Mike [Argonne National Lab. (ANL), Argonne, IL (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.

  3. Analytical work on local faults in LMFBR subassembly

    International Nuclear Information System (INIS)

    Yoshikawa, H.; Miyaguchi, K.; Hirata, N.; Kasahara, F.

    1979-01-01

    Analytical codes have been developed for evaluating various severe but highly unlikely events of local faults in the LMFBR subassembly (S/A). These include: (1) local flow blockage, (2) two-phase thermohydraulics under fission gas release, and (3) inter-S/A failure propagation. A simple inter-S/A thermal failure propagation analysis code, FUMES, is described that allows an easy parametric study of propagation potential of fuel fog in a S/A. 7 refs

  4. LMFBR technology. FFTF cover-gas leakage calculation

    International Nuclear Information System (INIS)

    Deboi, H.

    1974-01-01

    The FFTF LMFBR is intended to have a near zero release of radioactive gases during normal reactor operation with 1% failed fuel. This report presents calculations which provide an approximation of these cover gas leakages. Data from ongoing static and dynamic seal leak tests at AI are utilized. Leakage through both elastomeric and metallic seals in all sub-assemblies and penetrations comprising the reactor cover gas containment during reactor operation system are included

  5. Fluid-mixing studies in a hexagonal 61-pin wire-wrapped rod bundle

    International Nuclear Information System (INIS)

    Symolon, P.D.; Todreas, N.E.

    1981-02-01

    Mixing, pressure drop, and flow split experiments were performed on a 61 pin LMFBR fuel bundle with a pitch to diameter ratio of 1.25, and lead lengths of 6 and 12 inches. The mixing results obtained from salt injection experiments were found to depend on injection depth. The deeper injection is expected to give the more accurate results. The pressure drop data was presented as friction factor versus Reynolds number, and the results were compared to the correlation of Hawley. The flowsplit data presented was flawed by corroded bundle walls, but some insight was obtained on the effect of rough surfaces on flowsplit, and how to account for its effect in the correlations

  6. Inner wall attack and its inhibition method for FBR fuel pin cladding at high burnup

    International Nuclear Information System (INIS)

    Xu Yongli; Long Bin; Li Jingang; Wan Jiaying

    1998-01-01

    The inner wall attack of the modified 316-Ti S.S. cladding tubes manufactured in China used FBR at 10at.% burnup was investigated by means of the out of pile simulation tests. The inner surface morphologies of the cladding tubes attached by fission products Cs, Te, I and Se at 700 deg. C under lower and high oxygen potentials were observed respectively, and the depth of attack was also measured. The burst strength, maximum circum expansion and the appearances of fracture were measured and observed respectively for the cladding tubes attacked by fission products. Based on the mechanism of FBR fuel cladding chemical interaction (FCCI), Cr, Zr and Nb were used as the oxygen absorbers respectively, in order to inhibit the inner wall attack of the cladding tubes. The corrosion morphologies and depth, the penetration depth of the fission products in the inner surface of the cladding tubes were detected. The inhibition effectiveness of the oxygen absorbers for the inner wall attack of the cladding tubes was evaluated. (author)

  7. LMFBR subassembly response to simulated local pressure loadings

    International Nuclear Information System (INIS)

    Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

    1976-01-01

    The structural response of liquid metal fast breeder reactor (LMFBR) subassemblies to local accidental events is of interest in assessing the safety of such systems. Problems to be resolved include failure propagation modes from pin to pin and from subassembly to subassembly. Factors which must be considered include: (a) the geometry of the structure, (b) uncertainty of the pressure-energy source, (c) uncertainty of materials properties under reactor operating conditions, and (d) the difficulty in performing in-pile or out-of-pile experiments which would simulate the above conditions. The main effort in evaluating the subassembly response has been centered around the development of appropriate analyses based on the finite element technique. Analysis has been extended to include not only the subassembly duct structure itself, but also the fluid environment, both within subassemblies and between them. These models and codes have been devised to cover a wide range of accident loading conditions, and can treat various materials as their properties become known. The effort described here is centered mainly around an experimental effort aimed at verfying, modifying or extending the models used in treating subassembly damage propagation. To verify the finite element codes under development, a series of out-of-pile room temperature experiments has been performed on LMFBR-type subassembly ducts under various loading conditions. (Auth.)

  8. Methods used and kind of results obtained in SCARABEE facility about fuel and clad motion diagnostics. Previsional development for SCARABEE N and CABRI facilities

    International Nuclear Information System (INIS)

    Bardy, J.; Manent, G.; Roche, L.; Tattegrain, A.

    1975-01-01

    Techniques are described for planned LMFBR loss of flow simulation studies using irradiated fuel (SCARABEE N program) and for LMFBR transient overpower studies using both fresh and irradiated fuel (CABRI program)

  9. CAMDYN: a new model to describe the axial motion of molten fuel inside the pin of a fast breeder reactor during accident conditions

    International Nuclear Information System (INIS)

    Peter, G.

    1991-01-01

    The new in-pin fuel motion model CAMDYN (Cavity Material Dynamics) describes the axial motion of both partially and fully molten fuel inside the pin of a fast breeder reactor during accident conditions. The motion of the two types of molten fuel and the imbedded fission gas bubbles is treated both before and after cladding failure. The basic modelling approach consists of the treatment of two one-dimensional flows which are coupled by interaction terms. Each of these flows is treated compressively and with axially variable flow cross sections. The mass and energy equations of both fields are solved explicitly using upwind differencing on a fixed Eulerian grid. The two momentum equations are solved simultaneously, using the convective momentum fluxes of the previous timestep. Both partially and fully molten fuel can move axially into a central hole extending to the plenum in the case of certain hollow pellet designs. The fuel temperature calculation includes the determination of a radial temperature profile. A simple conduction freezing model is included. After cladding failure, ejection into the coolant channel is modeled

  10. LMFBR models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.

    1981-10-01

    Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th- 238 U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given

  11. 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)

  12. Analysis of three idealized reactor configurations: plate, pin, and homogeneous

    International Nuclear Information System (INIS)

    McKnight, R.D.

    1983-01-01

    Detailed Monte Carlo calculations have been performed for three distinct configurations of an idealized fast critical assembly. This idealized assembly was based on the LMFBR benchmark critical assembly ZPR-6/7. In the first configuration, the entire core was loaded with the plate unit cell of ZPR-6/7. In the second configuration, the entire core was loaded with the ZPR sodium-filled pin calandria. The actual ZPR pin calandria are loaded with mixed (U,Pu) oxide pins which closely match the composition of the ZPR-6/7 plate unit cell. For the present study, slight adjustments were made in the atom concentrations and the length of the pin calandria in order to make the core boundaries and average composition for the pin-cell configuration identical to those of the plate-cell configuration. In the third configuration, the core was homogeneous, again with identical core boundaries and average composition as the plate and pin configurations

  13. LMFBR safety program. Annual technical progress report. Government fiscal year, 1977

    International Nuclear Information System (INIS)

    1977-01-01

    Information is presented concerning the development of the SOMIX-1 computer code for sodium drop burning analysis; experimental analysis of burning sodium drops; aerosol leakage from containment buildings; high-temperature-concentration aerosols; aerosol source term from vaporized fuel; properties of high-temperature fuel mixtures; and development of the COMRADEX computer code for analysis of radiological doses in the environment from LMFBR accidents

  14. Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

    International Nuclear Information System (INIS)

    Henderson, J.M.; Wood, S.A.; Knight, D.D.

    1981-01-01

    The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting

  15. Pin care

    Science.gov (United States)

    ... Drugs & Supplements Videos & Tools Español You Are Here: Home → Medical Encyclopedia → Pin care URL of this page: //medlineplus.gov/ency/patientinstructions/000481.htm Pin care To use the sharing features on this page, please enable JavaScript. Broken bones can be fixed in surgery with metal ...

  16. Coupling analysis of deformation and thermal-hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV Codes

    International Nuclear Information System (INIS)

    Ito, Masahiro; Imai, Yasutomo; Uwaba, Tomoyuki; Ohshima, Hiroyuki

    2004-03-01

    The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)

  17. Some observations on pitting corrosion in the zircaloy cladding of fuel pins irradiated in a PWR loop

    International Nuclear Information System (INIS)

    Linde, A. van der; Letsch, A.C.; Hornsveld, E.M.

    1978-11-01

    A three-pins, zircaloy-4 clad, sphere-pac bundle was irradiated in a 280 0 C PWR loop in the HFR at Petten during 131 effective full power days to a bundle average burnup of 0.84 % FIMA. The pins contained a mixture of 61.5 w/o of 1050 μm (U,Pu) 0 2 spheres, 18.5 w/o of 115 μm UO 2 spheres and 20.0 w/o of 2 spheres. The as-fabricated smear density of the vibratory compacted mixture was 81-85 % T.D. The pressure of the pin filling gas was 1 bar helium for pin 306 and 25 bar helium for the pins 308 and 309. The cladding was zircaloy-4 tubing, stress relieved for 4 hours at 540 0 C, with an inner diameter of 9.30 mm and a wall thickness of 0.73 mm. Exposure of the pins in the loop started in the as-pickled, degreased surface condition. The pins operated at an average heat rating of 335 W/cm and at a peak rating of 620 W/cm. The end-of -life peak rating was 425 W/cm. Unfavourable water chemistry conditions of the coolant during the last weeks of the irradition, in particular low NH 3 concentrations resulting in low pH values, caused the deposition of heavy crud layers on the pin surfaces. This crud layer caused a small cladding defect in pin 306 at the axial position of the peak heat rating. The zircaloy-4 wall failed by complete oxidation, which started at and progressed from the outer, coolant side, surface. Immediately after the detection of fission product activity in the loop water, the irradiation of the bundle was terminated. Microscopic investigations on cross sections of the pins 306 and 309 revealed the presence of oxide pits at the outer surface of the zircalloy-4 wall

  18. 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

  19. The effect of core design changes on the doubling time and the fuel cycle cost of a 1,000 MWe LMFBR

    International Nuclear Information System (INIS)

    Otake, I.; Inoue, T.; Tomabechi, K.; Osada, H.; Aoki, K.

    1978-01-01

    Core design studies were performed to improve the doubling time and to minimize the fuel cycle cost of a 1,000 MWe Fast Demonstration Reactor. A core was designed mainly based on the technology being used for the design of a prototype fast reactor MONJU, because much valuable experience will be forthcoming from this reactor. Design parameters with a wide variable range were used to clarify the relations between breeding characteristics, fuel economics and various designs. (author)

  20. Investigation of the physical and numerical foundations of two-fluid representation of sodium boiling with applications to LMFBR experiments

    International Nuclear Information System (INIS)

    No, H.C.; Kazimi, M.S.

    1983-03-01

    This work involves the development of physical models for the constitutive relations of a two-fluid, three-dimensional sodium boiling code, THERMIT-6S. The code is equipped with a fluid conduction model, a fuel pin model, and a subassembly wall model suitable for stimulating LMFBR transient events. Mathematically rigorous derivations of time-volume averaged conservation equations are used to establish the differential equations of THERMIT-6S. These equations are then discretized in a manner identical to the original THERMIT code. A virtual mass term is incorporated in THERMIT-6S to solve the ill-posed problem. Based on a simplified flow regime, namely cocurrent annular flow, constitutive relations for two-phase flow of sodium are derived. The wall heat transfer coefficient is based on momentum-heat transfer analogy and a logarithmic law for liquid film velocity distribution. A broad literature review is given for two-phase friction factors. It is concluded that entrainment can account for some of the discrepancies in the literature. Mass and energy exchanges are modelled by generalization of the turbulent flux concept. Interfacial drag coefficients are derived for annular flows with entrainment. Code assessment is performed by simulating three experiments for low flow-high power accidents and one experiment for low flow/low power accidents in the LMFBR. While the numerical results for pre-dryout are in good agreement with the data, those for post-dryout reveal the need for improvement of the physical models. The benefits of two-dimensional non-equilibrium representation of sodium boiling are studied

  1. Post irradiation examination and analysis of 13(U,Pu) C-fuel pins irradiated in the thermal flux of FR 2

    International Nuclear Information System (INIS)

    Weimar, P.; Steiner, H.

    1979-01-01

    The post-irradiation examination of the pins at Karlsruhe Hot Cells revealed the following results: Nearly all specimens showed noteworthy clad deformations (up to 3%). Defects in the form of cracks in the clad were found at three pins. The observed clad deformations resulted from mechanical interaction between fuel and cladding in consequence of an inexorable fuel swelling. A linear relationship between burnup and clad deformation was found. Defects were observed for burnups greater than 50 MWd/kgM and can be explained by the small fabrications clearances between clad and fuel pellets (50-90 μm) and high smear densities. Fission gas measurements were performed in a three fold way, gas release, gas trapped in pores and gas in solid solution in the lattice of the mixed carbide were determined. The gas release fraction showed values between 10 and 15%. Whereas the fission gas content trapped in large pores (> 1 μm) was linearly dependent on burnup, fission gas in small pores and in solid solution reached a saturation value at about 20 MWd/kgM. Measurements of micro-hardness revealed carburization depths of the clad of up to 40% at temperatures of about 650 0 C. Furtermore, it could be confirmed that the carburization depth followed an Arrhenius law. (orig.)

  2. Mechanisms of fuel-cladding chemical interaction: US interpretation

    International Nuclear Information System (INIS)

    Adamson, M.G.

    1977-01-01

    Proposed mechanisms of fuel-cladding chemical interaction (FCCI) in LMFBR fuel pins are reviewed and examined in terms of in-pile and out-of-pile data. From this examination several factors are identified which may govern the occurrence of localized deep intergranular penetrations of Type-316SS cladding. Using a plausible mechanistic hypothesis for FCCI, first steps have been taken towards developing a quantitative, physically-meaningful, mathematical method of predicting cladding wastage in operating fuel pins. Both kinetic and thermodynamic aspects of FCCI are considered in the development of this prediction method, together with a fuel chemistry model that describes the evolution of thermochemical conditions at the fuel-cladding gap. On the basis of results from recent fuel pin and laboratory tests a thermal transport mechanism has been proposed to explain the thermal gradient-induced migration of Fe, Cr, and Ni from cladding into the fuel. This mechanism involves chemical transport of the metallic cladding components (as tellurides) in liquid Cs-Te. (author)

  3. Mechanisms of fuel-cladding chemical interaction: US interpretation

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States)

    1977-04-01

    Proposed mechanisms of fuel-cladding chemical interaction (FCCI) in LMFBR fuel pins are reviewed and examined in terms of in-pile and out-of-pile data. From this examination several factors are identified which may govern the occurrence of localized deep intergranular penetrations of Type-316SS cladding. Using a plausible mechanistic hypothesis for FCCI, first steps have been taken towards developing a quantitative, physically-meaningful, mathematical method of predicting cladding wastage in operating fuel pins. Both kinetic and thermodynamic aspects of FCCI are considered in the development of this prediction method, together with a fuel chemistry model that describes the evolution of thermochemical conditions at the fuel-cladding gap. On the basis of results from recent fuel pin and laboratory tests a thermal transport mechanism has been proposed to explain the thermal gradient-induced migration of Fe, Cr, and Ni from cladding into the fuel. This mechanism involves chemical transport of the metallic cladding components (as tellurides) in liquid Cs-Te. (author)

  4. Computer modelling of the WWER fuel elements under high burnup conditions by the computer codes PIN-W and RODQ2D

    International Nuclear Information System (INIS)

    Valach, M.; Zymak, J.; Svoboda, R.

    1997-01-01

    This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs

  5. Computer modelling of the WWER fuel elements under high burnup conditions by the computer codes PIN-W and RODQ2D

    Energy Technology Data Exchange (ETDEWEB)

    Valach, M; Zymak, J; Svoboda, R [Nuclear Research Inst. Rez plc, Rez (Czech Republic)

    1997-08-01

    This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs.

  6. Damping in LMFBR pipe systems

    International Nuclear Information System (INIS)

    Anderson, M.J.; Barta, D.A.; Lindquist, M.R.; Renkey, E.J.; Ryan, J.A.

    1983-06-01

    LMFBR pipe systems typically utilize a thicker insulation package than that used on water plant pipe systems. They are supported with special insulated pipe clamps. Mechanical snubbers are employed to resist seismic loads. Recent laboratory testing has indicated that these features provide significantly more damping than presently allowed by Regulatory Guide 1.61 for water plant pipe systems. This paper presents results of additional in-situ vibration tests conducted on FFTF pipe systems. Pipe damping values obtained at various excitation levels are presented. Effects of filtering data to provide damping values at discrete frequencies and the alternate use of a single equivalent modal damping value are discussed. These tests further confirm that damping in typical LMFBR pipe systems is larger than presently used in pipe design. Although some increase in damping occurred with increased excitation amplitude, the effect was not significant. Recommendations are made to use an increased damping value for both the OBE and DBE seismic events in design of LMFBR pipe systems

  7. Quasi-steady state boiling downstream of a central blockage in a 19-rod simulated LMFBR subassembly (FFM bundle 3B)

    International Nuclear Information System (INIS)

    Hanus, N.; Fontana, M.H.; Gnadt, P.A.; MacPherson, R.E.; Smith, C.M.; Wantland, J.L.

    1976-01-01

    Results of sodium boiling tests in a centrally blocked 19-rod simulated LMFBR subassembly are discussed. The tests were part of the experimental series conducted with bundle 3B in the Fuel Failure Mockup (FFM) at ORNL

  8. Major results on the development of high density U-Mo fuel and pin-type fuel elements executed under the Russian RERTR program and in cooperation with ANL (USA)

    International Nuclear Information System (INIS)

    Vatulin, A.; Morozov, A.; Stetsky, Y.; Suprun, V.; Dobrikova, I.; Trifonov, Y.; Mishunin, V.; Sorokin, V.

    2003-01-01

    VNIINM is active participant of 'Russian program on Reduced Enrichment for Research and Test Reactors'. Institute Works in two main directions: 1) development of new high-density fuels (HDF) and 2) development of new design of fuel elements with LEU. The development of the new type fuel element is carried out both for existing reactors, and for developing new advanced reactors. The 'TVEL' concern is coordinator of works of this program. The majority enterprises of branch (NIIAR, PIYaF, RRC KI, NZChK) take part in this work. Since 2000 these works are being conducted in cooperation with Argonne National Laboratory (USA) within the RERTR program under VNIINM with ANL contract. At the present, a large set of pre-pile investigations has been completed. All necessary fabrication procedures have been developed for utilization of U-Mo dispersion fuel in Russian-designed research reactors. For irradiation tests the pin-type mini-fuel elements with HDF dispersion fuel with LEU and the uranium density equaled to 4,0 and 6,0 g/cm 3 (up to 40 vol.%) have been manufactured. Their irradiation began in August 2003 in the MIR reactor (NIIAR, Dimitrovgrad). A large set of works for preparation of lifetime tests (WWR-M reactor in Gatchina) of two full-scale fuel assemblies with new pin-type fuel elements on basis LEU UO 2 -Al and UMo-Al fuels has been completed. The in-pile tests of fuel assemblies began in September 2003. The summary of important results of performed works and their near-term future are presented in paper. (author)

  9. ACPR upgrade fuel motion detection system

    International Nuclear Information System (INIS)

    Kelly, J.G.; Stalker, K.T.

    1976-01-01

    Coded aperture imaging techniques are being applied to the problem of fuel motion detection in core disruptive safety experiments in the LMFBR. The upgraded Sandia Annular Core Pulse Reactor will be used for fuel motion studies, and a system is now being developed to image fuel motion through a slot in the core in l- and 7-pin experiments. The γ-rays emitted by the fuel will be modulated by a coded aperture and will form a pseudohologram on a scintillator. This shadowgram will be lens coupled first to an optical image intensifier, and then to a rotating prism framing camera. The goal is to obtain quality images of fuel with approximately 1 mm spatial resolution at about 10 4 frames/sec. Some of the feasibility milestones which have been reached thus far are the following: (1) A fuel pin shaped 252 Cf fission gamma ray source has been imaged through 1.27 cm of steel (simulating reactor containment). (2) A 252 Cf source has been imaged through 1.27 cm of steel with x-ray image intensifier amplification. (3) Collimation and shielding tests indicate that at ACPR it should be possible to obtain adequate signal to noise at the detector when the fuel is observed through a slot in the core

  10. General considerations on the oxide fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    Pascard, R.

    1977-01-01

    Since the very first experimental irradiations in thermal reactors, performed in view of the future Rapsodie fuel general study, corrosion cladding anomalies were observed. After 10 years of Rapsodie and more than two years of Phenix, performance brought definite confirmation of the chemical reactions between the irradiated fuel and cladding. That is the reason for which the fuel designers express an urgent need for determining the corrosion rates. Semi-empirical laws and mechanisms describing corrosion processes are proposed. Erratic conditions for appearance of the oxide-cladding corrosion are stressed upon. Obviously such a problem can be fully appreciated only by a statistical approach based on a large number of observations on the true LMFBR fuel pins

  11. Effect of boron and gadolinium concentration on the calculated neutron multiplication factor of U(3)O2 fuel pins in optimum geometries

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1984-10-01

    The KENO-Va improved Monte Carlo criticality program is used to calculate the neutron multiplication factor for TMI-U2 fuel compositions in a variety of configurations and to display parametric regions giving rise to maximum reactivity contributions. The lattice pitch of UO 2 fuel pins producing a maximum k/sub eff/ is determined as a function of boron concentrations in the coolant for infinite and finite systems. The characteristics of U 3 O 8 -coolant mixtures of interest to modeling the rubble region of the core are presented. Several disrupted core configurations are calculated and comparisons made. The results should be useful to proposed defueling of the TMI-U2 reactor

  12. Advanced foil activation techniques for the measurement of within-pin distributions of the 63Cu(n,γ)64Cu reaction rate in nuclear fuel

    International Nuclear Information System (INIS)

    Macku, K.; Jatuff, F.; Murphy, M.F.; Joneja, O.P.; Bischofberger, R.; Chawla, R.

    2006-01-01

    Different foil activation techniques have been used for measuring spatial distributions of the 63 Cu(n,γ) 64 Cu reaction within two pins of a SVEA-96 Optima2 boiling water reactor fuel assembly, at the critical facility PROTEUS. This reaction is of interest because its 1/v cross-section gives it a good representation of the 235 U fission rate. Initially, radial capture rate profiles were measured with mechanically punched copper foils. More detailed profiles were then determined by using a 0.2 mm copper wire spiral (∼200 μm resolution), as well as 5-, 10-, and 20-ring UV-lithography, electroplating, and molding (UV-LIGA) foils (up to a 100 μm resolution). For azimuthal measurements, apart from manually cut activation foils (into 8 sectors), 8- and 12-sector LIGA foils were used. The highly versatile LIGA foils have the additional advantage of being very easily separated into individual pieces after irradiation without the use of punches or other cutting tools. In order to account for the invasive character of the foil activation techniques, corrections to account for sample perturbations and for self-shielding effects were determined via simplified Monte Carlo (MCNP4C) modeling of the experimental setup. The final results from the various measurements of 63 Cu(n,γ) 64 Cu within-pin distributions have been compared with MCNP computations employing a detailed model of the full SVEA Optima2 fuel assembly

  13. LARA: Expert system for acoustic localization of robot in a LMFBR

    International Nuclear Information System (INIS)

    Lhuillier, C.; Malvache, P.

    1986-12-01

    The expert system LARA (Acoustic Localization of Autonomic Robot) has been developed to show the interest of introducing artificial intelligency for fine automatic positioning of refuelling machine in a LMFBR reactor. LARA which is equipped with an acoustic detector gives rapidly a good positioning on the fuel [fr

  14. Cover gas seals. 11 - FFTF-LMFBR seal-test program, January-March 1974

    International Nuclear Information System (INIS)

    Kurzeka, W.; Oliva, R.; Welch, F.

    1974-01-01

    The objectives of this program are to: (1) conduct static and dynamic tests to demonstrate or determine the mechanical performance of full-size (cross section) FFTF fuel transfer machine and reactor vessel head seals intended for use in a sodium vapor - inert gas environment, (2) demonstrate that these FFTF seals or new seal configuration provide acceptable fission product and cover gas retention capabilities at LMFBR Clinch River Plant operating environmental conditions other than radiation, and (3) develop improved seals and seal technology for the LMFBR Clinch River Plant to support the national objective to reduce all atmospheric contaminations to low levels

  15. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  16. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched 235U fuel pins

    International Nuclear Information System (INIS)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched 235 U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are rather

  17. Experimental validation of radial reconstructed pin-power distributions in full-scale BWR fuel assemblies with and without control blade

    Energy Technology Data Exchange (ETDEWEB)

    Giust, Flavio, E-mail: flavio.giust@axpo.c [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne (Switzerland); Axpo Kernenergie AG, Parkstrasse 23, CH-5401 Baden (Switzerland); Grimm, Peter [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)

    2010-12-15

    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 paper presents comparisons of reconstructed 2D pin fission rates from nodal diffusion calculations to the experimental results in two configurations: one 'regular' (I-1A) and the other 'controlled' (I-2A). Both configurations consist of an array of 3 x 3 SVEA-96+ fuel assemblies moderated with light water at 20 {sup o}C. In configuration I-2A, an L-shaped hafnium control blade (half of a real cruciform blade) is inserted adjacent to the north-west corner of the central fuel assembly. To minimise the impact of the surroundings, all measurements were done in fuel pins belonging to the central assembly. The 3 x 3 experimental configuration (test zone) was modelled 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 partial current ratios (PCRs) calculated at the test zone boundary using a 3D Monte Carlo (MCNPX) model of the whole PROTEUS reactor. Sensitivity cases are analysed to show the impact of changes in the 2D lattice modelling on the calculated fission rate distribution and reactivity. Further, the effects of variations in the test zone boundary PCRs and their behaviour in energy are investigated. For the test zone configuration without control blade, the pin-power reconstruction methodology delivers the same level of accuracy as the 2D transport calculations. On the other hand, larger deviations that are inherent to the use of reflected geometry in the lattice calculations are observed for the configuration with the control blade inserted. In the basic (reference) simulation cases, the calculated-to-experimental (C

  18. Fission and corrosion products behavior in primary circuits of LMFBR's

    International Nuclear Information System (INIS)

    Feuerstein, H.; Thorley, A.W.

    1987-08-01

    Most of the 20 presented papers report items belonging to more than one session. The equipment results of primary circuits of LMFBR's relative to corrosion and fission products, release and chemistry of fuel, measurement techniques and analytical procedures of sodium sampling, difficulties with radionuclides and particles, reactor experiences with EBR-II, FFTF, BR10, BOR60, BN350, BN600, JOYO, and KNK-II, DFR, PFR, RAPSODIE, PHENIX, and SUPERPHENIX, and at least the verification of codes for calculation models of radioactive products accumulation and distribution are described. All 20 papers presented at the meeting are separately indexed in the database. (DG)

  19. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

  20. Happy Pinning

    DEFF Research Database (Denmark)

    Fausing, Bent

    2012-01-01

    This is about Pinterest, but with a different approach than usual to social networks. Pinterest is an image site par excellence. The images are as Windows that open outwards and also lets us look inwards and displays the soul and heart, the unintentional or pre-conscious desires. Happy Pinning!...