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Sample records for irradiated fuel pin

  1. 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)

  2. 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

  3. 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.

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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)

  9. 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)

  10. 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)

  11. 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

  12. 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.)

  13. 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)

  14. 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.

  15. 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

  16. 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

  17. 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.)

  18. 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

  19. 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.

  20. 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

  1. 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.)

  2. 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)

  3. 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

  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. 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)

  6. 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

  7. 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)

  8. 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

  9. 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

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

    International Nuclear Information System (INIS)

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

    1982-06-01

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

  11. 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

  12. 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

  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. 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.

  15. 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

  16. 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)

  17. 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

  18. 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.)

  19. 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

  20. 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

  1. Irradiation performance of full-length metallic IFR fuels

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.

    1992-07-01

    An assembly irradiation of 169 full-length U-Pu-Zr metallic fuel pins was successfully completed in FFTF to a goal burnup of 10 at.%. All test fuel pins maintained their cladding integrity during the irradiation. Postirradiation examination showed minimal fuel/cladding mechanical interaction and excellent stability of the fuel column. Fission-gas release was normal and consistent with the existing data base from irradiation testing of shorter metallic fuel pins in EBR-II

  2. 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

  3. 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)

  4. 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

  5. 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

  6. Fuel or irradiation subassembly

    International Nuclear Information System (INIS)

    Seim, O.S.; Hutter, E.

    1975-01-01

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

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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

  12. 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 %

  13. 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

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

    International Nuclear Information System (INIS)

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

    1982-03-01

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

  15. 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)

  16. CEA fuel pencil qualification under irradiation: from component conception to fuel assembly irradiation in a power reactor

    International Nuclear Information System (INIS)

    Marin, J.-F.; Pillet, Claude; Francois, Bernard; Morize, Pierre; Petitgrand, Sylvie; Atabek, R.-M.; Houdaille, Brigitte.

    1981-06-01

    Fabrication of fuel pins made of uranium oxide pellets and of a zircaloy 4 cladding is described. Irradiation experiment results are given. Thermomechanical behavior of the fuel pin in a power reactor is examined [fr

  17. 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)

  18. 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

  19. Nitride fuels irradiation performance data base

    International Nuclear Information System (INIS)

    Brozak, D.E.; Thomas, J.K.; Peddicord, K.L.

    1987-01-01

    An irradiation performance data base for nitride fuels has been developed from an extensive literature search and review that emphasized uranium nitride, but also included performance data for mixed nitrides [(U,Pu)N] and carbonitrides [(U,Pu)C,N] to increase the quantity and depth of pin data available. This work represents a very extensive effort to systematically collect and organize irradiation data for nitride-based fuels. The data base has many potential applications. First, it can facilitate parametric studies of nitride-based fuels to be performed using a wide range of pin designs and operating conditions. This should aid in the identification of important parameters and design requirements for multimegawatt and SP-100 fuel systems. Secondly, the data base can be used to evaluate fuel performance models. For detailed studies, it can serve as a guide to selecting a small group of pin specimens for extensive characterization. Finally, the data base will serve as an easily accessible and expandable source of irradiation performance information for nitride fuels

  20. 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

  1. 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

  2. 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)

  3. 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

  4. 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.

  5. Design and fuel fabrication processes for the AC-3 mixed-carbide irradiation test

    International Nuclear Information System (INIS)

    Latimer, T.W.; Chidester, K.M.; Stratton, R.W.; Ledergerber, G.; Ingold, F.

    1992-01-01

    The AC-3 test was a cooperative U.S./Swiss irradiation test of 91 wire-wrapped helium-bonded U-20% Pu carbide fuel pins irradiated to 8.3 at % peak burnup in the Fast Flux Test Facility. The test consisted of 25 pins that contained spherepac fuel fabricated by the Paul Scherrer Institute (PSI) and 66 pins that contained pelletized fuel fabricated by the Los Alamos National Laboratory. Design of AC-3 by LANL and PSI was begun in 1981, the fuel pins were fabricated from 1983 to 1985, and the test was irradiated from 1986 to 1988. The principal objective of the AC-3 test was to compare the irradiation performance of mixed-carbide fuel pins that contained either pelletized or sphere-pac fuel at prototypic fluence and burnup levels for a fast breeder reactor

  6. Thermal/hydraulic bowing stability analysis of grid-supported multi-pin bundles with differential swelling and irradiation creep

    International Nuclear Information System (INIS)

    McAreavey, G.

    1977-01-01

    Azimuthal variations of clad temperature in fuel pin bundles leads to pin bowing by differential thermal expansion. During irradiation in a fast flux further possibly more severe bowing is caused by differential neutron induced voidage swelling, which, being temperature sensitive, will also vary azimuthally. The problem of pin bowing in a fuel element cluster involves consideration of the thermal/hydraulic behaviour, allowing for both inherent and induced clad temperature non-uniformities, coupled with the restrained bowing behaviour, including differential thermal expansion, differential swelling, and irradiation creep. All pins must be considered simultaneously. In the temperature and stress ranges of interest thermal creep may be neglected. An existing computer code, IAMBIC solves the zero time thermal bowing problem for a cluster of up to 61 pins on hexagonal pitch, with up to 21 supports at arbitrary axial spacing. The present paper describes the basis of TRIAMBIC, a time dependent code which analyses the irradiation induced effects in fuel pin bunbles due to fast neutrons. (Auth.)

  7. 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

  8. 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

  9. 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

  10. 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)

  11. Fission product phases in irradiated carbide fuels

    International Nuclear Information System (INIS)

    Ewart, F.T.; Sharpe, B.M.; Taylor, R.G.

    1975-09-01

    Oxide fuels have been widely adopted as 'first charge' fuels for demonstration fast reactors. However, because of the improved breeding characteristics, carbides are being investigated in a number of laboratories as possible advanced fuels. Irradiation experiments on uranium and mixed uranium-plutonium carbides have been widely reported but the instances where segregate phases have been found and subjected to electron probe analysis are relatively few. Several observations of such segregate phases have now been made over a period of time and these are collected together in this document. Some seven fuel pins have been examined. Two of the irradiations were in thermal materials testing reactors (MTR); the remainder were experimental assemblies of carbide gas bonded oxycarbide and sodium bonded oxycarbide in the Dounreay Fast Reactor (DFR). All fuel pins completed their irradiation without failure. (author)

  12. Irradiation test of fuel containing minor actinides in the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Soga, Tomonori; Sekine, Takashi; Wootan, David; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

    2007-01-01

    The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP, accounting for both prompt and delayed heating components, and then adjusted using E/C for 10 B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO 2-x or AmO 2-x in the (U, Pu)O 2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel. (author)

  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. 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

  15. Post-irradiation examinations of uranium-plutonium mixed nitride fuel irradiated in JMTR (89F-3A capsule)

    International Nuclear Information System (INIS)

    Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Arai, Yasuo; Kimura, Yasuhiko; Nagashima, Hisao; Sekita, Noriaki

    2000-03-01

    Two helium-bonded fuel pins filled with uranium-plutonium mixed nitride pellets were encapsulated in 89F-3A and irradiated in JMTR up to 5.5% FIMA at a maximum linear power of 73 kW/m. The capsule cooled for ∼5 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pins. Very low fission gas release rate of about 2 ∼ 3% was observed, while the diametric increase of fuel pin was limited to ∼0.4% at the position of maximum reading. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

  16. 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)

  17. 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

  18. 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

  19. 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

  20. 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.)

  1. 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

  2. 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)

  3. 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)

  4. 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.

  5. Post-irradiation examinations of inert matrix nitride fuel irradiated in JMTR (01F-51A capsule)

    International Nuclear Information System (INIS)

    Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Honda, Junichi; Hatakeyama, Yuichi; Ono, Katsuto; Matsui, Hiroki; Arai, Yasuo

    2007-03-01

    A plutonium nitride fuel pin containing inert matrix such as ZrN and TiN was encapsulated in 01F-51A and irradiated in JMTR. Minor actinides are surrogated by plutonium. Average linear powers and burnups were 408W/cm, 30000MWd/t(Zr+Pu) [132000MWd/t-Pu] for (Zr,Pu)N and 355W/cm, 38000MWd/t(Ti+Pu) [153000MWd/t-Pu] for (TiN,PuN). The irradiated capsule was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pin. Very low fission gas release rate of about 1.6% was measured. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

  6. 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

  7. The post irradiation examination of a sphere-pac (UPu)C fuel pin irradiated in the BR-2 reactor (MFBS 7 experiment)

    International Nuclear Information System (INIS)

    Smith, L.; Aerne, E.T.; Buergisser, B.; Flueckiger, U.; Hofer, R.; Petrik, F.

    1979-09-01

    A pin fuelled with Swiss made (UPu)C microspheres has been successfully irradiated to a peak burn-up of 6% fima in the Belgian BR2 Reactor. The pin, rated up to 95 kW/m, was intact after irradiation and exhibited a peak strain of just over 0.5%. The results of the post irradiation examination are reported. (Auth.)

  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. 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

  10. 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

  11. 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

  12. 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%)

  13. 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.

  14. French approach in fuel pin modelling for fast reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-12-01

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

  15. 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

  16. 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.)

  17. Status of steady-state irradiation testing of mixed-carbide fuel designs

    International Nuclear Information System (INIS)

    Harry, G.R.

    1983-01-01

    The steady-state irradiation program of mixed-carbide fuels has demonstrated clearly the ability of carbide fuel pins to attain peak burnup greater than 12 at.% and peak fluences of 1.4 x 10 23 n/cm 2 (E > 0.1 MeV). Helium-bonded fuel pins in 316SS cladding have achieved peak burnups of 20.7 at.% (192 MWd/kg), and no breaches have occurred in pins of this design. Sodium-bonded fuel pins in 316SS cladding have achieved peak burnups of 15.8 at.% (146 MWd/kg). Breaches have occurred in helium-bonded fuel pins in PE-16 cladding (approx. 5 at.% burnup) and in D21 cladding (approx. 4 at.% burnup). Sodium-bonded fuel pins achieved burnups over 11 at.% in PE-16 cladding and over 6 at.% in D9 and D21 cladding

  18. Post-irradiation examination of the vipac fuel assemblies IFA-104 and IFA-203, irradiated in the Halden boiling water reactor

    International Nuclear Information System (INIS)

    Linde, A. van der; Lucas Luyckx, H.J.B.

    1976-11-01

    Two seven pin assemblies, IFA-104 and IFA-203, have been irradiated in the Halding Boiling Water Reactor. The Zircaloy clad pins contained vibrationally compacted, sharp edged UO 2 particles with smear densities of 84-88% theoretical density and a fuel stack length of about 1520 mm. The IFA-104 pins operated satisfactorily during approximately 240 effective full power days at an average linear heat rating of about 340 W/cm, and a peak rating of about 480 W/cm. Ingress of water via a pressure transducer into one of the IFA-203 pins necessitated the premature termination of the irradiation after approximately 30 effective full power days at an average rating of about 460 W/cm, and a peak rating of about 550 W/cm. One IFA-104 pin and two IFA-203 pins failed early. The primary cause of these failures has been sought in the oxidation reaction of the tantalum tube which protected the W/Re thermocouples with the surrounding UO 2 or with moisture in the fuel at temperatures in excess of approximately 1700degC. The fuel centre temperature and the internal gas pressure measurements during the beginning of lifetime and the results of the post-irradiation chemical and microscopic analysis on fuel pin cross-sections have been correlated with corresponding data calculated on a Gapcon-Thermal computer programme written for pellet fuel pins. Fairly good agreement could be achieved between the measured data on the vipac pins and the calculated data of a pellet pin. During the corrosion of the Zircaloy-2 cladding of the IFA-104 pins which were autoclaved in 400degC steam prior to the irradiation, relatively large amounts of H 2 entered the cladding from the D 2 O coolant side. With 0.66 mole% H 2 O in the D 2 O the calculated hydrogen uptake preference ratio was 33

  19. 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

  20. 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.)

  1. Status on the construction of the fuel irradiation test facility

    International Nuclear Information System (INIS)

    Park, Kook Nam; Sim, Bong Shick; Lee, Chung Young; Yoo, Seong Yeon

    2005-01-01

    As a facility to examine general performance of nuclear fuel under irradiation condition in HANARO, Fuel Test Loop(FTL) has been developed which can accommodate 3 fuel pins at the core irradiation hole(IR1 hole) taking consideration user's test requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of- Pile System (OPS). Test condition in IPS such as pressure, temperature and the water quality, can be controlled by OPS. 3-Pin FTL Conceptual design was set up in 2001 and had completed detail design including a design requirement and basic Piping and Instrument Diagram (P and ID) in 2004. The safety analysis report was prepared and submitted in early 2005 to the regulatory body(KINS) for review and approval of FTL. In 2005, the development team is going to purchase and manufacture hardware and make a contract for construction work. In 2006, the development team is going to install an FTL system performance test shall be done as a part of commissioning. After a 3-Pin FTL development which is expected to be finished by the 2007, FTL will be used for the irradiation test of the new PWR-type fuel and the usage of HANARO will be enhanced

  2. 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.

  3. 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

  4. Post-irradiation examination of a failed PHWR fuel bundle of KAPS-2

    International Nuclear Information System (INIS)

    Mishra, Prerna; Unnikrishnan, K.; Viswanathan, U.K.; Shriwastaw, R.S.; Singh, J.L.; Ouseph, P.M.; Alur, V.D.; Singh, H.N.; Anantharaman, S.; Sah, D.N.

    2006-08-01

    Detailed post irradiation examination was carried out on a PHWR fuel bundle irradiated at Kakrapar Atomic Power Station unit 2 (KAPS-2). The fuel bundle had failed early in life at a low burnup of 387 MWd/T. Non destructive and destructive examination was carried out to identify the cause of fuel failure. Visual examination and leak testing indicated failure in two fuel pins of the outer ring of the bundle in the form of axial cracks near the end plug location. Ultrasonic testing of the end cap weld indicated presence of lack of fusion type defect in the two fuel pins. No defect was found in other fuel pins of the bundle. Metallographic examination of fuel sections taken from the crack location in the failed fuel pin showed extensive restructuring of fuel. The centre temperature of the fuel had exceeded 1700 degC at this location in the failed fuel pin, whereas fuel centre temperature in the un-failed fuel pin was only about 1300 degC. Severe fuel clad interaction was observed in the failed fuel pin at and near the location of failure but no such interaction was observed in the un-failed fuel pins. Several incipient cracks originating from the inside surface were found in the cladding near failure location in addition to the main through wall crack. The incipient cracks were filled with interaction products and hydride platelets were present at tip of the cracks. It was concluded from the observations that the primary cause of failure was the presence of a part-wall defect in the end cap weld of the fuel pins. These defects opened up during reactor operation leading to steam ingress into the fuel, which caused high fuel centre temperature and severe fuel-cladding interaction resulting in secondary failures. A more stringent inspection and quality control of end plug weld during fabrication using ultrasonic test has been recommended to avoid such failure. (author)

  5. 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

  6. 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

  7. 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

  8. In-pile measurement of the thermal conductivity of irradiated metallic fuel

    International Nuclear Information System (INIS)

    Bauer, T.H.; Holland, J.W.

    1995-01-01

    Transient test data and posttest measurements from recent in-pile overpower transient experiments are used for an in situ determination of metallic fuel thermal conductivity. For test pins that undergo melting but remain intact, a technique is described that relates fuel thermal conductivity to peak pin power during the transient and a posttest measured melt radius. Conductivity estimates and their uncertainty are made for a database of four irradiated Integral Fast Reactor-type metal fuel pins of relatively low burnup (<3 at.%). In the assessment of results, averages and trends of measured fuel thermal conductivity are correlated to local burnup. Emphasis is placed on the changes of conductivity that take place with burnup-induced swelling and sodium logging. Measurements are used to validate simple empirically based analytical models that describe thermal conductivity of porous media and that are recommended for general thermal analyses of irradiated metallic fuel

  9. 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

  10. 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)

  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. 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.)

  13. 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

  14. 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.))

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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)

  1. 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)

  2. Safety aspects of advanced fuels irradiations in EBR-II

    International Nuclear Information System (INIS)

    Lehto, W.K.

    1975-09-01

    Basic safety questions such as MFCI, loss-of-Na bond, pin behavior during design basis transients, and failure propagation were evaluated as they pertain to advanced fuels in EBR-II. With the exception of pin response to the unlikely loss-of-flow transient, the study indicates that irradiation of significant numbers of advanced fueled subassemblies in EBR-II should pose no safety problems. The analysis predicts, however, that Na boiling may occur during the postulated design basis unlikely loss-of-flow transient in subassemblies containing He-bonded fuel pins with the larger fuel-clad gaps. The calculations indicate that coolant temperatures at top of core in the limiting S/A's, containing the He bonded pins, would reach approximately 1480 0 F during the transient without application of uncertainty factors. Inclusion of uncertainties could result in temperature predictions which approach coolant boiling temperatures (1640 0 F). Further analysis of He-bonded pins is being done in this potential problem area, e.g., to apply best estimates of uncertainty factors and to determine the sensitivity of the preliminary results to gap conductance

  3. Finite element method programs to analyze irradiation behavior of fuel pellets

    International Nuclear Information System (INIS)

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

    1979-09-01

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

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 75 MeV boron ion irradiation studies on Si PIN photodiodes

    Energy Technology Data Exchange (ETDEWEB)

    Prabhakara Rao, Y.P.; Praveen, K.C. [Department of Studies in Physics, University of Mysore, Manasagangotri, Mysore 570006 (India); Rejeena Rani, Y. [Integrated Circuits Division, Bharat Electronics Limited, Bangalore 560013, Karnataka (India); Tripathi, Ambuj [Inter University Accelerator Centre, Aruna Asaf Ali Marg, New Delhi 110067 (India); Gnana Prakash, A.P., E-mail: gnanap@hotmail.com [Department of Studies in Physics, University of Mysore, Manasagangotri, Mysore 570006 (India)

    2013-12-01

    The highly sensitive silicon PIN photodiodes were fabricated to use in radiation environments. The Si PIN photodiodes are coated with 150 nm silicon dioxide (SiO{sub 2}) as anti-reflective (AR) coating. The presence of AR coating on the performance of irradiated PIN photodiodes is studied up to a total dose of 10 Mrad. The effects of 75 MeV boron (B{sup 5+}) ions and {sup 60}Co gamma radiation on the I–V, C–V and spectral responses of PIN photodiodes were studied systematically to understand the radiation tolerance of the devices. The 75 MeV B{sup 5+} irradiation results are compared with {sup 60}Co gamma irradiated results in the same dose range for 1 mm × 1 mm and 10 mm × 10 mm active area PIN photodiodes. The irradiation results show that the ion irradiated PIN photodiodes show more degradation when compared {sup 60}Co gamma irradiated devices. The irradiation results are presented in this paper and the possible mechanism behind the degradation of photodiodes is also discussed in the paper.

  9. 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

  10. 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

  11. Post irradiation examinations of uranium-plutonium mixed carbide fuels irradiated at low linear power rate

    International Nuclear Information System (INIS)

    Maeda, Atsushi; Sasayama, Tatsuo; Iwai, Takashi; Aizawa, Sakuei; Ohwada, Isao; Aizawa, Masao; Ohmichi, Toshihiko; Handa, Muneo

    1988-11-01

    Two pins containing uranium-plutonium carbide fuels which are different in stoichiometry, i.e. (U,Pu)C 1.0 and (U,Pu)C 1.1 , were constructed into a capsule, ICF-37H, and were irradiated in JRR-2 up to 1.0 at % burnup at the linear heat rate of 420 W/cm. After being cooled for about one year, the irradiated capsule was transferred to the Reactor Fuel Examination Facility where the non-destructive examinations of the fuel pins in the β-γ cells and the destructive ones in two α-γ inert gas atmosphere cells were carried out. The release rates of fission gas were low enough, 0.44 % from (U,Pu)C 1.0 fuel pin and 0.09% from (U,Pu)C 1.1 fuel pin, which is reasonable because of the low central temperature of fuel pellets, about 1000 deg C and is estimated that the release is mainly governed by recoil and knock-out mechanisms. Volume swelling of the fuels was observed to be in the range of 1.3 ∼ 1.6 % for carbide fuels below 1000 deg C. Respective open porosities of (U,Pu)C 1.0 and (U,Pu)C 1.1 fuel were 1.3 % and 0.45 %, being in accordance with the release behavior of fission gas. Metallographic observation of the radial sections of pellets showed the increase of pore size and crystal grain size in the center and middle region of (U,Pu)C 1.0 pellets. The chemical interaction between fuel pellets and claddings in the carbide fuels is the penetration of carbon in the fuels to stainless steel tubes. The depth of corrosion layer in inner sides of cladding tubes ranged 10 ∼ 15 μm in the (U,Pu)C 1.0 fuel and 15 #approx #25 μm in the (U,Pu)C 1.1 fuel, which is correlative with the carbon potential of fuels posibly affecting the amount of carbon penetration. (author)

  12. Measurement of burnup in FBR MOX fuel irradiated to high burnup

    International Nuclear Information System (INIS)

    Koyama, Shin-ichi; Osaka, Masahiko; Sekine, Takashi; Morozumi, Katsufumi; Namekawa, Takashi; Itoh, Masahiko

    2003-01-01

    The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28% FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector. (author)

  13. Accurate analysis of pins irradiated in fast neutron reactors

    International Nuclear Information System (INIS)

    Regnaud, F.; Guichard, C.

    1976-01-01

    A device was developed for the chemical analysis of fissile materials in a whole fuel pin in order to determine U and Pu with an accuracy of about 0.1% in samples from 100 to 200g, irradiated at burn-up between 50 and 100.000MWd/t. This unit is provided with two shielded α hot cells. A dissolving device was developed which enables the dissolution of residues. Pu is determined by oxydation-reduction titration and U by the isotopic dilution method [fr

  14. 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)

  15. 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.

  16. 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

  17. 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)

  18. 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

  19. 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.

  20. Irradiation tests report of the 32nd cycle in 'JOYO'

    International Nuclear Information System (INIS)

    1998-09-01

    This report summarizes the operating and irradiation data of the experimental reactor 'JOYO' 32nd cycle, and estimates the 33rd cycle irradiation condition. Irradiation tests in the 31st cycle are as follows: (1) B-type irradiation rig (B9). (a) High burn up performance tests of MONJU' fuel pins, advanced austenitic steel cladding fuel pins, large diameter fuel pins, ferrite steel cladding fuel pins (in collaboration with the USA) and large diameter annular pellet fuel pins. (b) Mixed carbide and nitride fuel pins irradiation tests (in collaboration with JAERI). (2) C-type irradiation rig (C4F). (a) High burn up performance test of advanced austenitic steel cladding fuel pins (in collaboration with France). (3) C-type irradiation rig (C6D). (a) Large diameter fuel pins irradiation test. (4) Absorber Materials Irradiation Rig (AMIR-6). (a) Run to absorber pin's cladding breach. (5) Absorber Materials Irradiation Rig (AMIR-8). (a) High-temperature shroud and Na-bond elements tests. (6) Core Materials Irradiation Rig (CMIR-5-1). (a) Core materials irradiation tests. (7) Structure Materials Irradiation Rigs (SMIR). (a) Material irradiation tests (in collaboration with universities). (b) Surveillance back up tests for MONJU'. (8) MAterial testing RIg with temperature COntrol (MARICO-1). (a) Material irradiation tests (in collaboration with universities), (b) Creep rupture tests of the core materials for the demonstration reactor. (9) Upper core structure irradiation Plug Rig (UPR-1-5). (a) Upper core neutron spectrum effect and accelerated irradiation effect. The maximum burn-up driver assembly 'PFD503' reached 65,600 MWd/t (pin average). (author)

  1. 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.

  2. 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)

  3. 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

  4. 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

  5. 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

  6. 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.

  7. 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)

  8. 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.)

  9. 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

  10. 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}.

  11. 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)

  12. 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.)

  13. 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)

  14. 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

  15. 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.)

  16. 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.)

  17. 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)

  18. 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

  19. 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)

  20. 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.)

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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

  6. Breached-pin testing in the US

    International Nuclear Information System (INIS)

    Mahagin, D.E.; Lambert, J.D.B.

    1981-04-01

    Experience gained at EBR-II by the late 1970's from a significant number of failures in experimental fuel-pin irradiations forms the basis of a program directed towards the characterization of breached pins. The questions to be answered and the issues raised by further testing are discussed

  7. 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

  8. 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

  9. 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

  10. 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

  11. An Analysis of the Thermal and Structure Behaviour of the UO2-PuO2-Fuel in the Irradiation Experiment of the UO2-PuO2-Fuel in the Irradiation Experiment FR2 Capsule Test Series 5a

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Helmut, E.

    1981-01-01

    In the Karlsruhe research reactor FR2 nine fuel pins were irradiated within three irradiation capsules in the course of the test series 5a. The pins contained UO 2 -PuO 2 fuel pellets. They reached bump values of about 6, 17 and 47 Mwd/Kg Me with linear rod powers of 400 to 600 W/cm and clad surface temperature between 500 and 700 degree centigree. A detailed analysis of the fuel structuration data (columnar-grain and equiaxed- -grain growth regions) have allowed to determine, with the help of physic-mathematical models, the radii of these regions and the heat transfer through the contact zone between fuel and clad depending on the bump. The results of the analysis showed that the fuel surface temperature rose with increasing burnup. (Author) 16 refs

  12. 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

  13. 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.)

  14. 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

  15. 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)

  16. Volume pinning force and upper critical field of irradiated Nb3Sn

    International Nuclear Information System (INIS)

    Maier, P.; Seibt, E.

    1981-01-01

    Irradiation by neutrons and ions in A15 superconductors (Nb 3 Sn, V 3 Ga) exerts a stronger influence on the pinning behavior than in nonordered alloys (NbTi). In this work it is shown for deuteron irradiated Nb 3 /Sn wires prepared by the bronze process that the dose curve of the volume pinning force P/sub V/ can be conveniently described by a sum of two terms, due to the grain boundary pinning and to the radiation pinning, respectively. After deduction of the contribution by the radiation-induced pinning centers, good agreement is obtained between the measured P/sub V/ values and those calculated using the upper critical field B/sub c/2 and the transition temperature T/sub c/ on the basis of the irradiation fluence. The use of a theoretical relationship between B/sub c/2 and T/sub c/ is supported by measured values. Application to multifilamentary superconductors with high current carrying capabilities simplifies the calculation of P/sub V/, since the radiation induced volume pinning force can be neglected

  17. 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

  18. 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

  19. Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

    International Nuclear Information System (INIS)

    Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

    1985-01-01

    The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

  20. 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

  1. 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)

  2. Analysis of the porosity distribution of mixed oxide pins

    International Nuclear Information System (INIS)

    Lieblich, M.; Lopez, J.

    1987-01-01

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

  3. 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

  4. Method for monitoring irradiated nuclear fuel using cerenkov radiation

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  5. 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

  6. 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

  7. 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.

  8. 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)

  9. 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.)

  10. 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

  11. 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

  12. Molybdenum-UO2 cerment irradiation at 1145 K

    Science.gov (United States)

    Mcdonald, G.

    1971-01-01

    Two molybdenum-UO2 cermet fuel pins were fission heated in a helium-cooled loop at a temperature of 1145 K and to a total burnup of 5.3 % of the U-235. After irradiation the fuel pins were measured to check dimensional stability, punctured at the plenums to determine fission gas release, and examined metallographically to determine the effect of irradiation. Burnup was determined in several sections of the fuel pin. The results of the postirradiation examination indicated: (1) There was no visible change in the fuel pins on irradiation under the above conditions. (2) The maximum swelling of the fuel pins was less than 1%. (3) There was no migration of UO2 and no visible interaction between the molybdenum and the UO2. (4) Approximately 12% of the fission gas formed was released from the cermet cone into the gas plenum.

  13. Post-irradiation examinations on the KNK II/1 fuel element NY-203 with 400 equivalent full-power days residence time and 10 % burnup

    International Nuclear Information System (INIS)

    Patzer, G.; Geier, F.

    1984-09-01

    The fuel assembly NY-203 has been irradiated in the first core of KNK II up to a burnup of about 10 % and a residence time of 400 equivalent full-power days. The assembly contained 211 fuel pins with 6.0 mm outer diameter and fuel pellets with the composition (U 0 .7Pu 0 .3)O 2 .00. The cladding material was the austenitic steel 1.4988 lg. Some selected pins were examined in the hot cells of the KfK Karlsruhe. The post-irradiation examinations did not reveal any critical design aspects [de

  14. 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.)

  15. 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)

  16. Fission gas retention and axial expansion of irradiated metallic fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

    1986-05-01

    Out-of-reactor experiments utilizing direct electrical heating and infrared heating techniques were performed on irradiated metallic fuel. The results indicate accelerated expansion can occur during thermal transients and that the accelerated expansion is driven by retained fission gases. The results also demonstrate gas retention and, hence, expansion behavior is a function of axial position within the pin

  17. Experimental confirmation of the design to minimize vibration and wear in 61-pin wire-spaced EBR-II subassemblies

    International Nuclear Information System (INIS)

    Fukuda, S.K.

    1978-05-01

    Examinations of HEDL 61-pin subassemblies comprised of 5.84 mm (0.230) inch diameter mixed-oxide fuel pins with 1.02 mm (0.040'') diameter spacer wire (PNL-9, -10, -11, HEDL-N-E, -N-F), showed severe cladding and spacer wire wear after irradiation in EBR-II. A comparison of a large number of design, fabrication, and irradiation parameters for all of the HEDL subassemblies indicated that the porosity per ring of fuel pins correlated significantly with the occurrence of wear on the fuel pins. The porosity per ring is the clearance between the flat-to-flat pin bundle dimension and the inner hex can dimension divided by the number of hexagonal fuel pin rings in the subassembly. The porosity per ring for PNL-9, -10, -11 and HEDL-N-E was 0.15 mm/ring (6 mils/ring) and 0.18 mm/ring (7 mils/ring) for the HEDL-N-F subassembly. Since the original FTR subassembly design had a porosity/ring spread of 0.04 mm/ring to 0.16 mm/ring (1.67 to 6.11 mils/ring) an additional series of irradiation tests was conducted to confirm that a tighter fuel pin bundle would eliminate the wear

  18. The German carbide program: Performance, experimental findings, and evaluation of irradiation results

    International Nuclear Information System (INIS)

    Steiner, H.; Freund, D.; Geithoff, D.

    1982-09-01

    In this report a synopsis of the German carbide program is presented. The program comprises the irradiation of about 100 carbide pins equipped with pelletted fuel. Most of these fuel pins were He-bonded, the sodium bonding concept taken as a back-up solution. The main design parameters such as smear and pellet density, gap size, pin diameter and wall thickness as well as the irradiation conditions were varied mostly within wide ranges. Based on a compilation of relevant pin parameters, irradiation conditions, and the results of various irradiation experiments conclusions on the optimum ranges of the main design parameters are drawn. Furthermore, some important aspects of fuel pin behaviour are discussed based on quantitative results from post irradiation examinations. (orig.) [de

  19. 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

  20. 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

  1. 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

  2. Long term storage effects of irradiated fuel elements on power distribution and reactivity

    Energy Technology Data Exchange (ETDEWEB)

    Ponzoni Filho, P.; Sato, Sadakatu; Santos, Teresinha Ipojuca Cardoso T.I.C.; Fernandes Vanderlei Borba [FURNAS, Rio de Janeiro, RJ (Brazil); Fetterman, R.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-12-31

    The ALPHA/PHOENIX-P/ANC (APA) code package was used to calculate the pin by pin power distribution and reactivity for Angra 1 Power Plant, Cycle 5. The Angra 1 Cycle 5 core was loaded with several irradiated fuel elements which were stored in the Spent Fuel Pool (SFP) for more than 8 years. Generally, neutronic codes take into account the buildup and depletion of just a few key fission, products such as Sm-149. In this paper it is shown that the buildup effects of other fission products must be considered for fuel which has been out of the core for significant periods of time. Impacts of these other fission products can change core reactivity and power distribution. (author). 3 refs, 4 figs, 4 tabs.

  3. Long term storage effects of irradiated fuel elements on power distribution and reactivity

    International Nuclear Information System (INIS)

    Ponzoni Filho, P.; Sato, Sadakatu; Santos, Teresinha Ipojuca Cardoso T.I.C.; Fernandes Vanderlei Borba; Fetterman, R.J.

    1995-01-01

    The ALPHA/PHOENIX-P/ANC (APA) code package was used to calculate the pin by pin power distribution and reactivity for Angra 1 Power Plant, Cycle 5. The Angra 1 Cycle 5 core was loaded with several irradiated fuel elements which were stored in the Spent Fuel Pool (SFP) for more than 8 years. Generally, neutronic codes take into account the buildup and depletion of just a few key fission, products such as Sm-149. In this paper it is shown that the buildup effects of other fission products must be considered for fuel which has been out of the core for significant periods of time. Impacts of these other fission products can change core reactivity and power distribution. (author). 3 refs, 4 figs, 4 tabs

  4. 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

  5. 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

  6. 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)

  7. Precise measurement of fuel content of irradiated and nonirradiated materials

    International Nuclear Information System (INIS)

    Harker, Y.D.; Napper, P.R.; Proctor, A.E.

    1984-01-01

    This paper discusses the application of precise reactivity measurements in the Advanced Reactivity Measurement Facility at Idaho National Engineering Laboratory (INEL) to determine th fuel content in irradiated and nonirradiated materials. Different methods of reactivity measurements and examples of how they have been are presented, which provides an insight in capabilities available to analyze samples with different geometrical sizes from small volumes approx. 100 cc to 12 ft long fuel pins and also samples with different fuel content ranges from approx. 2 mg to approx. 600 g. The overall accuracy of these measurements is approx. 0.5% (1sigma)

  8. Behaviour of irradiated uranium silicide fuel revisited

    International Nuclear Information System (INIS)

    Finlay, M. Ross; Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

    2002-01-01

    Irradiated U 3 Si 2 dispersion fuels demonstrate very low levels of swelling, even at extremely high burn-up. This behaviour is attributed to the stability of fission gas bubbles that develop during irradiation. The bubbles remain uniformly distributed throughout the fuel and show no obvious signs of coalescence. Close examination of high burn-up samples during the U 3 Si 2 qualification program revealed a bimodal distribution of fission gas bubbles. Those observations suggested that an underlying microstructure was responsible for the behaviour. An irradiation induced recrystallisation model was developed that relied on the presence of sufficient grain boundary surface to trap and pin fission gas bubbles and prevent coalescence. However, more recent work has revealed that the U 3 Si 2 becomes amorphous almost instantaneously upon irradiation. Consequently, the recrystallisation model does not adequately explain the nucleation and growth of fission gas bubbles in U 3 Si 2 . Whilst it appears to work well within the range of measured data, it cannot be relied on to extrapolate beyond that range since it is not mechanistically valid. A review of the mini-plates irradiated in the Oak Ridge Research Reactor from the U 3 Si 2 qualification program has been performed. This has yielded a new understanding of U 3 Si 2 behaviour under irradiation. (author)

  9. 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

  10. 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

  11. 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)

  12. 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.)

  13. 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

  14. 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

  15. GCFR Fuels and Materials Program at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Neimark, L.A.; Greenberg, S.; Johnson, C.E.; Purohit, A.; Liu, Y.Y.; Rest, J.; Reimann, K.J.; McLennan, G.A.

    1979-01-01

    The F-5 fuel-pin irradiation experiment in EBR-II is a cornerstone of the GCFR program. It is the largest-scale fuel-pin experiment in the present program and will provide data on the performance of pins and a pin-support structure that are prototypic of the GCFR Demonstration Plant. The fuel pins are presently undergoing interim examination after successfully achieving 4.6 at.% burnup. Studies of the thermodynamics and kinetics of the U--Cs--O system, supplemented by analysis of the results of previously irradiated fuel pins, have led to the incorporation of fuel-design modifications in the F-5 experiment to insure adequate performance of the vented fuel. The effect of ribbing, as well as the ribbing process, on the short- and long-term structural performance of fuel-pin cladding is being evaluated via in-reactor and out-of-reactor tests and with the fuel-element modeling code LIFE-GCFR and the finite element program, ADINA

  16. Composite fuel behaviour under and after irradiation

    International Nuclear Information System (INIS)

    Dehaudt, P.; Mocellin, A.; Eminet, G.; Caillot, L.; Delette, G.; Bauer, M.; Viallard, I.

    1997-01-01

    Two kinds of composite fuels have been irradiated in the SILOE reactor. They are made of UO 2 particles dispersed in a molybdenum metallic (CERMET) or a MgAl 2 O 4 ceramic (CERCER) matrix. The irradiation conditions have allowed to reach a 50000 MWd/t U burn-up in these composite fuels after a hundred equivalent full power days long irradiation. The irradiation is controlled by a continuous measure of the pellet centre line temperature. It allows to have information about the TANOX rods thermal behaviour and the fuels thermal conductivities in comparing the centre line temperature versus linear power curves among themselves. Our results show that the CERMET centre line temperature is much lower than the CERCER and UO 2 ones: 520 deg. C against 980 deg. C at a 300W/cm linear power. After pin puncturing tests the rods are dismantled to recover each fuel pellet. In the CERCER case, the cladding peeling off has revealed that the fuel came into contact with the cladding and that some of the pellets were linked together. Optical microscopy observations show a changing of the MgAl 2 O 4 matrix state around the UO 2 particles at the pellets periphery. This transformation may have caused a swelling and would be at the origin of the pellet-cladding and the pellet-pellet interactions. No specific damage is seen after irradiation. The CERMET pellets are not cracked and remain as they were before irradiation. The CERCER crack network is slightly different from that observed in UO 2 . Kr retention was evaluated by annealing tests under vacuum at 1580 deg. C or 1700 deg. C for 30 minutes. The CERMET fission gas release is lower than the CERCER one. Inter- and intragranular fission gas bubbles are observed in the UO 2 particles after heat treatments. The CERCER pellet periphery has also cracked and the matrix has transformed again around UO 2 particles to present a granular and porous aspect. (author). 4 refs, 6 figs, 2 tabs

  17. Preliminary calculations of stress change of fuel pin using SiC/SiC composites for GFR with changing of thermal conductivity degradation by irradiation

    International Nuclear Information System (INIS)

    Lee, J. K.; Naganuma, M.

    2006-01-01

    Gas cooled Fast Reactor (GFR) is being researched as a candidate concept of Generation IV international Forum. As a main feature of GFR, it should be maintained high temperature and pressure of coolant gas for heat transfer efficiency. Such a demanding environment requires high-temperature-resistant structural materials distinguished from traditional steel material. Consequently, ceramics are promising candidate material of core components. Especially, Silicon Carbide fiber reinforced Silicon Carbide composites (SiC/SiC) have encouraging characteristics such as refractoriness, low activation and toughness. Application of new material to core components must be explained by the viewpoint of engineering validity. Therefore, present study surveyed that current report for mechanical strength and thermal conductivity of SiC/SiC composites. According to the reports, neutron irradiation environment degraded mechanical properties of SiC/SiC composites. To confirm applicability to core components, model of fuel pin using SiC/SiC composites was assumed with feasible mechanical properties. Furthermore, it was calculated and estimated that the stress caused by temperature variation of inner and outer side of assumed model of cladding tube. Stress was calculated by changing of input date such as thickness of cladding tube, temperature variation, thermal conductivity and linear power. In the range of this study, the most important factor was identified as degradation of thermal conductivity by irradiation. It caused a significant stress and limited a geometrical design of fuel pin. It was discussed that the differences of heat transfer between isotropic and anisotropic materials like a metal and composites. These results should be helpful not only to determine a design factor of core component but also to indicate an improvement direction of SiC/SiC composites. Through these work, reliability and safety of GFR will be increased

  18. 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

  19. Irradiation and examination results of the AC-3 mixed-carbide test

    International Nuclear Information System (INIS)

    Mason, R.E.; Hoth, C.W.; Stratton, R.W.; Botta, F.

    1992-01-01

    The AC-3 test was a cooperative Swiss/US irradiation test of mixed-carbide, (U,Pr)C, fuel pins in the Fast Flux Test Facility. The test included 25 Swiss-fabricated sphere-pac-type fuel pins and 66 U.S. fabricated pellet-type fuel pins. The test was designed to operate at prototypical fast reactor conditions to provide a direct comparison of the irradiation performance of the two fuel types. The test design and fuel fabrication processes used for the AC-3 test are presented

  20. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydin; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  1. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States)

    2010-01-31

    An engineering code to model the irradiation behavior of UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  2. 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

  3. Performance evaluation of UO2-Zr fuel in power ramp tests

    International Nuclear Information System (INIS)

    Knudsen, P.; Bagger, C.

    1977-01-01

    In power reactors using UO 2 -Zr fuel, rapid power increases may lead to failures in fuel pins that have been irradiated at steady or decreasing heat loads. This paper presents results which extend the experience with power ramp performance of high burn-up fuel pins. A test fuel element containing both pellet and vipac UO 2 -Zr fuel pins was irradiated in the HBWR at Halden for effectively 2 1/2 years to an average burn-up of 21,000 MWD/te UO 2 at gradually decreasing power levels. The subsequent non-destructive characterization revealed formation of transverse cracks in the vipac fuel columns. After the HBWR irradiation, five of the fuel pins were power ramp tested individually in the DR 3 Reactor at Riso. The ramp rates in this test series were in the range 3-60 W/cm min. The maximum local heat loads seen in the ramp tests were 20-120% above the highest levels experienced at the same axial positions during the HBWR irradiation. Three pellets and one vipac fuel pin failed, whereas another vipac pin gave no indication of clad penetration. Profilometry after the ramp testing indicated the formation of small ridges for both types of fuel pins. For vipac fuel, the ridges were less regularly distributed along the pin length than for pellet fuel. Neutron radiography revealed the formation of additional transverse and longitudinal fuel cracks during the power ramps for both types of fuel pins. The observed failures seemed to be marginal since little or no indication as to the locations of the clad penetrations could be derived from the non-destructive post-irradiation examinations. The cases have been analyzed by means of the Danish fuel performance codes. The calculations, which are in general agreement with the observations, are discussed. The results of the investigations indicate qualitative similarities in over power performance of the two fuel types

  4. 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.

  5. 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.)

  6. Effect of interaction between irradiation-induced defects and intrinsic defects in the pinning improvement of neutron irradiated YBaCuO sample

    International Nuclear Information System (INIS)

    Topal, Ugur; Sozeri, Huseyin; Yavuz, Hasbi

    2004-01-01

    Interaction between the intrinsic (native) defects and the irradiation-induced defects created by neutron irradiation was examined for the YBCO sample. For this purpose, non-superconducting Y-211 phase was included to the Y-123 samples at different contents as a source of large pinning center. The critical current density enhancement with the irradiation for these samples were analysed and then the role of defects on pinning improvement was discussed

  7. Effect of interaction between irradiation-induced defects and intrinsic defects in the pinning improvement of neutron irradiated YBaCuO sample

    Energy Technology Data Exchange (ETDEWEB)

    Topal, Ugur; Sozeri, Huseyin; Yavuz, Hasbi

    2004-08-01

    Interaction between the intrinsic (native) defects and the irradiation-induced defects created by neutron irradiation was examined for the YBCO sample. For this purpose, non-superconducting Y-211 phase was included to the Y-123 samples at different contents as a source of large pinning center. The critical current density enhancement with the irradiation for these samples were analysed and then the role of defects on pinning improvement was discussed.

  8. An Investigation of an Irradiated Fuel Pin by Measurement of the Production of Fast Neutrons in a Thermal Column and by Pile Oscillation Technique

    International Nuclear Information System (INIS)

    Gustavsson, Veine

    1968-05-01

    A fuel pin irradiated to about 3400 MWd/tU from the Halden reactor has been investigated by a measurement of the production of fast neutrons in a thermal column and by pile oscillator technique in the central channel of the reactor R1. Calibration was made by using samples with different U 235 enrichments. The thermal column experiment gives the quantity ave(νΣ f ) (average in the thermal column spectrum) for the Halden sample. Σ f is the macroscopic fission cross section and ν is the number of fast neutrons produced per fission. The result of the oscillator measurements is a value of ave(Σ a ) - w ave(Σ f ) (average in the central channel spectrum) for the irradiated sample, w is the importance of a fast neutron relative to a thermal one and ave(Σ a ) is the macroscopic absorption cross section. The results from both the experiments have been compared with values calculated by the REBUS code and the agreement was good

  9. Measuring a ballooning gap size of irradiated fuels by the indirect method of neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Cheul Muu; Lee, Seung Wook; Lim, In Cheol; Hong, Kwang Pyo; Kim, Young Jin

    2003-11-01

    An indirect method of a neutron radiography is mobilized for inspecting post irradiated nuclear fuel pins, UO{sub 2}Si clad with Al, which swallowing, crack, ballooning, plug gap, thinning and so on, occurred. The system of an indirect method consists of a cask of carrying fuel pins, Dy converter, linear controller of converter, camera of monitoring fuel cassette. The nuclear sample pins of RISO and KAERI are exposed for 25 min. at the first exposure room, 10{sup 7} cm{sup 2}/sec flux. An activation image formed in the Dy foil is subsequently transferred in a dark room for a more than 8 hours to SR film using the decay radiation. Due to L/D ratio an unsharpness of 9.82{approx}14{mu}m and a magnification of 1.0003 are given. After digitizing an image of SR film, the ballooning gap of plug is discernible by H/V filter of image processing.

  10. Measuring a ballooning gap size of irradiated fuels by the indirect method of neutron radiography

    International Nuclear Information System (INIS)

    Sim, Cheul Muu; Lee, Seung Wook; Lim, In Cheol; Hong, Kwang Pyo; Kim, Young Jin

    2003-11-01

    An indirect method of a neutron radiography is mobilized for inspecting post irradiated nuclear fuel pins, UO 2 Si clad with Al, which swallowing, crack, ballooning, plug gap, thinning and so on, occurred. The system of an indirect method consists of a cask of carrying fuel pins, Dy converter, linear controller of converter, camera of monitoring fuel cassette. The nuclear sample pins of RISO and KAERI are exposed for 25 min. at the first exposure room, 10 7 cm 2 /sec flux. An activation image formed in the Dy foil is subsequently transferred in a dark room for a more than 8 hours to SR film using the decay radiation. Due to L/D ratio an unsharpness of 9.82∼14μm and a magnification of 1.0003 are given. After digitizing an image of SR film, the ballooning gap of plug is discernible by H/V filter of image processing

  11. 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

  12. 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

  13. Second phase precipitation in irradiated Type 316 stainless steel cladding

    International Nuclear Information System (INIS)

    Hales, J.W.

    1978-05-01

    Differences in the phase composition of FFTF fuel cans following irradiation in the General Electric Test Reactor compared to HEDL fuel cans prompted laboratory studies to be conducted using cladding from the same lots of material used to fabricate the fuel pins and on cladding sections removed from the plenum area of the irradiated fuel pins to help establish the cause of the observed differences

  14. Release of fission products from a fuel rod with an artificial hole through cladding irradiated in an in-pile water loop, (2)

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi

    1978-11-01

    To make clear the iodine spiking phenomenon from a defective fuel rod into the primary coolant, the fuel rod (UO 2 pellets, with stainless steel sheath) with an artificial pin hole was irradiated in the inpile test section of water loop JMTR.OWL-1. Experimental conditions were depressurization and temperature drop of the primary loop coolant and diameter and position of the pin hole. Iodine 131 and cesium 137 in loop coolant were measured under various coolant conditions. The inventory and translation rate of iodine 131 in fuel rod related to irradiation histories were calculated. The levels of I-131 and Cs-137 released to loop coolant from fuel rod were compared. Comparison of the results with LWRs was made by way of the spiked amount and release rate of iodine 131. (author)

  15. 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

  16. Irradiation experience with KNK II Fast Breeder Fuel Subassemblies

    International Nuclear Information System (INIS)

    Hess, B.

    1993-02-01

    During the operation of the second core of KNK II fuel pin failures occurred, which were caused by local cladding weakening due to mechanical interaction between fuel pins and pin spacers. The present report gives a summarizing presentation of the consequences of these interactions, of the experimental and theoretical investigations to clarify the reason for the interactions and of measures to reduce their consequences in the extended residence time of the second core of KNK II. This type of interaction is caused by thermo-elastic instabilities of the fuel pin bundle, and its strength depends sensitively on the geometry of the pin bundle and the pin power. Finally, measures are described, which were taken for the fuel subassemblies of the third core of KNK II to avoid the wear causing instabilities [de

  17. 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)

  18. 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

  19. 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)

  20. Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Venkiteswaran, C.N., E-mail: cnv@igcar.gov.in; Jayaraj, V.V.; Ojha, B.K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B.P.C.; Kasiviswanathan, K.V.; Jayakumar, T.

    2014-06-01

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel–clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel–clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

  1. Examination in hot laboratories of irradiated fuels from fast reactors

    International Nuclear Information System (INIS)

    Clottes, G.; Peray, R.; Ratier, J.L.

    1980-05-01

    Low irradiation rate examinations were carried out soon after the Rapsodie, Rapsodie Fortissimo and Phenix reactors were started up for the first time in order to check the level of maximum temperatures reached and the radial migration of oxygen and plutonium and to assess the movements of fuels inside the cladding. The other examinations were effected at a high specific burnup in order to defines the limit specific burnup securing the integrity of the fuel pin claddings (distortion, ruptures and possible consequences). The examinations carried out so far on fuel elements coming from Phenix or Rapsodie have allowed good fuel surveillance to be undertaken and the acquisition of a large number of data, thanks to which the fuel characteristics of future reactors of the system have been developed [fr

  2. Reusable fuel test assembly for the FFTF

    International Nuclear Information System (INIS)

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

    1992-01-01

    A fuel test assembly that provides re-irradiation capability after interim discharge and reconstitution of the test pin bundle has been developed for use in the Fast Flux Test Facility (FFTF). This test vehicle permits irradiation test data to be obtained at multiple exposures on a few select test pins without the substantial expense of fabricating individual test assemblies as would otherwise be required. A variety of test pin types can be loaded in the reusable test assembly. A reusable test vehicle for irradiation testing in the FFTF has long been desired, but a number of obstacles previously prevented the implementation of such an experimental rig. The MFF-8A test assembly employs a 169-pin bundle using HT-9 alloy for duct and cladding material. The standard driver pins in the fuel bundle are sodium-bonded metal fuel (U-10 wt% Zr). Thirty-seven positions in the bundle are replaceable pin positions. Standard MFF-8A driver pins can be loaded in any test pin location to fill the bundle if necessary. Application of the MFF-8A reusable test assembly in the FFTF constitutes a considerable cost-saving measure with regard to irradiation testing. Only a few well-characterized test pins need be fabricated to conduct a test program rather than constructing entire test assemblies

  3. Calculation of oxygen distribution in uranium-plutonium oxide fuels during irradiation (programme CODIF)

    International Nuclear Information System (INIS)

    Moreno, A.; Sari, C.

    1978-01-01

    Radial gradients of oxygen to metal ratio, O/M, in uranium-plutonium oxide fuel pins, during irradiation and at the end of life, have been calculated on the basis of solid-state thermal diffusion using measured values of the heat of transport. A detailed computer model which includes the calculation of temperature profiles and the variation of the average O/M ratio as a function of burn-up is given. Calculations show that oxygen profiles are affected by the isotopic composition of the fuel, by the temperature profiles and by fuel-cladding interactions

  4. Postirradiation examination of JOYO MK-II control rod (CRM601). Irradiation performance of shroud type absorber pin

    International Nuclear Information System (INIS)

    Tanaka, Kosuke; Kikuchi, Shin; Katsuyama, Kozo; Nagamine, Tsuyoshi; Mitsugi, Takeshi; Uto, Manabu; Tatebe, Kazuaki; Onose, Shoji; Maruyama, Tadashi

    1998-10-01

    This paper describes the results of postirradiation examination and analysis by CORAL code for irradiation performance of CRM601 control rod, which was the 6th reloaded control rod with shroud type absorber pins for use in JOYO MK-II core. The detailed visual examination indicated that there was no cladding breach in absorber pins. However, sodium ingress from the vent tube was observed in four absorber pins among seven pins. While a remarkable oval deformation occurred in cladding tube of helium bonded absorber pins, a little or no diametral change was observed in the absorber pins in which sodium ingress took place. From metallurgical observations and the analysis by CORAL code, it was estimated that the shroud tube installed in helium bonded absorber pins were irradiated at 720degC, and those in sodium bonded absorber pins were irradiated at 420degC. It was confirmed that diametral change of cladding depended on the initial gap between shroud and cladding tube. The results of present investigation indicate that it is desirable to use the materials with low thermal expansion coefficient for shroud tubes, and that sodium bonded absorber pins were advantageous for obtaining long life control rods. (author)

  5. 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.

  6. 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

  7. 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

  8. Design and Operation of 3-Pin FTL HVAC System

    International Nuclear Information System (INIS)

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

    2005-01-01

    According to the increasing demand for irradiation tests to develop new fuels, the 3-Pin FTL(Fuel Test Loop for 3 pin test fuel) facility has now been under design to conduct in-core fuel performance tests at the operating conditions, which will be installed at HANARO. The HVAC system of the FTL will be dependent on that of the HANARO. The FTL has three equipments rooms, which are the room 1, room 2 and the control room. The high pressure and high temperature equipments will be installed in the room 1. The atmosphere of the room 1 shall be maintained under the designed condition. This paper describes the design of the FTL HVAC system in the room 1

  9. 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

  10. 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

  11. 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.)

  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. Collective radiation doses following a hypothetical, very severe accident to an irradiated fuel transport flask containing AGR fuel

    International Nuclear Information System (INIS)

    Corbett, J.O.

    1985-05-01

    Studies of the consequences of very severe, although unlikely, accidents to irradiated fuel transport flasks are made in order to evaluate risks. If an irradiated fuel transport flask carrying AGR fuel were damaged in a hypothetical accident involving a severe impact followed by a prolonged fire, a small proportion of caesium and other fission products might be released to the atmosphere from the gap inventory of broken fuel pins. The consequent radiation dose to the public would arise predominantly by direct irradiation from ground deposits and the ingestion of slightly contaminated foodstuffs. Although these collective doses must generally be estimated with the aid of computer codes, it is shown here that the worst case, when a high proportion of the radioactivity is deposited in a densely population area, can be assessed approximately by a much simpler method, an approach which is of great value in explaining the calculation in a manner that can be readily understood. A comparison is made between the simple approach and equivalent results from the NECTAR code, the worst case is compared with an ensemble average over all weather conditions, and the relative contributions of the two main routes to collective dose are discussed. (author)

  14. gamma-Zr-Hydride Precipitate in Irradiated Massive delta- Zr-Hydride

    DEFF Research Database (Denmark)

    Warren, M. R.; Bhattacharya, D. K.

    1975-01-01

    During examination of A Zircaloy-2-clad fuel pin, which had been part of a test fuel assembly in a boiling water reactor, several regions of severe internal hydriding were noticed in the upper-plenum end of the pin. Examination of similar fuel pins has shown that hydride of this type is caused by...... to irradiation-induced swelling....

  15. 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

  16. 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.)

  17. Comparison of proton microbeam and gamma irradiation for the radiation hardness testing of silicon PIN diodes

    Science.gov (United States)

    Jakšić, M.; Grilj, V.; Skukan, N.; Majer, M.; Jung, H. K.; Kim, J. Y.; Lee, N. H.

    2013-09-01

    Simple and cost-effective solutions using Si PIN diodes as detectors are presently utilized in various radiation-related applications in which excessive exposure to radiation degrades their charge transport properties. One of the conventional methods for the radiation hardness testing of such devices is time-consuming irradiation with electron beam or gamma-ray irradiation facilities, high-energy proton accelerators, or with neutrons from research reactors. Recently, for the purpose of radiation hardness testing, a much faster nuclear microprobe based approach utilizing proton irradiation has been developed. To compare the two different irradiation techniques, silicon PIN diodes have been irradiated with a Co-60 gamma radiation source and with a 6 MeV proton microbeam. The signal degradation in the silicon PIN diodes for both irradiation conditions has been probed by the IBIC (ion beam induced charge) technique, which can precisely monitor changes in charge collection efficiency. The results presented are reviewed on the basis of displacement damage calculations and NIEL (non-ionizing energy loss) concept.

  18. Evaluation of the ceramographies of the KNK II/1 test zone fuel assembly NY-202-IA

    International Nuclear Information System (INIS)

    Geier, F.

    1983-12-01

    From the 211 fuel pins of the KNK II/1 fuel assembly NY-202-IA six intact fuel pins were selected in addition to the defective pin for destructive post-irradiation examinations in the Hot Cells of the KfK Karlsruhe. The assembly had been unloaded due to a pin failure after 192 equivalent full-power days and a maximum burnup of 5.4 %. The main aspect of these investigations was to record the fuel and fuel pin behavior and thus to allow a comparison of the status before and after irradiation. The results can also be used for comparative calculations and adaptations of existing calculational models. This report documents in detailed form the results of the fuel and fuel pin examinations [de

  19. An Investigation of an Irradiated Fuel Pin by Measurement of the Production of Fast Neutrons in a Thermal Column and by Pile Oscillation Technique

    Energy Technology Data Exchange (ETDEWEB)

    Gustavsson, Veine

    1968-05-15

    A fuel pin irradiated to about 3400 MWd/tU from the Halden reactor has been investigated by a measurement of the production of fast neutrons in a thermal column and by pile oscillator technique in the central channel of the reactor R1. Calibration was made by using samples with different U 235 enrichments. The thermal column experiment gives the quantity ave({nu}{sigma}{sub f}) (average in the thermal column spectrum) for the Halden sample. {sigma}{sub f} is the macroscopic fission cross section and {nu} is the number of fast neutrons produced per fission. The result of the oscillator measurements is a value of ave({sigma}{sub a}) - w ave({sigma}{sub f}) (average in the central channel spectrum) for the irradiated sample, w is the importance of a fast neutron relative to a thermal one and ave({sigma}{sub a}) is the macroscopic absorption cross section. The results from both the experiments have been compared with values calculated by the REBUS code and the agreement was good.

  20. 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

  1. Comparison of Calculated and Measured Neutron Fluence in Fuel/Cladding Irradiation Experiments in HFIR

    International Nuclear Information System (INIS)

    Ellis, Ronald James

    2011-01-01

    permanent reflector. It is surrounded by a water reflector of effectively infinite thickness. In the axial direction, the reactor is reflected by water above and below the reactor. The irradiation facilities, one for UN and the other for UO 2 pellets, utilize a thin cylindrical hafnium shield approximately 4 cm in diameter surrounding the facility basket to reduce the thermal neutron flux sufficiently such that the linear power rating in the irradiated fuel pins will be similar to PWR operating conditions. The facilities each contain nine fuel pins, each comprising 10 fuel pellets, arranged as if three fuel rods.

  2. Advanced fuel cycle on the basis of pyroelectrochemical process for irradiated fuel reprocessing and vibropacking technology

    International Nuclear Information System (INIS)

    Mayorshin, A.A.; Skiba, O.V.; Tsykanov, V.A.; Golovanov, V.N.; Bychkov, A.V.; Kisly, V.A.; Bobrov, D.A.

    2000-01-01

    For advanced nuclear fuel cycle in SSC RIAR there is developed the pyroelectrochemical process to reprocess irradiated fuel and produce granulated oxide fuel UO 2 , PuO 2 or (U,Pu)O 2 from chloride melts. The basic technological stage is the extraction of oxides as a crystal product with the methods either of the electrolysis (UO 2 and UO 2 -PuO 2 ) or of the precipitating crystalIization (PuO 2 ). After treating the granulated fuel is ready for direct use to manufacture vibropacking fuel pins. Electrochemical model for (U,Pu)O 2 coprecipitation is described. There are new processes being developed: electroprecipitation of mixed oxides - (U,Np)O 2 , (U,Pu,Np)O 2 , (U,Am)O 2 and (U,Pu,Am)O 2 . Pyroelectrochemical production of mixed actinide oxides is used both for reprocessing spent fuel and for producing actinide fuel. Both the efficiency of pyroelectrochemical methods application for reprocessing nuclear fuel and of vibropac technology for plutonium recovery are estimated. (author)

  3. 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

  4. Fuel sub-assembly

    International Nuclear Information System (INIS)

    Jolly, R.

    1982-01-01

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

  5. 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

  6. 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

  7. Development of an End-plug Welding Technology for an Instrumented Fuel Irradiation Test

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Lee, Chul Yong; Shin, Yoon Taek; Choo, Kee Nam

    2010-01-01

    The irradiation test of end-plug specimens was planned for the evaluation of nuclear fuels performance. To establish the fabrication process, and for satisfying the requirements of the irradiation test, an orbital-GTA weld machine for the specimens of the dual rods was developed, and the preliminary welding experiments for optimizing the process conditions of the specimens of the dual rods were performed. Dual rods with a 9.5mm diameter and a 0.6mm wall thickness of the cladding tubes and end-plugs have been used and the optimum conditions of the pin-hole welding have also been selected. This paper describes the experimental results of the GTA welds of the specimens of the dual rods and the metallography examinations of the GTA welded specimens for various welding conditions for the instrumented fuel irradiation test. These investigations satisfied the requirements of the instrumented irradiation test and the GTA welds for the specimens of the dual rods at the HANARO research reactor

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

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

    International Nuclear Information System (INIS)

    Boussard, F.; Huillery, R.

    1994-01-01

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

  11. Means for supporting nuclear fuel

    International Nuclear Information System (INIS)

    Cocker, P.; Price, M.A.

    1975-01-01

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

  12. 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

  13. 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

  14. 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

  15. 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

  16. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    International Nuclear Information System (INIS)

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-01-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at ∼2.4, ∼7 and ∼11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of ∼7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10 15 n/cm 2 /s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between ∼410 deg. C and ∼645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  17. Fabrication and testing of uranium nitride fuel for space power reactors

    Science.gov (United States)

    Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

    1988-02-01

    Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

  18. 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

  19. The development of fuel pins and material specimens mixed loading irradiation test rig in the experimental fast reactor Joyo. The development of the fuel-material hybrid rig

    International Nuclear Information System (INIS)

    Oyamatsu, Yasuko; Someya, Hiroyuki

    2013-02-01

    In the experimental fast reactor Joyo, there were many tests using the irradiation rigs that it was possible to be set irradiation conditions for each compartment independently. In case of no alternative fuel element to irradiate after unloading the irradiated compartments, the irradiation test was restarted with the dummy compartment which the fuel elements was not mounted. If the material specimens are mounted in this space, it is possible to use the irradiation space effectively. For these reasons, the irradiation rig (hybrid rig) is developed that is consolidated with material specimens compartment and fuel elements compartment. Fuel elements and material specimens differ greatly with heat generation, so that the most important issue in developing of hybrid rig is being able to distribute appropriately the coolant flow which satisfies irradiation conditions. The following is described by this report. (1) It was confirmed that the flow distribution of loading the same irradiation rig with the compartment from which a flow demand differs could be satisfied. (2) It was confirmed that temperature setting range of hybrid rig could be equivalent to that of irradiation condition. (3) By standardizing the coolant entrance structure of the compartment lower part, the prospect which can perform easily recombination of the compartment from which a type differs between irradiation rigs was acquired. (author)

  20. Post irradiation test report of irradiated DUPIC simulated fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

    2001-12-01

    The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are γ-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

  1. Development of a FBR fuel bundle-duct interaction analysis code-BAMBOO. Analysis model and verification by Phenix high burn-up fuel subassemblies

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Ito, Masahiro; Ukai, Shigeharu

    2005-01-01

    The bundle-duct interaction analysis code ''BAMBOO'' has been developed for the purpose of predicting deformation of a wire-wrapped fuel pin bundle of a fast breeder reactor (FBR). The BAMBOO code calculates helical bowing and oval-distortion of all the fuel pins in a fuel subassembly. We developed deformation models in order to precisely analyze the irradiation induced deformation by the code: a model to analyze fuel pin self-bowing induced by circumferential gradient of void swelling as well as thermal expansion, and a model to analyze dispersion of the orderly arrangement of a fuel pin bundle. We made deformation analyses of high burn-up fuel subassemblies in Phenix reactor and compared the calculated results with the post irradiation examination data of these subassemblies for the verification of these models. From the comparison we confirmed that the calculated values of the oval-distortion and bowing reasonably agreed with the PIE results if these models were used in the analysis of the code. (author)

  2. 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

  3. 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.

  4. Irradiation performance and post-irradiation examinations of the instrumented sphere-pac UO2 assembly IFA-204, irradiated up to 1.7% FIMA in the Halden Boiling Water Reactor

    International Nuclear Information System (INIS)

    Linde, A. van der.

    1982-12-01

    Fuel assembly IFA-204 consisted of four pins filled with a blend of three sizes of UO 2 spheres. This blend was compacted in the zircaloy-4 cladding tube by vibration to achieve a 87-88% T.D. smear density of the 1500 mm long sphere-pac fuel column. IFA-204 was irradiated during 597 days, equivalent to 436 full power days, in the Halden BWR from November 1971 to April 1974 when the achieved assembly average burnup amounted to 13.7 MWd/kg UO 2 . Two of the four pins were equipped with fuel column length and pin length extensometers. The measurements showed that at average pin powers less than about 25 kW.m -1 only fuel-clad-thermal-interaction, FCTI, occurred. The clad thermal expansion coefficient was 4.8 ppm/kW.m -1 . For calculations of the irradiation behaviour of sphere-pac LWR pins, filled with 0.1 MPa helium, with the Gapcon-Thermal-2 code a new thermal conductivity-temperature relationship has been developed. The calculated data agreed reasonably well with the measured data when taking into account a restructured fuel density of 90% T.D., a fuel surface roughness of 1 μm, the absence of a fuel-cladding gap and an effective full power fuel restructuring time of 40 days. It is concluded that the porous structure of the sphere-pac fuel column, whether restructured or not, has the inherent disadvantage of a high mobility of gaseous fission products and the inherent advantage of a practically stress free operation of the cladding

  5. 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

  6. HANARO fuel irradiation test(II)

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, D. S.; Kim, H. R.; Chae, H. T.; Lee, B. C.; Lee, C. S.; Kim, B. G.; Lee, C. B.; Hwang, W

    2001-04-01

    In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiatied at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%.

  7. 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)

  8. Structure and flux pinning properties of irradiation defects in YBa2Cu3O7-x

    International Nuclear Information System (INIS)

    Kirk, M.A.

    1992-06-01

    We review our investigations of defects produced in YBa 2 Cu 3 O 7-x by various forms of irradiation. The defect microstructure has been studied by transmission electron microscopy (TEM). Irradiation enhancements of flux pinning have been studied by SQUID magnetometry on single crystals. In many cases the same single crystals were used in both TEM and SQUID investigations. The primary atom recoil spectra for all the irradiations studied have been carefully calculated and used to correlate the TEM and magnetization results for the different types of irradiation. Correlation of annealing experiments, employing both TEM and SQUID measurements, among several types of irradiation has also yielded information on the different defect structures present. Defect densities, sizes and strain field anisotropies have been determined by TEM. Defect flux pinning anisotropies have been determined for two field orientations in twinned single crystals. The temperature dependences of the flux pinning have been measured. The maximum field of irreversibility at 70 K is shown to change markedly upon both neutron and proton irradiations in some crystals and not others. The defect structure, chemistry and location in the unit cell has been determined in some cases. Some interaction with existing defect structure has been observed in proton and electron irradiations. The damage character and directionality has been determined in GeV ion irradiated crystals

  9. 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.)

  10. Chemical interaction between (Cs-Te) doped fuels and cladding material under irradiation

    International Nuclear Information System (INIS)

    Delbrassine, A.; Flipot, A.J.

    1977-01-01

    Pins containing UO 2 -30 wt.% PuO 2 low density pellets and or caesium and or tellurium as doping elements have been irradiated for about 40 days in the BR 2 reactor. The effect of two Cs/Te ratios, namely 1.3 and 4, and a wide range of O/M ratios on the inner corrosion of the clad has been investigated. The influence of tellurium on the attack of the cladding has been pointed out. It may be responsible for the chromium and nickel depletion in the grain boundaries of the steal. The corrosion patterns and the thickness of the corroded layer could be different in the total length of a fuel pin. It seems therefore necessary to measure the effective Cs/Te ratio associated with the local corrosion layers. This local Cs/Te ratio should be more useful than the initial mean Cs/Te ratio in a pin for understanding the corrosion phenomena. (author)

  11. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States)

    2010-01-31

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium

  12. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydin; Buongiorno, Jacopo

    2010-01-01

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

  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. Re-irradiation and limit testing of the fuels PWR type reactors

    International Nuclear Information System (INIS)

    Roche, M.; Molvault, M.

    1978-01-01

    In view of investigating the neutron radiation behavior of PWR fuel pins, the S.P.S. (Services des Piles de Saclay) developed a set of experimental means used at OSIRIS in Saclay Nuclear Research Center. Said devices are shown to be able to meet present problems concerning can failures, power and temperature cycling, remote-control studies. These means can also be used either for statistical studies, they can then receive several samples, or for analytical studies in instrumented devices of large capacity and accelerated irradiation rate [fr

  15. Main results and status of the development of LEU fuel for Russian research reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Morozov, A.; Suprun, V.; Dobrikova, I.

    2005-01-01

    VNIINM develops low enrichment uranium (LEU) fuel on base U-Mo alloys and a novel design of pin-type fuel elements. The development is carried out both for existing reactors, and for new advanced designs of reactors. The work is carried on the following main directions: - irradiate LEU U-Mo dispersion fuel (the uranium density up to 6,0 g/cm 3 ) in two Russian research reactors: MIR (RIAR, Dimitrovgrad) as pin type fuel mini-elements and in WWR-M (PINP, Gatchina) within full-scaled fuel assembly (FA) with pin type fuel elements; - finalize development of design and fabrication process of IRT type FA with pin type fuel elements; - develop methods of reducing of U-Mo fuel --Al matrix interaction under irradiation; - develop fabricating methods of fuel elements on base of monolithic U-Mo fuel. The paper generally reviews the results of calculation, design and technology investigations accomplished by now. (author)

  16. Pin clad strains in Phenix

    International Nuclear Information System (INIS)

    Languille, A.

    1979-07-01

    The Phenix reactor has operated for 4 years in a satisfactory manner. The first 2 sub-assembly loadings contained pins clad in solution treated 316. The principal pin strains are: diametral strain (swelling and irradiation creep), ovality and spiral bending of the pin (interaction of wire and pin cluster and wrapper). A pin cluster irradiated to a dose of 80 dpa F reached a pin diameter strain of 5%. This strain is principally due to swelling (low fission gas pressure). The principal parameters governing the swelling are instantaneous dose, time and temperature for a given type of pin cladding. Other types of steel are or will be irradiated in Phenix. In particular, cold-worked titanium stabilised 316 steel should contribute towards a reduction in the pin clad strains and increase the target burn-up in this reactor. (author)

  17. Magnetic studies of current conduction and flux pinning in high-Tc cuprates: Virgin, irradiated, and oxygen deficient materials

    International Nuclear Information System (INIS)

    Thompson, J.R.; Civale, L.; Marwick, A.D.; Holtzberg, F.

    1992-09-01

    To increase the current density and pinning of magnetic flux in high temperature superconductors, defects with point-like and line-like geometries were created in controlled numbers using ion irradiation methods. Single crystals of Y 1 Ba 2 Cu 3 O 7 and Bi 2 Sr 2 Ca 1 Cu 2 0 8 superconductors were studied using dc magnetic methods. These studies showed greatly increased irreversibility in the vortex state magnetization and enhanced intragrain current density J c following irradiation. Linear defects, created by irradiation with energetic heavy ions, are particularly effective in pinning vortices at higher temperatures and magnetic fields. Further investigations of ''flux creep'' (the time dependence of magnetization) are well described by recent vortex glass and collective pinning theories. Complementary investigations have delineated the role of oxygen deficiency δ on pinning in aligned Y 1 Ba 2 Cu 3 O 7-δ materials

  18. An Analysis of the Thermal and Structure Behaviour of the UO{sub 2}-PuO{sub 2}-Fuel in the Irradiation Experiment of the UO{sub 2}-PuO{sub 2}-Fuel in the Irradiation Experiment FR2 Capsule Test Series 5a; Analisis termico y estructural del combustible UO{sub 2}-PuO{sub 2} irradiado en el reactor FR2 dentro del experimento KVE-Vg.5a

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.; Helmut, E.

    1981-07-01

    In the Karlsruhe research reactor FR2 nine fuel pins were irradiated within three irradiation capsules in the course of the test series 5a. The pins contained UO{sub 2}-PuO{sub 2} fuel pellets. They reached bump values of about 6, 17 and 47 Mwd/Kg Me with linear rod powers of 400 to 600 W/cm and clad surface temperature between 500 and 700 degree centigree. A detailed analysis of the fuel structuration data (columnar-grain and equiaxed- -grain growth regions) have allowed to determine, with the help of physic-mathematical models, the radii of these regions and the heat transfer through the contact zone between fuel and clad depending on the bump. The results of the analysis showed that the fuel surface temperature rose with increasing burnup. (Author) 16 refs.

  19. 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

  20. 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.

  1. 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.

  2. Interpretation of the CABRI-RAFT LTX test up to pin failure based on detailed data evaluation and PARAS-2S code analysis

    International Nuclear Information System (INIS)

    Fukano, Yoshitaka; Sato, Ikken

    2001-09-01

    The CABRI-RAFT LTX test aims at a study on the fuel-pin-failure mechanism, in-pin fuel motion and post-failure fuel relocation with an annular fuel pin which was pre-irradiated up to peak burn-up of 6.4 at%. The transient test conditions similar to those of the LT4 test were selected in the LTX test using the same type of fuel pin, allowing an effective direct comparison between the two tests. In contrast to the LT4 test which showed a large PCMI-mitigation potential of the annular fuel-pin design, early pin failure occurred in the LTX test when fuel does not seem to have molten. In order to clarify the fuel pin failure mechanism, interpretation of the LTX test up to pin failure is performed in this study, through an experimental data evaluation and a PAPAS-2S-code analysis. The PAPAS-2S code simulates reasonably the fuel thermal conditions such as transient fuel-pin heat-up and fuel melting. The present detailed data evaluation shows that the earlier cladding failure compared with the LT4 test is mainly attributed to the local cladding heat-up. Under the high-temperature condition, plenum gas pressure has a certain potential to explain the observed failure. Fuel swelling-induced PCMI does not seem significant in the LTX test and it may have contributed to the early pin failure only to a limited extent, if any. (author)

  3. 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.)

  4. Control of domain wall pinning by localised focused Ga + ion irradiation on Au capped NiFe nanowires

    International Nuclear Information System (INIS)

    Burn, D. M.; Atkinson, D.

    2014-01-01

    Understanding domain wall pinning and propagation in nanowires are important for future spintronics and nanoparticle manipulation technologies. Here, the effects of microscopic local modification of the magnetic properties, induced by focused-ion-beam intermixing, in NiFe/Au bilayer nanowires on the pinning behavior of domain walls was investigated. The effects of irradiation dose and the length of the irradiated features were investigated experimentally. The results are considered in the context of detailed quasi-static micromagnetic simulations, where the ion-induced modification was represented as a local reduction of the saturation magnetization. Simulations show that domain wall pinning behavior depends on the magnitude of the magnetization change, the length of the modified region, and the domain wall structure. Comparative analysis indicates that reduced saturation magnetisation is not solely responsible for the experimentally observed pinning behavior.

  5. The design of in-pile test section for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, K. N.; Lee, J. M.; Shim, B. S.; Zee, D. Y.; Park, S. H.; Ahn, S. H.; Lee, J. Y.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    As an equipment for nuclear fuel's general performance irradiation test in HANARO, Fuel Test Loop(FTL) has been developed that can irradiate the pin to the maximum number of 3 at the core irradiation hole(IR1 hole) by considering for it's utility and user's irradiation requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). IPS consists for IPS Vessel assembly, In-Pool Piping, IPS Support, In-Pool Piping Support etc. Design that such IPS considers interference item consisted to do not bear in existing facilities by one. IVA that is connected to the OPS are controlled and regulated by means of system pressure, system temperature and the water quality. IPS Vessel assembly is consisted of outer pressure vessel, inner pressure vessel, IPS head, inner assembly and test fuel carrier. After 3-Pin FTL development which is expected to be finished by the 2006, FTL will be used for the irradiation test of the new PWR-type fuel and can maximize the usage of HANARO.

  6. Development of Melting Crucible Materials of Metallic Fuel Slug for SFR

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  7. HANARO fuel irradiation test (II): revision

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, D. S.; Kim, H.; Chae, H. T.; Lee, C. S.; Kim, B. G.; Lee, C. B

    2001-04-01

    In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiated at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%. This report is the revision of KAERI/TR-1816/2001 on the irradiation test for HANARO fuel.

  8. Plutonium and minor actinide transmutation by long irradiation in LWR

    International Nuclear Information System (INIS)

    Facchini, A.; Sanjust, V.

    1993-01-01

    An investigation was made on the conceptual possibility of burning in a thermal reactor MOX fuel together with special pins containing plutonium, minor actinides and long lived fission products, recovered from the reprocessing of previously irradiated MOX fuel and mixed with an inter matrix. Preliminary calculations showed that the long term radiotoxicity of the above special pins is reduced to reasonable levels when they are irradiated up to 20 divided-by 30 years, and cooled for some centuries. In particular, during the whole life such a reactor should be able to burn a considerable fraction of plutonium, minor actinides and long lived fission products recovered from the MOX fuel irradiated along the same period of time

  9. Fuel rod computations. The COMETHE code in its CEA version

    International Nuclear Information System (INIS)

    Lenepveu, Dominique.

    1976-01-01

    The COMETHE code (COde d'evolution MEcanique et THermique) is intended for computing the irradiation behavior of water reactor fuel pins. It is concerned with steadily operated cylindrical pins, containing fuel pellet stacks (UO 2 or PuO 2 ). The pin consists in five different axial zones: two expansion chambers, two blankets, and a central core that may be divided into several stacks parted by plugs. As far as computation is concerned, the pin is divided into slices (maximum 15) in turn divided into rings (maximum 50). Information are obtained for each slice: the radial temperature distribution, heat transfer coefficients, thermal flux at the pin surface, changes in geometry according to temperature conditions, and specific burn-up. The physical models involved take account for: heat transfer, fission gas release, fuel expansion, and creep of the can. Results computed with COMETHE are compared with those from ELP and EPEL irradiation experiments [fr

  10. 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)

  11. Panoramic irradiator dose mapping with pin photodiodes

    International Nuclear Information System (INIS)

    Ferreira, Danilo Cardenuto; Napolitano, Celia Marina; Bueno, Carmen Cecilia

    2011-01-01

    In this work we study the possibility of using commercial silicon PIN photodiodes (Siemens, SFH 00206) for dose mapping in the Panoramic Irradiator facility at IPEN-CNEN/SP. The chosen photodiode, that is encased in 1.2 mm thickness polymer layer, displays promising dosimetric characteristics such as small size (sensitive area of 7.00 mm 2 ), high sensitivity and low dark current (≅ 300 pA, at 0 V) together with low-cost and wide availability. The Panoramic facility is an irradiator Type II with absorbed dose certificated by International Dose Assurance Service (IDAS) offered by the International Agency Energy Atomic (IAEA). The charge registered by the diode as a function of the absorbed dose was in excellent agreement with that one calibrated by IDAS. Besides this, the easy handling and fast response of the SFH00206 diode compared to Fricke chemical dosimeters encouraged us to perform dose mapping around the source. (author)

  12. 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

  13. Impact of fission gas on irradiated PWR fuel behaviour at extended burnup under RIA conditions

    International Nuclear Information System (INIS)

    Lemoine, F.; Schmitz, F.

    1996-01-01

    With the world-wide trend to increase the fuel burnup at discharge of the LWRs, the reliability of high burnup fuel must be proven, including its behaviour under energetic transient conditions, and in particular during RIAs. Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup. The potential for swelling and transient expansion work under rapid heating conditions characterizes the high burnup fuel behaviour by comparison to fresh fuel. This effect is resulting from the steadily increasing amount of gaseous and volatile fission products retained inside the fuel structure. An attempt is presented to quantify the gas behaviour which is motivated by the results from the global tests both in CABRI and in NSRR. A coherent understanding of specific results, either transient release or post transient residual retention has been reached. The early failure of REP Na1 with consideration given to the satisfactory behaviour of the father rod of the test pin at the end of the irradiation (under load follow conditions) is to be explained both by the transient loading from gas driven fuel swelling and from the reduced clad resistance due to hydriding. (R.P.)

  14. Influence of various stresses on diametral and axial plastic deformations of the Phenix reactor fuel cans

    International Nuclear Information System (INIS)

    Guerin, Y.; Boutard, J.L.

    1983-04-01

    Dimensions of fuel cans are modified during irradiation in fast reactors: diameter increase is produced by steel swelling and irradiation creep under the pressure of fission gases and length increase integrates swelling. Diameter and density measured on fuel cans in SS 316, irradiated in the Phenix reactor, show that interaction spacer-can and interaction between pins produce plastic deformations. The interaction spacer-can leads not only to a helical deflection of the pin but also a slight axial plastic compression associated to a diametral plastic deformation. There is also a leveling of elongation in these strained pins because of friction with neighbouring pins [fr

  15. Seven pin bundle fast top tests L01 and L02

    International Nuclear Information System (INIS)

    Davies, A.L.; Bowen, G.R.; Herbert, R.; Kear, K.L.; Tylka, J.P.; Holland, J.W.

    1984-01-01

    Tests L01 and L02 were the first two seven pin bundle tests in the PFR/TREAT program of fuel failure tests carried out jointly by the US and the UK. The two tests were on bottom plenum annular pellet mixed oxide fuel clad in 316 stainless steel. L01 used fresh fuel, while L02 used PFR irradiated 4% burn-up fuel, to determine any differences in the failure mechanism and subsequent fuel behavior due to irradiation. They were performed in flowing sodium in the Mark IIIA version of a TREAT integral loop. Both were fast transient overpower (TOP) tests intended to simulate 5 $/s reactivity ramp hypothetical accidents in a large fast reactor. The test objectives were to obtain information on fuel motion in the central hole before failure, the time and location of cladding failures, and material motion in the channel after failure, having particular regard to the effect of irradiation

  16. Fuel integrity project: analysis of light water reactor fuel rods test results

    International Nuclear Information System (INIS)

    Dallongeville, M.; Werle, J.; McCreesh, G.

    2004-01-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  17. Fuel integrity project: analysis of light water reactor fuel rods test results

    Energy Technology Data Exchange (ETDEWEB)

    Dallongeville, M.; Werle, J. [COGEMA Logistics (AREVA Group) (France); McCreesh, G. [BNFL Nuclear Sciences and Technology Services (United Kingdom)

    2004-07-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  18. Performance testing of refractory alloy-clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

    1985-01-01

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  19. Performance of a sphere-pac mixed carbide fuel pin irradiated in the Dounreay Fast Reactor (DFR 527/1 experiment)

    International Nuclear Information System (INIS)

    Bischoff, K.; Smith, L.; Stratton, R.W.

    1980-10-01

    The DFR 527/1 experiment was the first irradiation of EIR sphere-pac uranium-plutonium mixed carbide fuel in a fast flux. The experiment has been successfully irradiated to a burn-up of 7.3% FIMA at ratings between 45 and 62 kW m - 1 and clad temperatures between 300 and 600 0 C. Restructuring and elemental redistribution has been found to be similar to the pattern established for pellet type fuel and follows effects seen in earlier sphere-pac carbide tests. Gas release of 12-14% has been measured. A preliminary comparison of radial temperature distribution calculations using a first version of the fuel behaviour modelling code SPECKLE with the actual metallography has been attempted. (Auth.)

  20. 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)

  1. Analysis of the porosity distribution of mixed oxide pins; Analisis de distribucion de porosidad en barras combustibles de oxidos mixtos bajo irradiacion

    Energy Technology Data Exchange (ETDEWEB)

    Lieblich, M; Lopez, J

    1987-07-01

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

  2. 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.

  3. 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)

  4. 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)

  5. 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

  6. 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)

  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. 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

  9. 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...

  10. Post-irradiation examination and R and D programs using irradiated fuels at KAERI

    International Nuclear Information System (INIS)

    Chun, Yong Bum; Min, Duck Kee; Kim, Eun Ka and others

    2000-12-01

    This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

  11. Post-irradiation examination and R and D programs using irradiated fuels at KAERI

    International Nuclear Information System (INIS)

    Chun, Yong Bum; So, Dong Sup; Lee, Byung Doo; Lee, Song Ho; Min, Duck Kee

    2001-09-01

    This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

  12. 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

  13. 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

  14. Plutonium fuel program

    International Nuclear Information System (INIS)

    1979-09-01

    A review is presented of the development of the (UPu)C sphere-pac fuel project during 1978. In particular, the problems encountered in obtaining good fuel quality in the fabrication process and their solution is discussed. The development of a fabrication pilot plant is considered, and the post-irradiation examination of fuel pins is presented. (Auth.)

  15. Materials specific work at Forschungszentrum Karlsruhe and in cooperation with the industrial partners ALKEM and Interatom for the development of nuclear oxide fuels for fission reactors

    International Nuclear Information System (INIS)

    Kleykamp, H.; Muehling, G.

    2005-09-01

    The fabrication of uranium-plutonium oxide fuel started in Forschungszentrum Karlsruhe and at ALKEM company to begin for the criticality experiments in the SNEAK reactor and subsequently for stationary fuel pin irradiations in the FR2, BR2, DFR, Rapsodie, Phenix and KNK II reactors. The production methods comprised first the mechanical blending of UO2 and PuO2 followed by direct pressing and sintering of the pellets, later the advanced methods such as optimized comilling and ammonium uranyl plutonyl coprecititation. The fabrication of pellets was described in the main, further the alternative fuel pin manufacturing processes by vibrational compaction and hot-impact densification were discussed. The first capsule and pin irradiations in the FR2 and BR2 reactors contributed to the assessment of the maximum operation parameters within the fuel pin development such as linear heat rating, cladding temperature and burnup. Subsequently, small-bundle and largebundle irradiations were made in fast reactors in cooperation with Interatom company in order to verify the specifications for the commercial fast reactor SNR 300. Milestones were the maximum burnup of 175 GWd/t metal, corresponding 18.6 % of the heavy atoms, obtained in one of the KNK II fuel pin assemblies, and the displacement rates in the cladding materials of 140 dpa NRT attained in the Phenix reactor. Higher implications gained later the stationary irradiations of defected mixed-oxide pins, the mild fuel pin transient operations, the local blockage experiments and the severe hypothetic accidents in the respective Siloe, HFR, BR2 and CABRI reactors. These experiments were made solely in international partnership. Further activities were the chemical analyses of solid residues and coprecipitations of irradiated mixed-oxide fuels in the head-end of the reprocessing. All these actions were coordinated in the then fast breeder project. Furthermore, irradiated fuels and fuel pins of other reactor types were

  16. Fuel irradiation experience at Halden

    International Nuclear Information System (INIS)

    Vitanza, Carlo

    1996-01-01

    The OECD Halden Reactor Project is an international organisation devoted to improved safety and reliability of nuclear power station through an user-oriented experimental programme. A significant part of this programme consists of studies addressing fuel performance issues in a range of conditions realised in specialised irradiation. The key element of the irradiation carried out in the Halden reactor is the ability to monitor fuel performance parameters by means of in-pile instrumentation. The paper reviews some of the irradiation rigs and the related instrumentation and provides examples of experimental results on selected fuel performance items. In particular, current irradiation conducted on high/very high burn-up fuels are reviewed in some detail

  17. 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

  18. 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.

  19. Neutron Flux Depression in the UO2-PuO2 (15 to 30%) Fuel Rods from IVO-FR2-Vg7-Irradiation Experiment

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Fernandez Marron, J.L.

    1983-01-01

    The thermal-neutron flux depression within a fuel rod has a great influence in the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO 2 -PUO 2 (15 to 30% PUO 2 ) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (Author) 22 refs

  20. Neutron flux depression in the UO2-PuO2 (15 to 30%) fuel rods from IVO-FR2-Vg7-Irradiation experiment

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Fernandez Marron, J.L.

    1983-01-01

    The thermal-neutron flux depression within a fuel rod has a great influence on the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO 2 -PuO 2 (15 to 30% PuO 2 ) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (author)

  1. Fuel-cladding chemical interaction in mixed-oxide fuels

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Weber, J.W.; Devary, J.L.

    1978-10-01

    The character and extent of fuel-cladding chemical interaction (FCCI) was established for UO 2 -25 wt% PuO 2 clad with 20% cold worked Type 316 stainless steel irradiated at high cladding temperatures to peak burnups greater than 8 atom %. The data base consists of 153 data sets from fuel pins irradiated in EBR-II with peak burnups to 9.5 atom %, local cladding inner surface temperatures to 725 0 C, and exposure times to 415 equivalent full power days. As-fabricated oxygen-to-metal ratios (O/M) ranged from 1.938 to 1.984 with the bulk of the data in the range 1.96 to 1.98. HEDL P-15 pins provided data at low heat rates, approx. 200 W/cm, and P-23 series pins provided data at higher heat rates, approx. 400 W/cm. A design practice for breeder reactors is to consider an initial reduction of 50 microns in cladding thickness to compensate for possible FCCI. This approach was considered to be a conservative approximation in the absence of a comprehensive design correlation for extent of interaction. This work provides to the designer a statistically based correlation for depth of FCCI which reflects the influences of the major fuel and operating parameters on FCCI

  2. SATURN-FS 1: A computer code for thermo-mechanical fuel rod analysis

    International Nuclear Information System (INIS)

    Ritzhaupt-Kleissl, H.J.; Heck, M.

    1993-09-01

    The SATURN-FS code was written as a general revision of the SATURN-2 code. SATURN-FS is capable to perform a complete thermomechanical analysis of a fuel pin, with all thermal, mechanical and irradiation-based effects. Analysis is possible for LWR and for LMFBR fuel pins. The thermal analysis consists of calculations of the temperature profile in fuel, gap and in the cladding. Pore migration, stoichiometry change of oxide fuel, gas release and diffusion effects are taken into account. The mechanical modeling allows the non steady-state analysis of elastic and nonelastic fuel pin behaviour, such as creep, strain hardening, recovery and stress relaxation. Fuel cracking and healing is taken into account as well as contact and friction between fuel and cladding. The modeling of the irradiation effects comprises swelling and fission gas production, Pu-migration and irradiation induced creep. The code structure, the models and the requirements for running the code are described in the report. Recommendations for the application are given. Program runs for verification and typical examples of application are given in the last part of this report. (orig.) [de

  3. 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.

  4. 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

  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. Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.

    1985-01-01

    A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

  7. Results of the post-irradiation examination of a highly-rated mixed oxide fuel rod from the Mol 7B experiment

    International Nuclear Information System (INIS)

    Coquerelle, M.; Walker, C.T.; Whitlow, W.H.

    1980-01-01

    The experiment MOL 7B was carried out in a epithermal flux in the Belgian reactor BR2. The pin examined contained fuel of initial composition (Usub(0.7)Pusub(0.3))Osub(1.98). It had been irradiated to a maximum burn-up of 13.2 at% at a maximum linear power of 568Wcm -1 . The fuel was clad with coldworked stainless steel. Electron microprobe analysis indicated that a Cr 2 O 3 type oxide was the main constituent of the grey phases in the gap. A metallic phase on the fuel surface had apparently resulted from the mechanical compaction of fragments of cladding that had been depleted in chromium by oxidation. Thus the main components of the phase were iron and nickel. Chromium loss from the inner cladding surface was significant only in the upper regions of the pin. In pin sections that were metallographically examined sigma phase and carbides of the type M 23 C 6 were present at the grain boundaries of the cladding. Cladding corrosion was not an Arrhenius function of the cladding temperature: the amount of metal lost from the inner cladding surface decreased with rise in cladding temperature above 910 K. A contributor to metal loss was the mechanical detachment of fragments of cladding which reformed as a metallic layer on the surface of the fuel. Chromium depletion and sigma phase formation at grain boundaries lowered the cohesive forces between grains which were then mechanically detached. Chromium loss from grain boundaries is mainly the results of oxidation of the cladding by the mixed oxide fuel. Data are presented to support the view that the local average O/M of the fuel determined the rate of oxidation and consequently the extent of chromium depletion. Fuel-cladding mechanical interactions were weak in the upper regions of the pin where metal loss was small

  8. 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.)

  9. 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

  10. Blind prediction exercise on modeling of PHWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Sah, D.N.; Viswanathan, U.K.; Viswanadham, C.S.; Unnikrishnan, K.; Rath, B.N.

    2008-01-01

    A blind prediction exercise was organised on Indian Pressurised Heavy Water Reactor (PHWR) fuel to investigate the predictive capability of existing codes for their application at extended burnup and to identify areas of improvement. The blind problem for this exercise was based on a PHWR fuel bundle irradiated in Kakrapar Atomic Power Station-I (KAPS-I) up to about 15 000 MWd/tU and subjected to detailed post-irradiation examination (PIE) in the hot cells facility at BARC. Eleven computer codes from seven countries participated in this exercise. The participants provided blind predictions of fuel temperature, fission gas release, internal gas pressure and other performance parameters for the fuel pins. The predictions were compared with the experimental PIE data which included fuel temperature derived from fuel restructuring, fission gas release measured by fuel pin puncturing, internal gas pressure in pin, cladding oxidation and fuel microstructural data. The details of the blind problem and an analysis of the results of blind predictions by the codes vis-a-vis measured data are provided in this paper

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

    Indian Academy of Sciences (India)

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

  12. Magnetization and flux pinning in high-Tc cuprates: Irradiated and oxygen deficient materials

    International Nuclear Information System (INIS)

    Thompson, J.R.; Civale, L.; Marwick, A.D.; Holtzberg, F.

    1992-11-01

    This work surveys recent studies of the intragrain current density J and vortex pinning in high Tc superconductors. Materials include Y 1 Ba 2 Cu 3 O 7-δ and Bi 2 Sr 2 Ca 1 Cu 2 O 8 single crystals and aligned polycrystals. To probe the flux pinning, we modified the strength, number, and morphology of defects. Varying the oxygen content (7-δ) in YBa 2 Cu 3 O 7-δ or irradiating the materials with ions, having either light or heavy masses, gives systematic changes in the character of the all-important defects

  13. Status of RBCB testing of LMR oxide fuel in EBR-II

    International Nuclear Information System (INIS)

    Strain, R.V.; Bottcher, J.H.; Gross, K.C.; Lambert, J.D.B.; Ukai, S.; Nomura, S.; Shikakura, S.; Katsuragawa, M.

    1991-01-01

    The status is given of the the American-Japanese collaborative program in Experimental Breeder Reactor 2 to determine the run-beyond-cladding-breach performance of (UPu)O 2 fuel pins for liquid-metal cooled reactors. Phase 1 of the collaboration involved eighteen irradiation tests over 1981--86 with 5.84-mm pins in 316 or D9 stainless steel. Emphasis in Phase 2 tests from 1989 onwards is with larger diameter (7.5mm) pins in advanced claddings. Results include delayed neutron and fission gas release data from breached pins, the impact of fuel-sodium reaction product formation on pin performance, and fuel and fission product contamination from failures. 13 refs, 1 fig., 4 tabs

  14. In-reactor performance of methods to control fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    Weber, E.T.; Gibby, R.L.; Wilson, C.N.; Lawrence, L.A.; Adamson, M.G.

    1979-01-01

    Inner surface corrosion of austenitic stainless steel cladding by oxygen and reactive fission product elements requires a 50 μm wastage allowance in current FBR reference oxide fuel pin design. Elimination or reduction of this wastage allowance could result in better reactor efficiency and economics through improvements in fuel pin performance and reliability. Reduction in cladding thickness and replacement of equivalent volume with fuel result in improved breeding capability. Of the factors affecting fuel-cladding chemical interaction (FCCI), oxygen activity within the fuel pin can be most readily controlled and/or manipulated without degrading fuel pin performance or significantly increasing fuel fabrication costs. There are two major approaches to control oxygen activity within an oxide fuel pin: (1) control of total oxygen inventory and chemical activity (Δ anti GO 2 ) by use of low oxygen-to-metal ratio (O/M) fuel; and (2) incorporation of a material within the fuel pin to provide in-situ control of oxygen activity (Δ anti GO 2 ) and fixation of excess oxygen prior to, or in preference to reaction with the cladding. The paper describes irradiation tests which were conducted in EBR-II and GETR incorporating oxygen buffer/getter materials and very low O/M fuel to control oxygen activity in sealed fuel pins

  15. Plutonium fuel program

    International Nuclear Information System (INIS)

    1979-01-01

    The work of the Project-Fuel Development reached the apex of its current programme during the course of the year. Notable success was recorded in the area of irradiation testing with the completion of the examination of the MFBS-7 irradiation. The irradiation group also prepared the seventh Filos experiment and this, as well as the DIDO-III test, began irradiation at the end of the year. Consideration was given and plans prepared for a revised pin filling line for bundle tests. Work also began on the conceptual design study for a pilot production line having a nominal capacity of 500 kg fuel per year. (Auth.)

  16. Dosimetry work and calculations in connection with the irradiation of large devices in the high flux materials testing reactor BR2

    International Nuclear Information System (INIS)

    De Raedt, C.; Leenders, L.; Tourwe, H.; Farrar, H. IV.

    1982-01-01

    For about fifteen years the high flux reactor BR2 has been involved in the testing of fast reactor fuel pins. In order to simulate the fast reactor neutron environment most devices are irradiated under cadmium screen, cutting off the thermal flux component. Extensive neutronic calculations are performed to help the optimization of the fuel bundle design. The actual experiments are preceded by irradiations of their mock-ups in BR02, the zero power model of BR2. The mock-up irradiations, supported by supplementary calculations, are performed for the determination of the main neutronic characteristics of the irradiation proper in BR2 and for the determination of the corresponding operation data. At the end of the BR2 irradiation, the experimental results, such as burn-ups, neutron fluences, helium production in the fuel pin claddings, etc. are correlated by neutronic calculations in order to examine the consistency of the post-irradiation results and to validate the routine calculation procedure and cross-section data employed. A comparison is made in this paper between neutronic calculation results and some post-irradiation data for MOL 7D, a cadmium screened sodium cooled loop containing a nineteen fuel pin bundle

  17. HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

    1998-03-01

    As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

  18. Modelling of WWER-1000 fuel: state and prospects

    International Nuclear Information System (INIS)

    Medvedev, A.; Bibilashvili, Yu.; Bogatyr, S.; Khvostov, G.

    1994-01-01

    The role of START-3 code in studying and computerized modelling of post-irradiation behaviour of standard fuel rods in real operation conditions of WWER-1000 reactors is described. The models used in the code are based on experimental study of material properties, processes and post irradiation research on standard and experimental fuel pins. The code capability is verified by comparison with data from experiments on WWER test rods performed in MR reactor, the Russia-Finland tests SOFIT and the international program FUMEX. The comparison performed and the results thus obtained demonstrate the satisfactory ability of START-3 code to simulate fuel rod behaviour in normal operation condition. The calculations confirm the experimentally observed evidence of an essential margin on serviceability of WWER-1000 fuel pin with three year operation cycle permitting an increase in design fuel burnup. 2 tabs., 18 figs

  19. Modelling of WWER-1000 fuel: state and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Medvedev, A; Bibilashvili, Yu; Bogatyr, S; Khvostov, G [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

    1994-12-31

    The role of START-3 code in studying and computerized modelling of post-irradiation behaviour of standard fuel rods in real operation conditions of WWER-1000 reactors is described. The models used in the code are based on experimental study of material properties, processes and post irradiation research on standard and experimental fuel pins. The code capability is verified by comparison with data from experiments on WWER test rods performed in MR reactor, the Russia-Finland tests SOFIT and the international program FUMEX. The comparison performed and the results thus obtained demonstrate the satisfactory ability of START-3 code to simulate fuel rod behaviour in normal operation condition. The calculations confirm the experimentally observed evidence of an essential margin on serviceability of WWER-1000 fuel pin with three year operation cycle permitting an increase in design fuel burnup. 2 tabs., 18 figs.

  20. Safeguards approach for irradiated fuel

    International Nuclear Information System (INIS)

    Harms, N.L.; Roberts, F.P.

    1987-03-01

    IAEA verification of irradiated fuel has become more complicated because of the introduction of variations in what was once presumed to be a straightforward flow of fuel from reactors to reprocessing plants, with subsequent dissolution. These variations include fuel element disassembly and reassembly, rod consolidation, double-tiering of fuel assemblies in reactor pools, long term wet and dry storage, and use of fuel element containers. This paper reviews future patterns for the transfer and storage of irradiated LWR fuel and discusses appropriate safeguards approaches for at-reactor storage, reprocessing plant headend, independent wet storage, and independent dry storage facilities

  1. 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

  2. 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

  3. 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)

  4. 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

  5. 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)

  6. Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

    International Nuclear Information System (INIS)

    Harp, Jason M.; Demkowicz, Paul A.

    2014-01-01

    In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10"-"4 to 10"-"5) of as manufactured defects and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from intentionally failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application was considered. Previous experience utilizing similar techniques, the expected activities in AGR-3/4 rings, and analysis of this work indicate using GECT to evaluate AGR-3/4 will be feasible. The GECT technique was also applied to other irradiated nuclear fuel systems currently available in the HFEF hot cell, including oxide fuel pins, metallic fuel pins, and monolithic plate fuel. Results indicate GECT with the HFEF PGS is effective. (author)

  7. 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

  8. 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

  9. 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

  10. 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

  11. Influence of irradiation on defects creation in pin diode structure

    International Nuclear Information System (INIS)

    Sopko, V.; Dammer, J.; Sopko, B.; Chren, D.

    2012-01-01

    In this paper the manufacture of type S1 PIN diodes and radiation defect induce by fast neutrons were studied. A shift from VV"- to VV (neutral) is observed in neutron irradiated diodes. From the results obtained, an explanation that clearly offers itself is that the nature of the defects produced by irradiation of material exhibiting N type conductivity is different from those for type P material. Given that the experiments were conducted with the same material, i.e., the dopant present in the material remained unchanged, it can be stated that simply by changing the type of conductivity with increasing dose, a different kind of defects is produced, having different activation energies in the forbidden band. All these results are consistent with the ongoing RD 50 experiments at CERN.

  12. 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

  13. 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.

  14. 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)

  15. KEY RESULTS FROM IRRADIATION AND POST-IRRADIATION EXAMINATION OF AGR-1 UCO TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; Morris, Robert N.

    2016-11-01

    The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior within the US program. The fuel exhibited a very low incidence of TRISO coating failure during irradiation and post-irradiation safety testing at temperatures up to 1800°C. Advanced PIE methods have allowed particles with SiC coating failure to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of

  16. Fission product release from UO2 during irradiation. Diffusion data and their application to reactor fuel pins

    International Nuclear Information System (INIS)

    Findlay, J.R.; Johnson, F.A.; Turnbull, J.A.; Friskney, C.A.

    1980-01-01

    Release of fission product species from UO 2 , and to a limited extent from (U, Pu)0 2 was studied using small scale in-reactor experiments in which these interacting variables may be separated, as far as is possible, and their influences assessed. Experiments were at fuel ratings appropriate to water reactor fuel elements and both single crystal and poly-crystalline specimens were used. They employed highly enriched uranium such that the relative number of fissions occurring in plutonium formed by neutron capture was small. The surface to volume ratio (S/V) of the specimens was well defined thus reducing the uncertainties in the derivation of diffusion coefficients. These experiments demonstrate many of the important characteristics of fission product behaviour in UO 2 during irradiation. The samples used for these experiments were small being always less than 1g with a fissile content usually between 2 and 5mg. Polycrystalline materials were taken from batches of production fuel prepared by conventional pressing and sintering techniques. The enriched single crystals were grown from a melt of sodium and potassium chloride doped with UO 2 powder 20% 235 U content. The irradiations were performed in the DIDO reactor at Harwell. The neutron flux at the specimen was 4x10 16 neutrons m -2 s -1 providing a heat rating within the samples of 34.5 MW/teU

  17. 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)

  18. Techniques and devices developed by the CEA for hot cell and in-situ examinations of PWR components and PWR fuel assembliess after irradiation

    International Nuclear Information System (INIS)

    Van Craeynest, J.C.; Leseur, A.; Lhermenier, A.; Cytermann, R.

    1981-11-01

    Within the framework of the electro-nuclear development of the PWR system, the CEA has provided itself with facilities for developing techniques for analyzing assemblies, pins and fuels. These are examinations and tests on irradiated heads and assemblies with the aid of the Fuel Examination Module (FEM), of machining of assemblies and examinations in the Celimene hot laboratory or detailed examinations and analyses on fuel elements using eddy currents, the electronic microprobe and the Fisher ''permeascope'' which enables the outline of the oxide coat present on the cladding to be followed [fr

  19. 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

  20. Fast breeder fuel element development

    International Nuclear Information System (INIS)

    Marth, W.; Muehling, G.

    1983-08-01

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

  1. 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.

  2. 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)

  3. TS-1 and TS-2 transient overpower tests on FFTF fuel

    International Nuclear Information System (INIS)

    Pitner, A.L.; Ferrell, P.C.; Culley, G.E.; Weber, E.T.

    1985-01-01

    The TS-1 and TS-2 TREAT transient experiments subjected a low burnup (2 MWd/kg) and a medium burnup (58 MWd/kg), respectively, FFTF irradiated fuel pin to unprotected 5 cents/s overpower transient conditions. The fuel pin failure response was similar in the two tests, which demonstrated a large margin to failure (P/P 0 > 3) and a favorable upper level failure location. Thus, for these transient conditions, burnup effects on transient performance appeared to be minimal in the range tested. Pin disruption in the medium burnup TS-2 test was more severe due to the higher fission gas pressurization, but failure occurred at only a 5% lower power level than for the low burnup TS-1 fuel pin. Both tests exhibited axial extrusion of molten fuel to the region above the fuel column several seconds before pin failure, demonstrating a potentially beneficial inherent safety mechanism to delay failure and mitigate accident consequences

  4. Current Status of the LIFE Fast Reactors Fuel Performance Codes

    International Nuclear Information System (INIS)

    Yacout, A.M.; Billone, M.C.

    2013-01-01

    The LIFE-4 (Rev. 1) code was calibrated and validated using data from (U,Pu)O2 mixed-oxide fuel pins and UO2 blanket rods which were irradiation tested under steady-state and transient conditions. – It integrates a broad material and fuel-pin irradiation database into a consistent framework for use and extrapolation of the database to reactor design applications. – The code is available and running on different computer platforms (UNIX & PC) – Detailed documentations of the code’s models, routines, calibration and validation data sets are available. LIFE-METAL code is based on LIFE4 with modifications to include key phenomena applicable to metallic fuel, and metallic fuel properties – Calibrated with large database from irradiations in EBR-II – Further effort for calibration and detailed documentation. Recent activities with the codes are related to reactor design studies and support of licensing efforts for 4S and KAERI SFR designs. Future activities are related to re-assessment of the codes calibration and validation and inclusion of models for advanced fuels (transmutation fuels)

  5. Irradiated fuel performance evaluation technology development

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  6. Gamma scanning of the irradiated HANARO fuels

    International Nuclear Information System (INIS)

    Hong, Kwon Pyo; Lee, K. S.; Park, D. G.; Baik, S. Y.; Song, W. S.; Kim, T. Y.; Seo, C. K.

    1997-02-01

    To conform the burnup state of the fuels, we have transported the irradiated HANARO fuels from the reactor to IMEF (Irradiated Material Examination Facility), and executed gamma scanning for the fuels. By measuring the gamma-rays from the irradiated fuels we could see the features of the relative burnup distributions in the fuel bundles. All of 17 fuel bundles were taken in and out between HANARO and IMEF from March till August in 1996, and we carried out the related regulations. Longitudinal gamma scanning and angular gamma scanning are done for each fuel bundle without dismantlement of the bundles. (author). 5 tabs., 25 figs

  7. Fuel fabrication processes, design and experimental conditions for the joint US-Swiss mixed carbide test in FFTF (AC-3 test)

    International Nuclear Information System (INIS)

    Stratton, R.W.; Ledergerber, G.; Ingold, F.; Latimer, T.W.; Chidester, K.M.

    1993-01-01

    The preparation of mixed carbide fuel for a joint (US-Swiss) irradiation test in the US Fast Flux Test Facility (FFTF) is described, together with the experiment design and the irradiation conditions. Two fabrication routes were compared. The US produced 66 fuel pins containing pellet fuel via the powder-pellet (dry) route, and the Swiss group produced 25 sphere pac pins of mixed carbide using the internal gelation (wet) route. Both sets of fuel met all t the requirements of the specifications concerning soichiometry, chemical composition and structure. The pin designs were as similar as possible. The test operated successfully in the FFTF for 620 effective full power days until October 1988 and reached over 8% burn up with peak powers of around 80 kW/m. The conclusions were that the choice of sphere pac or pellet fuel for reactor application is dependent on preferred differences in fabrication (e.g. economics and environmental factors) and not on differences in irradiation behaviour. (orig.)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  9. Fuel temperature prediction during high burnup HTGR fuel irradiation test. US-JAERI irradiation test for HTGR fuel

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.

    1995-01-01

    This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)

  10. In-pile study of the reaction between breeder fuel and sodium

    International Nuclear Information System (INIS)

    Hugot, J.P.

    1982-10-01

    Studies carried out until now show that the determinant parameter of fuel can failure evolution is the development of the reaction between mixed uranium and plutonium dioxide and sodium. The parameters of the reaction are presented from results of an out of pile study, as also results obtained from examination on pins failed in reactors. The best way to study in pile the development of the reaction was to irradiate at a constant power a fuel pin containing sodium. In the experiment, the pin was equipped with a central thermocouple. It shows, that the reaction is developing intergranularly, from cracks and interpellet spaces, in an internal fringe of the fuel before spreading to the periphery. An overheating of the pin is associated to the development of the reaction as also a modification of the fuel pin geometry and a reduction of the oxide [fr

  11. The physics of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Robin, M.

    1980-01-01

    The knowledge of the neutron irradiation effect is essential in dealing with all subjects related to the fuel. Neutron irradiation provokes fission reactions within the fuel and produces new nuclides. The formation chains are described and the importance of each isotope in the fuel cycle is explained with regards to its own characteristics. To solve the system of equations giving the evolution of different nuclides concentrations, the corresponding effective cross-sections and flux received are given by standard codes used for reactor calculations. A good test for calculation methods is the experimental study of irradiated fuel. Many techniques have been developed for this purpose. The last chapter compares fuel evolution in different reactors, in connection with some specific characteristics. (author)

  12. Design of single-walled NaK capsules for fast breeder fuel pins irradiation (IVO-FR2-Vg7 program)

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Hafner, H.E.

    1979-01-01

    In Frame of the Joint Irradiation Program IVO-FR2 between the Nuclear Research Centre of Karlsruhe (RFA) and the Junta de Energia Nuclear (Spain) is carried out in the FR2 reactor (Karlsruhe) the irradiation of 12 mixed-oxide fuel rods of 172 mm length. These test rods are first irradiated under various conditions in four modified FR2 capsule (Typ 7). Two versions of single-walled NaK (78% K) are used for this purpose. This report contains the design and description of these two capsule versions as well as the considerations required to oftain the operations licence, supplemented by the relevant figures. (author)

  13. Silicon P.I.N. Junctions used for studies of radiation damage; Etude de l'irradiation aux neutrons rapides du silicium au moyen de jonctions P.I.N

    Energy Technology Data Exchange (ETDEWEB)

    Lanore, J [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-06-01

    Irradiation of silicon P.I.N. junction has been studied primarily for the purpose of developing a radiation damage dosimeter, but also for the purpose of investigating silicon itself. It is known that the rate of recombination of electrons and holes is a linear function of defects introduced by neutron irradiation. Two methods have been used to measure that rate of recombination: forward characteristic measurements, recovery time measurements. In order to explain how these two parameters depend on recombination rate we have given a theory of the P.I.N. junction. We have also given an idea of the carrier lifetime dependence versus temperature. Annealing effects in the range of 70 to 700 K have also been studied, we found five annealing stages with corresponding activation energies. As an application for these studies, we developed a radiation damage dosimeter with which we made several experiments in facilities such as Naiade or Marias. (author) [French] L'irradiation de structures P.I.N. etait faite dans le but d'etudier principalement la mise au point d'un dosimetre a ''radiation damage'' et aussi pour etudier plus profondement le silicium lui-meme. On sait que le taux de recombinaison electrons-trous est une fonction lineaire du taux de defauts introduits par irradiation aux neutrons. Deux methodes ont ete utilisees pour atteindre ce taux de recombinaison: mesures de la caracteristique directe, mesures du temps de retournement. Pour expliquer de quelle facon ces parametres dependent du taux de recombinaison. Nous avons donne une theorie de la jonction P.I.N. Nous avons aussi donne l'allure des variations du temps de vie des porteurs en fonction de la temperature. Nous avons d'autre part effectue des recuits entre 70 et 700 K, domaine dans lequel nous avons trouve cinq etapes de ''guerison'' avec les energies d'activation correspondantes. En application de ces etudes nous avons mis ou point un dosimetre a ''radiation damage'' avec lequel nous avons effectue des

  14. 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

  15. Irradiation test and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Song, K. C.; Moon, J. S.

    2002-05-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  16. Fission gas release during post irradiation annealing of large grain size fuels from Hinkley point B

    International Nuclear Information System (INIS)

    Killeen, J.C.

    1997-01-01

    A series of post-irradiation anneals has been carried out on fuel taken from an experimental stringer from Hinkley Point B AGR. The stringer was part of an experimental programme in the reactor to study the effect of large grain size fuel. Three differing fuel types were present in separate pins in the stringer. One variant of large grain size fuel had been prepared by using an MgO dopant during fuel manufactured, a second by high temperature sintering of standard fuel and the third was a reference, 12μm grain size fuel. Both large grain size variants had similar grain sizes around 35μm. The present experiments took fuel samples from highly rated pins from the stringer with local burn-up in excess of 25GWd/tU and annealed these to temperature of up to 1535 deg. C under reducing conditions to allow a comparison of fission gas behaviour at high release levels. The results demonstrate the beneficial effect of large grain size on release rate of 85 Kr following interlinkage. At low temperatures and release rates there was no difference between the fuel types, but at temperatures in excess of 1400 deg. C the release rate was found to be inversely dependent on the fuel grain size. The experiments showed some differences between the doped and undoped large grains size fuel in that the former became interlinked at a lower temperature, releasing fission gas at an increased rate at this temperature. At higher temperatures the grain size effect was dominant. The temperature dependence for fission gas release was determined over a narrow range of temperature and found to be similar for all three types and for both pre-interlinkage and post-interlinkage releases, the difference between the release rates is then seen to be controlled by grain size. (author). 4 refs, 7 figs, 3 tabs

  17. Fission gas release during post irradiation annealing of large grain size fuels from Hinkley point B

    Energy Technology Data Exchange (ETDEWEB)

    Killeen, J C [Nuclear Electric plc, Barnwood (United Kingdom)

    1997-08-01

    A series of post-irradiation anneals has been carried out on fuel taken from an experimental stringer from Hinkley Point B AGR. The stringer was part of an experimental programme in the reactor to study the effect of large grain size fuel. Three differing fuel types were present in separate pins in the stringer. One variant of large grain size fuel had been prepared by using an MgO dopant during fuel manufactured, a second by high temperature sintering of standard fuel and the third was a reference, 12{mu}m grain size fuel. Both large grain size variants had similar grain sizes around 35{mu}m. The present experiments took fuel samples from highly rated pins from the stringer with local burn-up in excess of 25GWd/tU and annealed these to temperature of up to 1535 deg. C under reducing conditions to allow a comparison of fission gas behaviour at high release levels. The results demonstrate the beneficial effect of large grain size on release rate of {sup 85}Kr following interlinkage. At low temperatures and release rates there was no difference between the fuel types, but at temperatures in excess of 1400 deg. C the release rate was found to be inversely dependent on the fuel grain size. The experiments showed some differences between the doped and undoped large grains size fuel in that the former became interlinked at a lower temperature, releasing fission gas at an increased rate at this temperature. At higher temperatures the grain size effect was dominant. The temperature dependence for fission gas release was determined over a narrow range of temperature and found to be similar for all three types and for both pre-interlinkage and post-interlinkage releases, the difference between the release rates is then seen to be controlled by grain size. (author). 4 refs, 7 figs, 3 tabs.

  18. Sodium erosion of boron carbide from breached absorber pins

    International Nuclear Information System (INIS)

    Basmajian, J.A.; Baker, D.E.

    1981-03-01

    The purpose of the irradiation experiment was to provide an engineering demonstration of the irradiation behavior of breached boron carbide absorber pins. By building defects into the cladding of prototypic absorber pins, and performing the irradiation under typical FFTF operating conditions, a qualitative assessment of the consequences of a breach was achieved. Additionally, a direct comparison of pin behavior with that of the ex-reactor test could be made

  19. 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

  20. In-pile irradiation of rock-like oxide fuels

    International Nuclear Information System (INIS)

    Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H.

    2001-01-01

    Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO 2 single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

  1. In-pile irradiation of rock-like oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2001-07-01

    Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO{sub 2} single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

  2. 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)

  3. 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)

  4. 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

  5. Apparatus for inspecting a irradiated nuclear fuel rod

    International Nuclear Information System (INIS)

    Saura, Hideaki; Yonemura, Eizo.

    1975-01-01

    Object: To increase safety and inspection efficiency by operating irradiated fuel rods, which are accommodated in a water-filled pool after being taken out from the reactor. Structure: When making inspection of irradiated fuel rods, particularly the cladding tube thereof, a fuel box which stores irradiated fuel rods in a water pool is secured to a securement mechanism with slime removal apparatus and inspection apparatus on either side capable of being vertically moved, and it is then stopped at a water depth of about 2 meters. When the lid of the box is opened, irradiated fuel rods are taken out with gripping means and then secured together with the gripping means to an operation base provided on the outside of the pool. Thereafter, the box is lowered by operating pedals on the operation base to completely pull out the irradiated fuel rods from the box, and the irradiated fuel rods are then horizontally moved and then held in a suspended state. Next a slime removal apparatus in raised by operating pedals and an inspection element assembly are progressively raised for inspection of the state of the cladding tube of each fuel rod after removal of slime therefrom. (Nakamura, S.)

  6. The 3rd irradiation test plan of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Park, J. H. and others

    2001-05-01

    The objective of the 3rd irradiation test of DUPIC fuel at the HANARO is to estimate the in-core behaviour of a DUPIC pellet that is irradiated up to more than average burnup of CANDU fuel. The irradiation of DUPIC fuel is planned to start at May 21, 2001, and will be continued at least for 8 months. The burnup of DUPIC fuel through this irradiation test is thought to be more than 7,000 MWd/tHE. The DUPIC irradiation rig instrumented with three SPN detectors will be used to accumulate the experience for the instrumented irradiation and to estimate the burnup of irradiated DUPIC fuel more accurately. Under normal operating condition, the maximum linear power of DUPIC fuel was estimated as 55.06 kW/m, and the centerline temperature of a pellet was calculated as 2510 deg C. In order to assess the integrity of DUPIC fuel under the accident condition postulated at the HANARO, safety analyses on the locked rotor and reactivity insertion accidents were carried out. The maximum centerline temperature of DUPIC fuel was estimated 2590 deg C and 2094 deg C for each accident, respectively. From the results of the safety analysis, the integrity of DUPIC fuel during the HANARO irradiation test will be secured. The irradiated DUPIC fuel will be transported to the IMEF. The post-irradiation examinations are planned to be performed at the PIEF and IMEF.

  7. Designing the KNK II-TOAST irradiation experiment with the saturn-FS code

    International Nuclear Information System (INIS)

    Ritzhaupt-Kleissl, H.J.; Elbel, H.; Heck, M.

    1991-01-01

    In order to study the existing specification of FBR fuel with respect to allowable fabrication tolerances with the objective to reduce the expense of fabrication and quality control, the TOAST irradiation experiment will be carried out in the 3 rd core of the KNK II. This experiment shall investigate the influence of the following fuel specification parameters on the operational behaviour: - Fuel diameter - Stoichiometry - Sintering atmosphere - Fill gas in the fuel pin. The combination of these test parameters led to a fabrication of 6 types of fuel pellets, giving together with two fill gas mixtures a total of 9 fuel pin types. Design calculations in the frame of the standard licensing procedure have been performed with the SATURN-FS fuel pin behaviour code. These calculations have been done for the steady-state behaviour as well as for some defined design transients, such as startup procedures and overpower ramps

  8. Development of out-of-pile version of instrumented irradiation capsule for determination of online creep deformation

    International Nuclear Information System (INIS)

    Venkatesu, Sadu; Saxena, Rajesh; Chaurasia, P.K.; Muthuganesh, M.; Murugan, S.; Venugopal, S.

    2016-01-01

    Materials used for fuel cladding and structural components in fast reactors can undergo significant dimensional and physical changes due to exposure to high energy neutrons. At high temperatures in nuclear environment, material undergoes considerable deformation due to thermal and irradiation creep. Diametral increase of fuel pin due to thermal and irradiation creep, apart from irradiation swelling, reduces the coolant flow area around the fuel pins affecting the effective removal of heat generated in the fuel pins. The changes due to creep can be determined by two types of material irradiation tests in reactor. The first type includes non-instrumented irradiation tests with specimen dimensional evaluations carried out in post-irradiation examinations. The second type includes instrumented irradiation tests with online monitoring and/or controlling of test conditions and real time measurement of changes in dimensions of the specimen. During instrumented irradiation tests, parameters such as specimen temperature, the load exerted on the specimen, specimen elongation, etc. can be monitored and/or controlled using suitable components such as linear variable differential transformers (LVDTs), bellows, thermocouples, etc. Instrumented irradiation experiments in reactors are relatively complex in design but can provide full information on the experimental parameters. Such benefits provide motivation for development of instrumented irradiation capsule to measure creep behavior online during in-pile instrumented irradiation tests. Out-of-pile version of the instrumented irradiation capsule for determination of online creep deformation has been developed and tested in the furnace by raising the temperature gradually up to 330 °C. This paper discusses the details of the design, assembly of experimental set up and experimental results of the out-of-pile version of instrumented capsule developed in our laboratory for determination of online creep deformation. (author)

  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. 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

  11. 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)

  12. Development of a new bench for puncturing of irradiated fuel rods in STAR hot laboratory

    Science.gov (United States)

    Petitprez, B.; Silvestre, P.; Valenza, P.; Boulore, A.; David, T.

    2018-01-01

    A new device for puncturing of irradiated fuel rods in commercial power plants has been designed by Fuel Research Department of CEA Cadarache in order to provide experimental data of high precision on fuel pins with various designs. It will replace the current set-up that has been used since 1998 in hot cell 2 of STAR facility with more than 200 rod puncturing experiments. Based on this consistent experimental feedback, the heavy-duty technique of rod perforation by clad punching has been preserved for the new bench. The method of double expansion of rod gases is also retained since it allows upgrading the confidence interval of volumetric results obtained from rod puncturing. Furthermore, many evolutions have been introduced in the new design in order to improve its reliability, to make the maintenance easier by remote handling and to reduce experimental uncertainties. Tightness components have been studied with Sealing Laboratory Maestral at Pierrelatte so as to make them able to work under mixed pressure conditions (from vacuum at 10-5 mbar up to pressure at 50 bars) and to lengthen their lifetime under permanent gamma irradiation in hot cell. Bench ergonomics has been optimized to make its operating by remote handling easier and to secure the critical phases of a puncturing experiment. A high pressure gas line equipped with high precision pressure sensors out of cell can be connected to the bench in cell for calibration purposes. Uncertainty analyses using Monte Carlo calculations have been performed in order to optimize capacity of the different volumes of the apparatus according to volumetric characteristics of the rod to be punctured. At last this device is composed of independent modules which allow puncturing fuel pins out of different geometries (PWR, BWR, VVER). After leak tests of the device and remote handling simulation in a mock-up cell, several punctures of calibrated specimens have been performed in 2016. The bench will be implemented soon in hot

  13. Development of cutting device for irradiated fuel rod

    International Nuclear Information System (INIS)

    Lee, E. P.; Jun, Y. B.; Hong, K. P.; Min, D. K.; Lee, H. K.; Su, H. S.; Kim, K. S.; Kwon, H. M.; Joo, Y. S.; Yoo, K. S.; Joo, J. S.; Kim, E. K.

    2004-01-01

    Post Irradiation Examination(PIE) on irradiated fuel rods is essential for the evaluation of integrity and irradiation performance of fuel rods of commercial reactor fuel. For PIE, fuel rods should be cut very precisely. The cutting positions selected from NDT data are very important for further destructive examination and analysis. A fuel rod cutting device was developed witch can cut fuel rods longitudinal very precisely and can also cut the fuels into the same length rod cuts repeatedly. It is also easy to remove the fuel cutting powder after cutting works and it can extend the life time of cutting device and lower the contamination level of hot cell

  14. SATURN-S - a program system for the description of the thermomechanical behaviour of reactor fuel pins under irradiation

    International Nuclear Information System (INIS)

    Pesl, R.; Freund, D.; Gaertner, H.; Steiner, H.

    1987-07-01

    On the basis of post irradiation examination results of various irradiation experiments with different fuel types real case calculations showed many of the existing models to be applicable to a restricted extent only. Therefore a re- and partially new formulation of models was necessary. Furthermore, the data base had been actualized and numerical procedures had been improved. This, together with the capabilities of modern computer systems, conducted the development of the program system SATURN-S with a strictly modular structure, specified by the requirements of the determination of the superposition of effects. In the present report the program SATURN-S as well as some analysis results are presented. (orig./HP) [de

  15. Status of fuel irradiation tests in HANARO

    International Nuclear Information System (INIS)

    Kim, Hark Rho; Lee, Choong Sung; Lee, Kye Hong; Jun, Byung Jin; Lee, Ji Bok

    1999-01-01

    Since 1996 after finishing the long-term operational test, HANARO (High-Flux Advanced Neutron Application Reactor) has been extensively used for material irradiation tests, beam application research, radioisotope production and neutron activation analysis. This paper presents the fuel irradiation test activities which are now conducted or have been finished in HANARO. KAERI developed LEU fuel using an atomization method for the research reactors. Using this LEU, we have set up and conducted three irradiation programs: (1) medium power irradiation test using a short-length mini-assembly made of 3.15 gU/cc U 3 Si, (2) high power irradiation tests using full-length test assemblies made of 3.15 gU/cc U 3 Si, and (3) irradiation test using a short-length mini-plate made of 4.8 gU/cc U 3 Si 2 . DUPIC (Direct Use of spent PWR fuels in CANDU Reactors) simulation fuel pellets, of which compositions are very similar to DUPIC pellets to keep the similarity in the thermo-mechanical property, were developed. Three mini-elements including 5 pellets each were installed in a capsule. This capsule has been irradiated for 2 months and unloaded from the HANARO core at the end of September 1999. Another very important test is the HANARO fuel qualification program at high power, which is required to resolve the licensing issue. This test is imposed on the HANARO operation license due to insufficient test data under high power environment. To resolve this licensing issue, we have been carrying out the required irradiation tests and PIE (Post-irradiation Examination) tests. Through this program, it is believed that the resolution of the licensing issue is achieved. In addition to these programs, several fuel test plans are under way. Through these vigorous activities of fuel irradiation test programs, HANARO is sure to significantly contribute to the national nuclear R and D programs. (author)

  16. 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

  17. Power-to-melt evaluation of fresh mixed-oxide fast reactor fuel. Technical improvements of the post-irradiation-experiment and the evaluation of the results for the power-to-melt test PTM-2 in 'JOYO'

    International Nuclear Information System (INIS)

    Yamamoto, Kazuya; Kushida, Naoya; Koizumi, Atsuhiro

    1999-11-01

    The second Power-To-Melt (PTM) test, PTM-2, was performed in the experimental fast reactor 'JOYO'. All of the twenty-four fuel pins of the irradiation vehicle, B5D-2, for the PTM-2 test, were provided for post-irradiation-experiment (PIE) to evaluate the PTM values. In this study, the PIE technique for PTM test was established and the PTM results were evaluated. The findings are as follows: The maximum fuel-melting ratio on the transverse section was 10.7%, and was within the limit of fuel-melting in this PTM test enough. Unexpected fuel-melting amount to a ratio of 11.8% was found at ∼24 mm below the peak power elevation in a test fuel pin. It is possible that this arose from secondary fuel-melting. Combination of metallographical observation with X-ray microanalysis of plutonium distribution was very effective for the identification of once-molten fuel zone. The PTM evaluation suggested that dependence of the PTM on the fuel pellet density was stronger than that of previous foreign PTM tests, while the dependence on the pellet-cladding gap and the oxygen-to-metal ratio was indistinctly. The dependence on the cladding temperature and the fill gas composition was not shown as well. (author)

  18. 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.

  19. Observations of in-reactor endurance and rupture life for fueled and unfueled FTR cladding

    International Nuclear Information System (INIS)

    Lovell, A.J.; Christensen, B.Y.; Chin, B.A.

    1979-01-01

    Reactor component endurance limits are important to nuclear experimenters and operators. This paper investigates endurance limits of 316 CW fuel pin cladding. The objective of this paper is to compare and analyze two different sets of FTR fuel pin cladding data. The first data set is from unfueled pressurized cladding irradiated in the Experimental Breeder Reactor No. II (EBR-II). This data set was generated in an assembly in which the temperature was monitored and controlled. The second data set contains observations of breached and unbreached EBR-II test fuel pins covering a large range of temperature, power and burnup conditions

  20. 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

  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. 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

  3. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    International Nuclear Information System (INIS)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-01-01

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

  4. Criticality safety evaluation for long term storage of FFTF fuel in interim storage casks

    International Nuclear Information System (INIS)

    Richard, R.F.

    1995-01-01

    It has been postulated that a degradation phenomenon, referred to as ''hot cell rot'', may affect irradiated FFTF mixed plutonium-uranium oxide (MOX) fuel during dry interim storage. ''Hot cell rot'' refers to a variety of phenomena that degrade fuel pin cladding during exposure to air and inert gas environments. It is thought to be a form of caustic stress corrosion cracking or environmentally assisted cracking. Here, a criticality safety analysis was performed to address the effect of the ''hot cell rot'' phenomenon on the long term storage of irradiated FFTF fuel in core component containers. The results show that seven FFTF fuel assemblies or six Ident-69 pin containers stored in core component containers within interim storage casks will remain safely subcritical

  5. Consequences of the improvement of fast reactor material behavior under irradiation on fuel element performance

    International Nuclear Information System (INIS)

    Leclere, J.; Dupouy, J.M.; Marcon, J.P.

    1979-01-01

    The most important problems in fast reactor fuel element come from the excessive swelling of the structural materials used. The limitations of irradiation time for a given reactor result from the cladding or hexagonal wrapper deformations. Irradiation creep plays a major role, either in inducing additional deformations, or in providing possible ways of accommodation of bending stresses. Progress has been made in designing swelling resistant and/or low irradiation creep modulus materials. For instance in FRANCE, annealed 316 SS has been eliminated from pin and subassembly, and replaced by cold worked 316; we are now considering introduction of stabilizing elements in 316 SS as a further improvement and studying different alloys (nickel alloys, or ferritic steels). It has to be checked that the improvement of irradiation characteristic is not counterbalanced by losses on other properties (embrittlement for instance). Considering that pushing off or eliminating a limit may lead to the onset of a new one, it is porposed to make a review of the consequences of substantial improvement of structural material behavior

  6. Development of 4S and related technologies (2). Long life metallic fuel

    International Nuclear Information System (INIS)

    Yacout, A.M.; Tsuboi, Y.; Ueda, N.

    2009-01-01

    This paper provides an overview of the long life metallic fuel to be used in the 4S reactor. The 4S fuel design is presented and implications of its characteristics on fuel performance are discussed. Main design characteristics include the long fuel life time of 30 years and the wider and longer fuel pins compared to EBR-II and FFTF fuel pins. The LIFE-METAL fuel performance code was used to evaluate the performance of the 4S fuel design. The code has been validated using post irradiation examination data of metallic fuel irradiated in EBR-II. The performance evaluation shows the benign nature of the design. The design enables the fuel to perform adequately during reactor operations without violating any of a conservative set of steady state design criteria. A survey evaluation of the fuel performance is also presented. This performance bounding evaluation took into account possible fuel swelling behavior and cladding temperature range that represents worst case scenarios. The evaluation showed that the fuel maintains its integrity even under those worst case conditions. (author)

  7. EDF energy generation UK transport of irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    James, R. [EDF Energy, London, (United Kingdom)

    2015-07-01

    This paper give an overview of irradiated fuel transport in the UK. It describes the design of irradiated fuel flask used by EDF Energy; operational experience and good practices learnt from over 50 years of irradiated fuel transport. The AGRs can store approximately 9 months generation of spent fuel, hence the ability to transport irradiated fuel is vital. Movements are by road to the nearest railhead, typically less than 2 miles and then by rail to Sellafield, up to 400 miles, for reprocessing or long term storage. Road and rail vehicles are covered. To date in the UK: over 30,000 Magnox flask journeys and over 15,000 AGR A2 flask journeys have been carried out.

  8. 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)

  9. Calculation simulation of equivalent irradiation swelling for dispersion nuclear fuel

    International Nuclear Information System (INIS)

    Cai Wei; Zhao Yunmei; Gong Xin; Ding Shurong; Huo Yongzhong

    2015-01-01

    The dispersion nuclear fuel was regarded as a kind of special particle composites. Assuming that the fuel particles are periodically distributed in the dispersion nuclear fuel meat, the finite element model to calculate its equivalent irradiation swelling was developed with the method of computational micro-mechanics. Considering irradiation swelling in the fuel particles and the irradiation hardening effect in the metal matrix, the stress update algorithms were established respectively for the fuel particles and metal matrix. The corresponding user subroutines were programmed, and the finite element simulation of equivalent irradiation swelling for the fuel meat was performed in Abaqus. The effects of the particle size and volume fraction on the equivalent irradiation swelling were investigated, and the fitting formula of equivalent irradiation swelling was obtained. The results indicate that the main factors to influence equivalent irradiation swelling of the fuel meat are the irradiation swelling and volume fraction of fuel particles. (authors)

  10. Studies and manufacture of plutonium fuel

    International Nuclear Information System (INIS)

    Bussy, P.; Mustelier, J.P.; Pascard, R.

    1964-01-01

    The studies carried out at the C.E.A. on the properties of fast neutron reactor fuels, the manufacture of fuel elements and their behaviour under irradiation are broadly outlined. The metal fuels studied are the ternary alloys U Pu Mo, U Pu Nb, U Pa Ti, U Pa Zr, the ceramic fuels being mixed uranium and plutonium oxides, carbides and nitrides obtained by sintering. Results are given on the manufacture of uranium fuel elements containing a small proportion of plutonium, used in a critical experiment, and on the first experiments in the manufacture of fuel elements for the reactor Rapsodie. Finally the results of irradiation tests carried out on the prototype fuel pins for Rapsodie are described. (authors) [fr

  11. 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)

  12. 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

  13. 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.

  14. High reliability fuel in the US

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Leggett, R.D.; Walters, L.C.; Matthews, R.B.

    1986-05-01

    The fuels development program of the United States is described for liquid metal reactors (LMR's). The experience base, status and future potential are discussed for the three systems - oxide, metal and carbide - that have proved to have high reliability. Information is presented showing burnup capability of the oxide fuel system in a large core, e.g., FFTF, to be 150 MWd/kgM with today's technology with the potential for a capability as high as 300 MWd/kgM. Data provided for the metal fuel system show 8 at. % being routinely achieved as the EBR-II driver fuel with good potential for extending this to 15 at. % since special test pins have already exceeded this burnup level. The data included for the carbide fuel system are from pin and assembly irradiations in EBR-II and FFTF, respectively. Burnup to 12 at. % appears readily achievable with burnups to 20 at. % being demonstrated in a few pins. Efforts continue on all three systems with the bulk of the activity on metal and oxide

  15. Advances in carbide fuel element development for fast reactor application

    International Nuclear Information System (INIS)

    Dienst, W.; Kleykamp, H.; Muehling, G.; Reiser, H.; Steiner, H.; Thuemmler, F.; Wedermeyer, H.; Weimar, P.

    1977-01-01

    The features of the carbide fuel development programme are reviewed and evaluated. Single pin and bundle irradiations are carried out under thermal, epithermal and fast flux conditions, the latter in the DFR and KNK-II reactors. Several fuel concepts in the region of representative SNR clad temperatures are compared by parameter and performance tests. A conservative concept is based on He-bonded 8 mm pins with (U,Pu)C pellets and a smear density of 75% TD, operating at 800 W/cm rod power and burnup to 70 MWd/kg. The preparation of mixed carbide fuels is carried out by carbothermic reduction of the oxides in different methods supported by equivalent carbon content, grain size and phase distribution analysis. The fuel for subassembly performance tests is produced in a pilot plant of 0,5 t/year capacity. Compatibility studies reveal that cladding carburization is the only chemical interaction with carbide fuels. This effect leads to a reduction in ductility of the stainless steel. Fission products apparently play no role in the compatibility behaviour. Comprehensive studies lead to reliable information on the chemical and thermodynamic state of the fuel under irradiation. The swelling of carbide fuels and the fission gas release are examined and analysed. Cladding plastic strain by fuel swelling occurs during steady-state operation because the irradiation creep is rather slow compared to oxide fuels. The cladding strain observed depends on the fuel porosity and the cladding strength. The development of carbide fuel pins is complemented by the application of comprehensive computer models. In addition to the steady-state tests power cycling and safety tests are under performance. Up to 1980 the results are summarized for the final design and specification. The development target of the present program is to fabricate several subassemblies for test operation in the SNR 300 by 1981

  16. Analysis of irradiation temperature in fuel rods of OGL-1 fuel assembly

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Minato, Kazuo; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-10-01

    Irradiation temperature in the fuel rods of 5th OGL-1 fuel assembly was analysed by the system composed by STPDSP2 and TRUMP codes. As the measured input-data, following parameters were allowed for; circumferential heating distribution around the fuel rod, which was measured in the JMTR critical assembly, axial heating distribution through the fuel rod, ratio of peak heatings of three fuel rods, and pre- and post-irradiation outer radii of the fuel compacts and inner radii of the graphite sleeves, which had been measured in PIE of the 5th OGL-1 fuel assembly. In computation the axial distributions of helium coolant temperature through the fuel rod and the heating value of each fuel rod were, firstly, calculated as input data for TRUMP. The TRUMP calculation yielded the temperatures which were fitted in those measured by all of the thermo-couples installed in the fuel rods, by adjusting only the value of the surface heat transfer coefficient, and consequently, the temperatures in all portions of the fuel rod were obtained. The apparent heat transfer coefficient changed to 60% of the initial values in the middle period of irradiation. For this reduction it was deduced that shoot had covered the surface of the fuel rod during irradiation, which was confirmed in PIE. Beside it, several things were found in this analysis. (author)

  17. 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)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  19. RECH-1 test fuel irradiation status report

    International Nuclear Information System (INIS)

    Marin, J.; Lisboa, J.; Olivares, L.; Chavez, J.

    2005-01-01

    Since May 2003, one RECH-1 fuel element has been submitted to irradiation at the HFR-Petten, Holland. By November 2004 the irradiation has achieved its pursued goal of 55% burn up. This irradiation qualification service will finish in the year 2005 with PIE tests, as established in a contractual agreement between the IAEA, NRG, and CCHEN. This report presents the objectives and the current results of this fuel qualification under irradiation. Besides, a brief description of CHI/4/021, IAEA's Technical Cooperation Project that has supported this irradiation test, is also presented here. (author)

  20. Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Tsai Hanchung [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439-4803 (United States)

    2012-08-15

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 Degree-Sign C, cooling to 522 Degree-Sign C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at {approx}0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher {Delta}T between fuel center and cladding than at core center, together providing more rare earths at the cladding and

  1. 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.)

  2. 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

  3. Irradiation testing of coated particle fuel at Hanaro

    International Nuclear Information System (INIS)

    Goo Kim, Bong; Sung Cho, Moo; Kim, Yong Wan

    2014-01-01

    TRISO-coated particle fuel is developing to support development of VHTR in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in VHTR in the HANARO at KAERI. This experiment is currently undergoing under the atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak burn-up of about 4 at% and a peak fast neutron fluence of about 1.7 x 10 21 n/cm 2 , PIE will be carried out at KAERI's Irradiated Material Examination Facility. This paper is described characteristics of coated particle fuel, the design of test rod and irradiation device for coated particle fuel, discusses the technical results for irradiation testing at HANARO. (authors)

  4. MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

    1997-08-01

    Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

  5. Silicon P.I.N. Junctions used for studies of radiation damage; Etude de l'irradiation aux neutrons rapides du silicium au moyen de jonctions P.I.N

    Energy Technology Data Exchange (ETDEWEB)

    Lanore, J. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-06-01

    Irradiation of silicon P.I.N. junction has been studied primarily for the purpose of developing a radiation damage dosimeter, but also for the purpose of investigating silicon itself. It is known that the rate of recombination of electrons and holes is a linear function of defects introduced by neutron irradiation. Two methods have been used to measure that rate of recombination: forward characteristic measurements, recovery time measurements. In order to explain how these two parameters depend on recombination rate we have given a theory of the P.I.N. junction. We have also given an idea of the carrier lifetime dependence versus temperature. Annealing effects in the range of 70 to 700 K have also been studied, we found five annealing stages with corresponding activation energies. As an application for these studies, we developed a radiation damage dosimeter with which we made several experiments in facilities such as Naiade or Marias. (author) [French] L'irradiation de structures P.I.N. etait faite dans le but d'etudier principalement la mise au point d'un dosimetre a ''radiation damage'' et aussi pour etudier plus profondement le silicium lui-meme. On sait que le taux de recombinaison electrons-trous est une fonction lineaire du taux de defauts introduits par irradiation aux neutrons. Deux methodes ont ete utilisees pour atteindre ce taux de recombinaison: mesures de la caracteristique directe, mesures du temps de retournement. Pour expliquer de quelle facon ces parametres dependent du taux de recombinaison. Nous avons donne une theorie de la jonction P.I.N. Nous avons aussi donne l'allure des variations du temps de vie des porteurs en fonction de la temperature. Nous avons d'autre part effectue des recuits entre 70 et 700 K, domaine dans lequel nous avons trouve cinq etapes de ''guerison'' avec les energies d'activation correspondantes. En application de ces etudes nous avons mis ou point un

  6. 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)

  7. Alternatives for water basin spent fuel storage: executive summary and comparative evaluation

    International Nuclear Information System (INIS)

    Viebrock, J.M.

    1979-09-01

    A five part report identifies and evaluates alternatives to conventional methods for water basin storage of irradiated light water reactor fuel assemblies (spent fuel). A recommendation is made for development or further evaluation of one attractive alternative: Proceed to develop fuel disassembly with subsequent high density storage of fuel pins (pin storage). The storage alternatives were evaluated for emplacement at reactor, in existing away-from-reactor storage facilities and in new away-from-reactor facilities. In the course of the study, the work effort necessarily extended beyond the pool wall in scope to properly assess the affects of storage alternatives on AFT systems

  8. Irradiation of novel MTR fuel plates in BR2

    International Nuclear Information System (INIS)

    Verboomen, B.; Aoust, Th.; Beeckmans De Westmeerbeeck, A.; De Raedt, Ch.

    2000-01-01

    Since the end of 1999, novel MTR fuel plates with very high-density meat are being irradiated in BR2. The purpose of the irradiation is to investigate the behaviour of these fuel plates under very severe reactor operation conditions. The novel fuel plates are inserted in two standard six-tube BR2 fuel elements in the locations normally occupied by the standard outer fuel plates. The irradiation in BR2 was prepared by carrying out detailed neutron Monte Carlo calculations of the whole BR2 core containing the two experimental fuel elements for various positions in the reactor and for various azimuthal orientations of the fuel elements. Comparing the thus determined fission density levels and azimuthal profiles in the new MTR fuel plates irradiated in the various channels allowed the experimenters to choose the most appropriate BR2 channel and the most appropriate fuel element orientation. (author)

  9. Irradiation effects on fuels for space reactors

    International Nuclear Information System (INIS)

    Ranken, W.A.; Cronenberg, A.W.

    1984-01-01

    A review of irradiation-induced swelling and gas release experience is presented here for the three principal fuels UO 2 , UC, and UN. The primary advantage of UC and UN over UO 2 is higher thermal conductivity and attendant lower fuel temperature at equivalent pellet diameter and power density, while UO 2 offers the distinct benefit of well-known irradiation performance. Irradiation test results indicate that at equivalent burnup, temperature, and porosity conditions, UC experiences higher swelling than UO 2 or UN. Fission gas swelling becomes important at fuel temperatures above 1320 K for UC, and at somewhat higher temperatures for UO 2 and UN. Evidence exists that at equivalent fuel temperatures and burnups, high density UO 2 and UN experience comparable swelling behavior; however, differences in thermal conductivity influence overall irradiation performance. The low conductivity of UO 2 results in higher thermal gradients which contribute to fuel microcracking and gas release. As a result UO 2 exhibits higher fractional gas release than UN, at least or burnups up to about 3%

  10. Calculation and experimental estimation of the equation of state of irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bober, M; Breitung, W; Karow, H U; Schumacher, G [Gesellschaft fuer Kernforschung mbH, INR Kernforschungszentrum, Karlsruhe (Germany)

    1977-07-01

    The gas pressure development in an irradiated mixed oxide fuel is mainly influenced by fission gases and volatile fission products in the temperature range below the melting point and by the fuel material itself and the less volatile fission products in the temperature region above 4000 K. Besides the temperature the important factors for the vapor pressure are the oxygen potential of the fuel and the concentration of fission products in the fuel. As demonstrated previously the oxygen potential influences strongly the pressure of vapor species above (U Pu)O{sub 2}. The pressure of the species U, UO, UO{sub 2}, Pu, PuO, PuO{sub 2} varies over a range of more than five orders of magnitude by variation of the oxygen potential at 2000 K. Similar effects were observed with oxides of the fission products. Fission products dissolved in mixed oxide fuel on the other hand can influence significantly the oxygen potential of the irradiated mixed oxide. In the first paragraph of the paper an attempt is made to calculate oxygen potentials of mixed oxides containing dissolved fission products. The model used is based on the equilibrium of oxygen defects in the mixed oxide. The chemical state and distribution of fission products is a further behavior that should be considered in calculation of the local and overall pressures and behavior of the fuel. Fission products were transported during the irradiation time and collect at different positions within the fuel pin. This process can produce high local concentrations of fission products, thus enabling elements with low overall concentrations to reach their saturation pressure. The distribution of fission products and their behavior in irradiated mixed oxide fuel is described in the second paragraph. The third paragraph deals with the calculation of vapor pressures that has been conducted using a model described for uranium-plutonium mixed oxides. This model is based on the law of mass action and provides vapor pressures as a

  11. Gamma-ray spectroscopy on irradiated fuel rods

    International Nuclear Information System (INIS)

    Terremoto, Luis Antonio Albiac

    2009-01-01

    The recording of gamma-ray spectra along an irradiated fuel rod allows the fission products to be qualitatively and quantitatively examined. Among all nondestructive examinations performed on irradiated fuel rods by gamma-ray spectroscopy, the most comprehensive one is the average burnup measurement, which is quantitative. Moreover, burnup measurements by means of gamma-ray spectroscopy are less time-consuming and waste-generating than burnup measurements by radiochemical, destructive methods. This work presents the theoretical foundations and experimental techniques necessary to measure, using nondestructive gamma-ray spectroscopy, the average burnup of irradiated fuel rods in a laboratory equipped with hot cells. (author)

  12. Oxide fuel fabrication technology development of the FaCT project (5). Current status on 9Cr-ODS steel cladding development for high burn-up fast reactor fuel

    International Nuclear Information System (INIS)

    Ohtsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya

    2011-01-01

    This paper describes evaluation results of in-reactor integrity of 9Cr and 12Cr-ODS steel cladding tubes and the plan for reliability improvement in homogeneous tube production, both of which are key points for the commercialized use of ODS steels as long-life fuel cladding tubes. A fuel assembly in the BOR-60 irradiation test including 9Cr and 12Cr-ODS fuel pins has achieved the highest burn-up, i.e. peak burn-up of 11.9at% and peak neutron dose of 51dpa, without any fuel pin rupture and microstructure instability. In another fuel assembly containing 9Cr and 12Cr-ODS steel fuel pins whose peak burn-up was 10.5at%, one 9Cr-ODS steel fuel pin failed near the upper end of the fuel column. A peculiar microstructure change occurred in the vicinity of the ruptured area. The primary cause of this fuel pin rupture and microstructure change was shown to be the presence of metallic Cr inclusions in the 9Cr-ODS steel tube, which had passed an ultrasonic inspection test for defects. In the next stage from 2011 to 2013, the fabrication technology of full pre-alloy 9Cr-ODS steel cladding tube will be developed, where the handling of elemental powder is prohibited in the process. (author)

  13. Irradiated fuel examination using the Cerenkov technique

    International Nuclear Information System (INIS)

    Nicholson, N.; Dowdy, E.J.

    1981-03-01

    A technique for monitoring irradiated nuclear fuel inventories located in water filled storage ponds has been developed and demonstrated. This technique provides sufficient qualitative information to be useful as a confirmatory technique to International Atomic Energy Agency inspectors. Measurements have been made on the Cerenkov glow light intensity from irradiated fuel that show the intensity of this light to be proportional to the cooling time. Fieldable instruments used in several tests confirm that such measurements can be made easily and rapidly, without fuel assembly movement or the introduction of apparatus into the storage ponds. The Cerenkov technique and instrumentation have been shown to be of potential use to operators of reactor spent fuel facilities and away from reactor storage facilities, and to the International Atomic Energy Agency inspectors who provide surveillance of the irradiated fuel stored in these facilities

  14. Chemical characterisation of experimental PFBR-MOX fuel - an overview

    International Nuclear Information System (INIS)

    Mallik, G.K.; Arun Kumar; Panakkal, J.P.; Kamath, H.S.

    2003-01-01

    Fuel pins for experimental PFBR-type fuel containing ( Nat U 0.335 233 U 0.37 Pu 0.295 )O 2 are being manufactured at AFFF and will be irradiated in FBTR. A number of chemical characterization experiments have been carried out on simulated and actual fuel pellets to meet the design specifications of the fuel. An overview is being presented to highlight the related facts. (author)

  15. Irradiation Testing of TRISO-Coated Particle Fuel in Korea

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Yeo, Sunghwan; Jeong, Kyung-Chai; Eom, Sung-Ho; Kim, Yeon-Ku; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung; Kim, Yong Wan

    2014-01-01

    In Korea, coated particle fuel is being developed to support development of a VHTR. At the end of March 2014, the first irradiation test in HANARO at KAERI to demonstrate and qualify TRISO-coated particle fuel for use in a VHTR was terminated. This experiment was conducted in an inert gas atmosphere without on-line temperature monitoring and control, or on-line fission product monitoring of the sweep gas. The irradiation device contained two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The duration of irradiation testing at HANARO was about 135 full power days from last August 2013. The maximum average power per particle was about 165 mW/particle. The calculated peak burnup of the TRISO-coated fuel was a little less than 4 atom percent. Post-irradiation examination is being carried out at KAERI’s Irradiated Material Examination Facility beginning in September of 2014. This paper describes characteristics of coated particle fuel, the design of the test rod and irradiation device for this coated particle fuel, and discusses the technical results of irradiation testing at HANARO. (author)

  16. 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)

  17. Behavior of pre-irradiated fuel under a simulated RIA condition

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide

    1994-07-01

    This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-3, conducted in the Nuclear Safety Research Reactor (NSSR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and analyses, interpretations, and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 19.6MWd/kgU with average linear heat rate of 25.3 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a deposited energy of 174±6 cal/g·fuel and a peak fuel enthalpy of 130±5 cal/g·fuel under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The cladding surface temperature increased to only 150degC, and the test resulted in slight fuel deformation and no fuel failure. An estimated rod-average fission gas release during the transient was about 2.2%. Through the detailed fuel examinations, the information concerning microstructural change in the fuel pellets were also obtained. (author)

  18. Molybdenum-UO2 cermet irradiation at 1145 K.

    Science.gov (United States)

    Mcdonald, G.

    1971-01-01

    Two molybdenum-uranium dioxide cermet fuel pins with molybdenum clad were fission-heated in a forced-convection helium coolant for sufficient time to achieve 5.3% burnup. The cermet core contained 20 wt % of 93.2% enriched uranium dioxide. The results were as follows: there was no visible change in the appearance of the molybdenum clad during irradiation; the maximum increase in diameter of the fuel pins was 0.8%; there was no migration of uranium dioxide along grain boundaries and no evident interaction between molybdenum and uranium dioxide; and, finally, approximately 12% of the fission gas formed was released from the cermet core into the gas plenum.

  19. Reliability of fast reactor mixed-oxide fuel during operational transients

    International Nuclear Information System (INIS)

    Boltax, A.; Neimark, L.A.; Tsai, Hanchung; Katsuragawa, M.; Shikakura, S.

    1991-07-01

    Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 figs., 4 tabs

  20. Preliminary test results for post irradiation examination on the HTTR fuel

    International Nuclear Information System (INIS)

    Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

    2007-01-01

    The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

  1. 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

  2. Testing program PIN on experiments reported in EPRI-NP-369

    International Nuclear Information System (INIS)

    Kunt, J.; Pazdera, F.

    1980-03-01

    The testing is described of the applicability of program PIN for the thermomechanical simulation of PWR fuel elements. Program PIN is a modified program GAPCON-THERMAL-2. Four modifications are presented mostly affecting the results, viz., the fuel restructuring model, the fuel densification model, the gaseous fission product release model, and the fuel cladding creep model. The results show that the initial data for simulating fuel element behaviour in accident situations are determined using program PIN with accuracy comparable to that using other programs. (J.P.)

  3. Theoretical and experimental analysis of fast reactor fuel performance

    International Nuclear Information System (INIS)

    Kummerer, K.R.; Freund, D.; Steiner, H.

    1982-09-01

    In order to predict behavior, performance, and capability of prototypic fuel pins a standard operational scheme for the SNR-300 fast breeder reactor is established considering besides normal operation unscheduled power changes and shutdowns. The behavior during the whole lifetime is calculated using the updated SATURN codes and - for special conditions as power transients and skewed fuel rod power - the new TRANSIENT and TEXDIF codes. The results of these calculations are compared to experimental findings. It is demonstrated that the level of modeling and the knowledge of material properties under irradiation are sufficient for a quantitative description of the fuel pin performance under the above mentioned conditions. (orig.) [de

  4. Design and construction of the Fuels and Materials Examination Facility

    International Nuclear Information System (INIS)

    Burgess, C.A.

    1979-01-01

    Final design is more than 85 percent complete on the Fuels and Materials Examination Facility, the facility for post-irradiation examination of the fuels and materials tests irradiated in the FFTF and for fuel process development, experimental test pin fabrication and supporting storage, assay, and analytical chemistry functions. The overall facility is generally described with specific information given on some of the design features. Construction has been initiated and more than 10% of the construction contracts have been awarded on a fixed price basis

  5. Focused ultrasonic wave testing, in immersion of spent fuel cans

    International Nuclear Information System (INIS)

    Poinboeuf, P.; Furlan, J.

    1984-10-01

    To detect weak and very weak damage of the fuel can, ultrasonic testing has been used. For that, a simple mechanical device, allowing to maintain an optimal ultrasonic focussing on irradiated cans, is presented. Its aim is to correct the variation of the incidence angle due to the possible ovalization of pins. After a description of the device, the results obtained with tests carried out on non-irradiated cans, including artificial ovalized regions, standard defects, are presented. After the description of the adaptation of this mechanism on a test bench which allows an helicoidal exploration of pins, some results obtained in hot cell during examinations experimental pins and previously tested by Foucault current [fr

  6. Irradiation behavior of uranium oxide - Aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products and as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show that, with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 g U/cm 3 ) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼63% 235 U burnup). (author)

  7. Irradiation behavior of uranium oxide-aluminum dispersion fuel

    International Nuclear Information System (INIS)

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

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼ 63% 235 U burnup)

  8. Status of irradiation testing and PIE of MOX (Pu-containing) fuel

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Zhou, Y.N.; Ryz, M.A.

    1995-01-01

    This paper describes AECL's mixed oxide (MOX) fuel-irradiation and post-irradiation examination (PIE) program. Post-irradiation examination results of two major irradiation experiments involving several (U, Pu)O 2 fuel bundles are highlighted. One experiment involved bundles irradiated to burnups ranging fro 400 to 1200 MWh/kgHe in the Nuclear Power Demonstration (NPD) reactor. The other experiment consisted of several (U, Pu)O 2 bundles irradiated to burnups of up to 500 Mwh/kgHe in the National Research Universal (NRU) reactor. Results of these experiments demonstrate the excellent performance of CANDU MOX fuel. This paper also outlines the status of current MOX fuel irradiation tests, including the irradiation of various (U, Pu)O 2 bundles. The strategic importance of MOX fuel to CANDU fuel-cycle flexibility is discussed. (author)

  9. Monitoring for fuel sheath defects in three shipments of irradiated CANDU nuclear fuel

    International Nuclear Information System (INIS)

    Johnson, H.M.

    1978-01-01

    Analyses of radioactive gases within the Pegase shipping flask were performed at the outset and at the completion of three shipments of irradiated nuclear fuel from the Douglas Point Generating Station to Whiteshell Nuclear Research Establishment. No increases in the concentration of active gases, volatiles or particulates were observed. The activity of the WR-1 bay water rose only marginally due to the storage of the fuel. Other tests indicated that minimal surface contamination was present. These data established that defects in fuel element sheaths did not arise during the transport or the handling of this irradiated fuel. The observation has significance for the prospect of irradiated nuclear fuel transfer and handling in preparation for storage or disposal. (author)

  10. 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)

  11. The Analysis Of Spent Fuel Utilization For A Gamma Irradiator

    International Nuclear Information System (INIS)

    MS, Pudjijanto; Setiyanto

    2002-01-01

    The gamma irradiator using RSG-GAS spent fuels was analyzed. The cylindrical geometry of the irradiator was designed by locating the spent fuels the cylindrical periphery. The analysis was focused to evaluate the feasibilities of the irradiator as a fruits and vegetables irradiator. The spent fuels activities were calculated using Origen2 code, while the dose rate at the irradiation positions was determined by linear attenuation model with transport coefficient. The evaluated results showed that the cylindrical geometry of irradiators with diameter around 1-1.5 m gave the effective dose rate for fruits and vegetables preservation. It can be concluded that one can use the RSG-GAS spent fuels effectively as a gamma irradiator for certain applications

  12. Computer-controlled gamma-ray scanner for irradiated reactor fuel

    International Nuclear Information System (INIS)

    Mandler, J.W.; Coates, R.A.; Killian, E.W.

    1979-01-01

    Gamma-ray scanning of irradiated fuel is an important nondestructive technique used in the thermal fuels behavior program currently under way at the Idaho National Engineering Laboratory. This paper is concerned with the computer-controlled isotopic gamma-ray-scanning system developed for postirradiation examination of fuel and includes a brief discussion of some scan results obtained from fuel rods irradiated in the Power-Burst Facility to illustrate gamma-ray spectrometry for this application. Both burnup profiles and information concerning fission-product migration in irradiated fuel are routinely obtained with the computer-controlled system

  13. Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

  14. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    Directory of Open Access Journals (Sweden)

    M.K. MEYER

    2014-04-01

    Full Text Available High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  15. Irradiation performance of U-Mo monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, M. K.; Gan, J.; Jue, J. F.; Keiser, D. D.; Perez, E.; Robinson, A.; Wachs, D. M.; Woolstenhulme, N. [Idaho National Laboratory, Idaho (Korea, Republic of); Kim, Y.S.; Hofman, G. L. [Argonne National Laboratory, Lemont (United States)

    2014-04-15

    High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  16. Nondestructive analysis of irradiated fuels

    International Nuclear Information System (INIS)

    Dudey, N.D.; Frick, D.C.

    1977-01-01

    The principal nondestructive examination techniques presently used to assess the physical integrity of reactor fuels and cladding materials include gamma-scanning, profilometry, eddy current, visual inspection, rod-to-rod spacing, and neutron radiography. LWR fuels are generally examined during annual refueling outages, and are conducted underwater in the spent fuel pool. FBR fuels are primarily examined in hot cells after fuel discharge. Although the NDE techniques are identical, LWR fuel examinations emphasize tests to demonstrate adherence to technical specification and reliable fuel performance; whereas, FBR fuel examinations emphasize aspects more related to the relative performance of different types of fuel and cladding materials subjected to variable irradiation conditions

  17. 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.

  18. Nuclear fuel cycle: (5) reprocessing of irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Williams, J.A.

    1977-09-01

    The evolution of the reprocessing of irradiated fuel and the recovery of plutonium from it is traced out, starting by following the Manhatten project up to the present time. A brief description of the plant and processes used for reprocessing is given, while the Purex process, which is used in all plants today, is given special attention. Some of the important safety problems of reprocessing plants are considered, together with the solutions which have been adopted. Some examples of the more important safety aspects are the control of activity, criticality control, and the environmental impact. The related topic of irradiated fuel transport is briefly discussed.

  19. Investigation of WWER fuel behaviour under MIR power ramps

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Novikov, V.V.; Agafonov, S.N.

    1996-01-01

    The paper discusses results of experimental WWER fuel investigation under power ramps. Specificity of using the research reactor ''MIR'' to accomplish scheduled power rating of fuel is considered. The paper presents the methodology of experiments using irradiation facility ''TEST''. Reactor experiments were performed at burn-up ∼ 10000 MW.day/t UO 2 using standard fuel pins and the ones having backfitted fuel and cladding. (author). 7 figs, 1 tab

  20. Results of Microstructural Examinations of Irradiated LEU U-Mo Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, ID 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organization (Australia)

    2009-06-15

    Introduction: The RERTR program is responsible for converting research reactors that use high-enriched uranium fuels to ones that use low-enriched uranium fuels [1]. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. Based on the results of initial fuel plate testing, adjustments have been made to the characteristics of fuel plates to improve the stability of the fuel microstructure. One improvement has been to add Si to the matrix of a dispersion fuel. This material is also being added at the fuel/cladding interface of a monolithic fuel. This paper will discuss the irradiation performance of these fuels, in terms of the stability of their microstructures during irradiation. Results and discussion: The post-irradiation examinations of fuel plates are performed at the Idaho National Laboratory. These examinations consist of visual examinations of fuel plates, gamma scanning, thickness measurements, oxide thickness measurements, and optical metallographic examinations of the fuel plate microstructures. Microstructural analysis is also performed using scanning electron microscopy. Overall, U-7Mo and U-10Mo alloy fuels have displayed the best irradiation performance, particularly, when a Si-containing Al alloy is used as the dispersion fuel matrix. The benefit of using this type of matrix is that the commonly observed fuel/cladding interaction that occurs during irradiation is reduced and the interaction layer that forms exhibit stable behavior during irradiation. Monolithic-type fuels, which consist of a U-Mo foil encased in Al alloy cladding, are also being developed. These types of fuels are also showing promise and will continue to be developed. One challenge with this type of fuel is in trying to maximize the bond strength at the foil/cladding interface. Fuel/cladding interactions can affect the quality of the boding at this interface. Si is being added to improve the characteristics

  1. Management of irradiated CANDU fuel

    International Nuclear Information System (INIS)

    Lupien, Mario

    1985-01-01

    The nuclear industry, like any other industrial activity, generates waste and, since these radioactive products are known to be hazardous both to man and his natural environment, they are subject to stringent controls. The irradiated fuel is also highly radioactive and remains so for thousands of years. It is estimated that by the year 2000, nuclear reactors in Canada alone will have produced some 50 Gg of radioactive fuel which is stored at the nuclear plant site itself. The nuclear industry plays a leading role in the research and development effort to find suitable waste-management methods. Its R and D programs cover many scientific fields, including chemistry, and therefore demand a considerable amount of coordination. The knowledge acquired in this multidisciplinary context should form a basis for solving many of today's industrial-waste problems. This paper describes the various stages in the long management process. In the medium term, the irradiated fuel will be stored in surface installations but the long-term solution proposed is to emplace the used fuel or the fuel recycle waste deep underground in a stable geologic formation

  2. 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.

  3. Some results on development, irradiation and post-irradiation examinations of fuels for fast reactor-actinide burner (MOX and inert matrix fuel)

    International Nuclear Information System (INIS)

    Poplavsky, V.; Zabudko, L.; Moseev, L.; Rogozkin, B.; Kurina, I.

    1996-01-01

    Studies performed have shown principal feasibility of the BN-600 and BN-800 cores to achieve high efficiency of Pu burning when MOX fuel with Pu content up to 45% is used. Valuable experience on irradiation behaviour of oxide fuel with high Pu content (100%) was gained as a result of operation of two BR-10 core loadings where the maximum burnup 14 at.% was reached. Post-irradiation examination (PIE) allowed to reveal some specific features of the fuel with high plutonium content. Principal irradiation and PIE results are presented in the paper. Use of new fuel without U-238 provides the maximum burning capability as in this case the conversion ratio is reduced to zero. Technological investigations of inert matrix fuels have been continued now. Zirconium carbide, zirconium nitride, magnesium oxide and other matrix materials are under consideration. Inert matrices selection criteria are discussed in the paper. Results of technological study, of irradiation in the BOR-60 reactor and PIE results of some inert matrix fuels are summarized in this report. (author). 2 refs, 1 fig., 3 tabs

  4. Irradiation testing of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-01-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

  5. Horizontal modular dry irradiated fuel storage system

    Science.gov (United States)

    Fischer, Larry E.;