WorldWideScience

Sample records for alloy-ht-9

  1. MINIMARS conceptual design: Report I. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.D. (ed.)

    1985-12-01

    This report contains separate articles of seven aspects of the MINIMARS programs. The areas discussed are Fusion Engineering Design Center, Halo Model and Computer Code, safety design, the University of Wisconsin blankets, activation product transport in a FLiBe-VANADIUM alloy HT-9 system, a halo scraper/direct converter system, and heat transport power conversion. The individual articles are cataloged separately. (WRF)

  2. MINIMARS conceptual design: Report I. Volume 2

    International Nuclear Information System (INIS)

    This report contains separate articles of seven aspects of the MINIMARS programs. The areas discussed are Fusion Engineering Design Center, Halo Model and Computer Code, safety design, the University of Wisconsin blankets, activation product transport in a FLiBe-VANADIUM alloy HT-9 system, a halo scraper/direct converter system, and heat transport power conversion. The individual articles are cataloged separately

  3. Transmutation of alloys in MFE facilities as calculated by REAC (a computer code system for activation and transmutation)

    International Nuclear Information System (INIS)

    A computer code system for fast calculation of activation and transmutation has been developed. The system consists of a driver code, cross-section libraries, flux libraries, a material library, and a decay library. The code is used to predict transmutations in a Ti-modified 316 stainless steel, a commercial ferritic alloy (HT9), and a V-15%Cr-5%Ti alloy in various magnetic fusion energy (MFE) test facilities and conceptual reactors

  4. TEM specimen preparation for the HFIR-MFE-RB1 experiment

    International Nuclear Information System (INIS)

    TEM specimens of the ferritic alloys HT-9, 9Cr-1Mo and 2 1/4Cr-1Mo have been prepared for inclusion in the HFIR-MFE-RB1 irradiation. The specimen matrix encompasses all the conditions being irradiated in the EBR-II AD-2 test. Additionally, the matrix includes specimens of HT-9 weld fusion zones and simulated heat-affected zones to study the microstructural response of weldments

  5. Activation analysis for different structural alloys considered for ITER

    International Nuclear Information System (INIS)

    Activation calculations have been made for the austentic steel 316SS, the ferritic alloy HT-9, the titanium alloy Ti6A14V, and the vanadium alloy V5Cr5Ti in a liquid metal (Na) design suggested recently for ITER. The calculations show that the vanadium alloy has the minimum short and long-term radioactivity and BHP. It also has the minimum decay heat at all the time. The titanium alloy has less radioactivity than the austenitic and this ferritic alloys. However, the decay heat of this alloy could exceed that of the conventional alloys

  6. Strength and rupture-life transitions caused by secondary carbide precipitation in HT-9 during high-temperature low-rate mechanical testing

    International Nuclear Information System (INIS)

    The martensitic-ferritic alloy HT-9 is slated for long-term use as a fuel-cladding material in the Integral Fast Reactor. Analysis of published high-temperature mechanical property data suggests that secondary carbide precipitation would occur during service life causing substantial strengthening of the as-heat-treated material. Aspects of the kinetics of this precipitation process are extracted from calculations of the back stress necessary to produce the observed strengthening effect under various creep loading conditions. The resulting Arrhenius factor is shown to agree quantitatively with shifts to higher strength of crept material in reference to the intrinsic strength of HT-9. The results of very low constant strain-rate high-temperature tensile tests on as-heat-treated HT-9 that focus on the transition in strength with precipitation will be presented and related to rupture-life

  7. Performance of HT9 clad metallic fuel at high temperature

    International Nuclear Information System (INIS)

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

  8. Pressure Resistance Welding of High Temperature Metallic Materials

    Energy Technology Data Exchange (ETDEWEB)

    N. Jerred; L. Zirker; I. Charit; J. Cole; M. Frary; D. Butt; M. Meyer; K. L. Murty

    2010-10-01

    Pressure Resistance Welding (PRW) is a solid state joining process used for various high temperature metallic materials (Oxide dispersion strengthened alloys of MA957, MA754; martensitic alloy HT-9, tungsten etc.) for advanced nuclear reactor applications. A new PRW machine has been installed at the Center for Advanced Energy Studies (CAES) in Idaho Falls for conducting joining research for nuclear applications. The key emphasis has been on understanding processing-microstructure-property relationships. Initial studies have shown that sound joints can be made between dissimilar materials such as MA957 alloy cladding tubes and HT-9 end plugs, and MA754 and HT-9 coupons. Limited burst testing of MA957/HT-9 joints carried out at various pressures up to 400oC has shown encouraging results in that the joint regions do not develop any cracking. Similar joint strength observations have also been made by performing simple bend tests. Detailed microstructural studies using SEM/EBSD tools and fatigue crack growth studies of MA754/HT-9 joints are ongoing.

  9. Comparison of fracture behavior for low-swelling ferritic and austenitic alloys irradiated in the Fast Flux Test Facility (FFTF) to 180 DPA

    International Nuclear Information System (INIS)

    Fracture toughness testing was conducted to investigate the radiation embrittlement of high-nickel superalloys, modified austenitic steels and ferritic steels. These materials have been experimentally proven to possess excellent resistance to void swelling after high neutron exposures. In addition to swelling resistance, post-irradiation fracture resistance is another important criterion for reactor material selection. By means of fracture mechanics techniques the fracture behavior of those highly irradiated alloys was characterized in terms of irradiation and test conditions. Precipitation-strengthened alloys failed by channel fracture with very low postirradiation ductility. The fracture toughness of titanium-modified austenitic stainless steel D9 deteriorates with increasing fluence to about 100 displacement per atom (dpa), the fluence level at which brittle fracture appears to occur. Ferritic steels such as HT9 are the most promising candidate materials for fast and fusion reactor applications. The upper-shelf fracture toughness of alloy HT9 remained adequate after irradiation to 180 dpa although its ductile- brittle transition temperature (DBTT) shift by low temperature irradiation rendered the material susceptible to brittle fracture at room temperature. Understanding the fracture characteristics under various irradiation and test conditions helps reduce the potential for brittle fracture by permitting appropriate measure to be taken

  10. Design optimization of ferritic alloy LMFBR fuel assemblies as affected by in-reactor deformation

    International Nuclear Information System (INIS)

    Because of the relatively high resistance to irradiation-induced creep and swelling exhibited by the high strength ferritic alloy HT-9, it is a major structural material candidate for use in Liquid Metal Fast Breeder Reactor (LMFBR) fuel assembly hexagonal ducts, principally to minimize the duct dilation due to the nuclear environment. The design of a fuel assembly with an HT-9 duct can be performed as part of an overall plant performance optimization to achieve minimum plant operational costs. Such evaluations were performed to quantify the effect on fuel cycle costs of possible future changes to either the current estimates of HT-9 swelling and creep or to the uncertainty band widths of creep and swelling. The evaluations were conducted using a reactor system design computer code capable of implementing core design tradeoff studies to obtain a minimum cost optimization of the reactor system performance. The results of the study quantify the incentive for minimum creep and swelling and the incentive to obtain sufficient data to minimize the creep and swelling uncertainty band widths

  11. Evaluation of the advanced mixed oxide fuel test FO-2 irradiated in Fast Flux Test Facility

    International Nuclear Information System (INIS)

    The advanced mixed-oxide (UO2-PuO2) test assembly, FO-2, irradiated in the Fast Flux Test Facility (FFTF), is undergoing postirradiation examination (PIE). This is one of the first FFTF tests examined that used the advanced ferrite-martensite alloy, HT9, which is highly resistant to irradiation swelling. The FO-2 includes the first annular fueled pins irradiated in FFTF to undergo destructive examination. The FO-2 is a lead assembly for the ongoing FFTF Core Demonstration Experiment (CDE) (Leggett and Omberg 1987) and was designed to evaluate the effects of fuel design variables, such as pellet density, smeared density, and fuel form (annular or solid fuel), on advanced pin performance. The assembly contains a total of 169 fuel pins of twelve different types. The test was irradiated for 312 equivalent full power days (EFPD) in FFTF. It had a peak pin power of 13.7 kW/ft and reached a peak burnup of 65.2 MWd/kgM with a peak fast fluence of 9.9 /times/ 1022 n/cm2 (E > 0.1 MeV). This document discusses the test and its results. 6 refs., 19 figs., 4 tabs

  12. Evaluation of the advanced mixed-oxide fuel test FO-2 irradiated in the FFTF [Fast Flux Test Facility

    International Nuclear Information System (INIS)

    The advanced mixed-oxide (UO2-PuO2) test assembly, FO-2, irradiated in the Fast Flux Test Facility (FFTF) is undergoing postirradiation examination. This is one of the first FFTF tests examined that used the advanced ferrite-martensite alloy, HT9, which is highly resistant to irradiation swelling. The FO-2 includes the first annular fueled pins irradiated in FFTF to undergo destructive examination. The FO-2 is a lead assembly for the ongoing FFTF Core Demonstration Experiment (CDE) and was designed to evaluate the effects of fuel design variables, such as pellet density, smeared density, and fuel form (annular or solid fuel), on advanced pin performance. The assembly contains a total of 169 fuel pins of 12 different types. Two L (annular) fuel pins, GF02L04 (FFTF and transient tested) and GF02L09 (FFTF only), were destructively examined. Evaluation of the FO-2 fuel pins and assembly shows the excellent and predictable performance of the mixed-oxide fuels with HT9 structural material. This, combined with the robust behavior of the pins in transient tests, and the continued excellent performance of the CDE indicate this is a superior fuel system for liquid-metal reactors. It offers greatly reduced deformation during irradiation, while maintaining good operating characteristics

  13. Irradiation creep in austenitic and ferritic steels irradiated in a tailored neutron spectrum to induce fusion reactor levels of helium

    Energy Technology Data Exchange (ETDEWEB)

    Grossbeck, M.L.; Gibson, L.T. [Oak Ridge National Laboratory, TN (United States); Jitsukawa, S.

    1996-04-01

    Six austenitic stainless steels and two ferritic alloys were irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor where an atomic displacement level of 7.4 dpa was achieved and was then transferred to the High Flux Isotope Reactor for the remainder of the irradiation to a total displacement level of 19 dpa. Temperatures of 60 and 330{degree}C are reported on. At 330{degree}C irradiation creep was found to be linear in stress and fluence with rates in the range of 1.7 - 5.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. Annealed and cold-worked materials exhibited similar creep rates. There is some indication that austenitic alloys with TiC or TiO precipitates had a slightly higher irradiation creep rate than those without. The ferritic alloys HT-9 and Fe-16Cr had irradiatoin creep rates about 0.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. No meaningful data could be obtained from the tubes irradiated at 60{degree}C because of damage to the tubes.

  14. Irradiation creep of various ferritic alloys irradiated {approximately}400 C in the PFR and FFTF reactors

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B. [Washington State Univ., WA (United States); Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Eiholzer, C.R. [Westinghouse Hanford Co., WA (United States)

    1998-03-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400 C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400 C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 {times} 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.