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Sample records for steel-ht-9

  1. Effect of microstructure on the mechanical properties of a commercial 12Cr-1Mo steel (HT-9)

    International Nuclear Information System (INIS)

    Lechtenberg, T.A.

    1981-08-01

    The microstructure of a commercial 12Cr-1Mo steel (HT-9) and its associated effect on strength and toughness properties is being studied in a continuing program aimed at qualifying the alloy for use in fusion energy machines. Interim results show this alloy is subject to a degree of tempered martensite embrittlement and temper embrittlement. For applications projected for fusion machines at lower temperatures, a new heat treatment (1000 0 C, 1 h, air-cooled followed by 650 0 C tempering) at lower temperatures and shorter times than the vendor-recommended heat treatment has been identified. Microstructural differences between the treatments are discussed, and mechanical properties are correlated. 6 figures

  2. Microstructural examination of commercial ferritic alloys at 299 DPA

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1995-11-01

    Microstructures and density change measurements are reported for Martensitic commercial steels HT-9 and Modified 9Cr-lMo (T9) and oxide dispersion strengthened ferritic alloys MA956 and NU957 following irradiation in the FFTF/MOTA at 420 degrees C to 200 DPA. Swelling as determined by density change remains below 2% for all conditions. Microstructures are found to be stable except in recrystallized grains of MA957, which are fabrication artifacts, with only minor swelling in the Martensitic steels and α' precipitation in alloys with 12% or more chromium. These results further demonstrate the high swelling resistance and microstructural stability of the ferritic alloy class

  3. Automatic inspection for remotely manufactured fuel elements

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  4. Fusion reactor materials research in China

    International Nuclear Information System (INIS)

    Qian Jiapu

    1994-10-01

    The fusion materials research in China is introduced. Many kinds of structural materials (such as Ti-modified stainless steel, ferritic steel, HT-9, HT-7, oxide dispersion strengthening ferritic steel), tritium breeders (lithium, Li 2 O, γ-LiAlO 2 ) and plasma facing materials (PFMs) (graphite with TiC and SiC coatings) have been developed or being developed. A systematic research activities on irradiation effects, compatibility, plasma materials interaction, thermal shock during disruption, tritium production, release and permeation, neutron multiplication in Be and Pb, etc. have been performed. The research activities are summarized and some experimental results are also given

  5. Effect of specimen size on the upper shelf energy of ferritic steels

    International Nuclear Information System (INIS)

    Kumar, A.S.

    1990-01-01

    A methodology is proposed that can be used to predict the upper shelf energy (USE) of ferritic steels based on subsize specimen data. The proposed methodology utilizes the partitioning of the USE into energies required for crack initiation and crack propagation. Notched-only Charpy specimens are used in conjunction with precracked specimens to separate the two components. An unirradiated ferritic steel, HT-9, was used to demonstrate the validity of the methodology. Unlike previous correlations that were limited in their applicability to either highly ductile or brittle material, the proposed methodology is expected to be applicable over a wide range of ductility and to be particularly useful for materials that harden significantly during irradiation

  6. Permeation of deuterium implanted into vanadium alloys

    International Nuclear Information System (INIS)

    Anderl, R.A.; Longhurst, G.R.; Struttmann, D.A.

    1986-05-01

    Permeation of deuterium through the vanadium alloy, V-15Cr-5Ti, was investigated using 3-keV, D 3 + ion beams from a small accelerator. The experiments consisted of measurements of the deuterium reemission and permeation rates as a function of implantation fluence for 0.5-mm thick specimens heated to tempertures from 623 to 823 0 K. Implantation-side surface characterization was made by simultaneous measurements of sputtered ions with a secondary ion mass spectrometer (SIMS). Analyses of these measurements indicate that for the experimental conditions used, the steady-state deuterium permeation flux in V-15Cr-5Ti is approximately 18% of the implantation flux. This corresponds to approximately 1000 times that seen in the ferritic steel, HT-9, under comparable conditions. Measurement of deuterium diffusivity in V-15Cr-5Ti using permeation break-through times indicates D = 1.4 x 10 -8 exp(-.11 eV/kT) (m 2 /s)

  7. Recent results on implantation and permeation into fusion reactor materials

    Science.gov (United States)

    Anderl, R. A.; Holland, D. F.; Longhurst, G. R.; Struttman, D. A.

    This paper reports on implantation-driven permeation experiments that have been made for primary candidate alloy (PCA) and the ferritic steel HT-9 using deuterium ion beams from an accelerator. The results include measurements of the implantation flux and fluence dependence of the deuterium reemission and permeation for specimens heated to approximately 430(0)C. Simultaneous measurements of the ions sputtered from the specimen front surface with a secondary ion mass spectrometer provided some characterization of the surface condition throughout an experiment. For both materials, the permeation rate was lowered by the implantation process. However, the steady state permeation rate for HT-9 was found to be at least a factor of 5 greater than that for PCA.

  8. Interdiffusion between U-Zr-Mo and stainless steel cladding

    International Nuclear Information System (INIS)

    Hwang, J. Y.; Lee, B. S.; Lee, J. T.; Kang, Y. H.

    1998-01-01

    Interdiffusion investigations were carried out at 700 deg C for 200 hours for the diffusion couples assembled with the U-Zr-Mo ternary fuel versus austenitic stainless steel D9 and the U-Zr-Mo ternary fuel versus martensitic stainless steel HT9 respectively to investigate the fuel-cladding compatibility. SEM-EDS analysis was utilized to determine the composition and the penetration depths of the reaction layers. In the case of Fuel/D9 couple, (Fe, Cr, Ni) of the cladding elements formed the precipitates with the Zr, Mo and diminished the U concentration upto 800μ length from the fuel side. Composition of the precipitates was varied with the penetrated elements. In Fuel/HT9 couple, reaction layer was smaller than that of D9 couples and was less affected by cladding elements. The eutectic reaction appeared partially in the Fuel/HT9 diffusion couple

  9. Activation product transport in a FLiBe-vanadium alloy-HT9 system

    International Nuclear Information System (INIS)

    Klein, A.C.; Sze, D.K.

    1985-01-01

    An assessment is made of the gamma radiation hazards likely to be found around a fusion reactor heat transfer and tritium breeding loop which employs a vanadium alloy for the blanket and first wall structure and the ferritic-steel HT9 for the remainder of the loop. The coolant/tritium breeding fluid is the molten metallic salt FliBe. Since the radiation levels near the primary loop components are found to be less than 100 mR/hr 3-5 days after shutdown after three years of continuous full power operation, limited hands-on maintenance could be allowed. The very short half-lives of the predominant corrosion products make this result possible and make such a system very attractive

  10. Permeation of deuterium implanted into V-15Cr-5Ti

    International Nuclear Information System (INIS)

    Anderl, R.A.; Longhurst, G.R.; Struttmann, D.A.

    1987-01-01

    Permeation and reemission of deuterium for the vanadium alloy, V-15Cr-5Ti, was investigated using 3 keV, D 3 + ion beams from a small accelerator. The experiments consisted of measurement of the deuterium reemission and permeation rates as a function of implantation fluence for 0.5 mm thick specimens heated to temperatures from 623 K to 823 K. Implantation-side surface characterization was made by simultaneous measurements of sputtered ions with a secondary ion mass spectrometer (SIMS). For the experimental conditions used, the steady-state deuterium permeation flux in V-15Cr-5Ti is approximately 18% of the implantation flux. This is approximately 1000 times that seen in the austenitic stainless steel, PCA, and 200 times that seen in the ferritic steel, HT-9, under comparable conditions. Measurement of deuterium diffusivity in V-15Cr-5Ti using permeation break-through times indicates that D = 1.4x10 -8 exp(-0.11 eV/kT)(m 2 /s), over the temperature range 723 K to 823 K. (orig.)

  11. Permeation of deuterium implanted into V-15Cr-5Ti

    Science.gov (United States)

    Anderl, R. A.; Longhurst, G. R.; Struttmann, D. A.

    1987-02-01

    Permeation and reemission of deuterium for the vanadium alloy, V-15Cr-5Ti, was investigated using 3 keV, D 3+ ion beams from a small accelerator. The experiments consisted of measurement of the deuterium reemission and permeation rates as a function of implantation fluence for 0.5 mm thick specimens heated to temperatures from 623 K to 823 K. Implantation-side surface characterization was made by simultaneous measurements of sputtered ions with a secondary ion mass spectrometer (SIMS). For the experimental conditions used, the steady-state deuterium permeation flux in V-15Cr-5Ti is approximately 18% of the implantation flux. This is approximately 1000 times that seen in the austenitic stainless steel, PCA, and 200 times that seen in the ferritic steel, HT-9, under comparable conditions. Measurement of deuterium diffusivity in V-15Cr-5Ti using permeation break-through times indicates that D = 1.4 × 10 -8 exp( -0.11 eV/ kT) (m 2/s), over the temperature range 723 K to 823 K.

  12. A comparison of implantation-driven permeation characteristics of fusion reactor structural materials

    Science.gov (United States)

    Longhurst, G. R.; Anderl, R. A.; Struttmann, D. A.

    1986-11-01

    Implantation-driven permeation experiments have been conducted on samples of the ferritic steel HT-9, the austenitic Primary Candidate Alloy (PCA) and the vanadium alloy V-15Cr-5Ti using D 3+ ions under conditions that simulate charge-exchange neutral loading on a fusion reactor first wall. The steels all exhibited an initially intense permeation "spike" followed by an exponential decrease to low steady-state values. That spike was not evident in the V-15Cr-5Ti experiments. Steady-state permeation was highest in the vanadium alloy and lowest in the austenitic steel. Though permeation rates in the HT-9 were lower than those in V-15Cr-5Ti, permeation transients were much faster in HT-9 than in other materials tested. Sputtering of the steel surface resulted in enhanced reemission, whereas in the vanadium tests, recombination and diffusivity both appeared to diminish as the deuterium concentration rose. We conclude that for conditions comparable to those of these experiments, tritium retention and permeation loss in first wall structures made of steels will be less than in structures made of V-15Cr-5Ti.

  13. Comparison on implantation-driven permeation characteristics of fusion reactor structural materials

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R.; Anderl, R.A.; Struttmann, D.A. (Idaho National Engineering Lab., Idaho Falls)

    Implantation-driven permeation experiments have been conducted on samples of the ferritic steel HT-9, the austenitic Primary Candidate Alloy (PCA) and the vanadium alloy V-15Cr-5Ti using D{sub 3}{sup +} ions under conditions that simulate charge-exchange neutral loading on a fusion reactor first wall. The steels all exhibited an initially intense permeation spike followed by an exponential decrease to low steady-state values. That spike was not evident in the V-15Cr-5Ti experiments. Steady-state permeation was highest in the vanadium alloy and lowest in the austenitic steel. Though permeation rates in the HT-9 were lower than those in V-15Cr-5Ti, permeation transients were much faster in HT-9 than in other materials tested. Sputtering of the steel surface resulted in enhanced reemission, whereas in the vanadium tests, recombination and diffusivity both appeared to diminish as the deuterium concentration rose. We conclude that for conditions comparable to those of these experiments, tritium retention and permeation loss in first wall structures made of steels will be less than in structures made of V-15Cr-5Ti.

  14. Permeation of deuterium implanted into V-15Cr-5Ti

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A.; Longhurst, G.R.; Struttmann, D.A.

    1987-02-01

    Permeation and reemission of deuterium for the vanadium alloy, V-15Cr-5Ti, was investigated using 3 keV, D/sub 3//sup +/ ion beams from a small accelerator. The experiments consisted of measurement of the deuterium reemission and permeation rates as a function of implantation fluence for 0.5 mm thick specimens heated to temperatures from 623 K to 823 K. Implantation-side surface characterization was made by simultaneous measurements of sputtered ions with a secondary ion mass spectrometer (SIMS). For the experimental conditions used, the steady-state deuterium permeation flux in V-15Cr-5Ti is approximately 18% of the implantation flux. This is approximately 1000 times that seen in the austenitic stainless steel, PCA, and 200 times that seen in the ferritic steel, HT-9, under comparable conditions. Measurement of deuterium diffusivity in V-15Cr-5Ti using permeation break-through times indicates that D = 1.4x10/sup -8/ exp(-0.11 eVkT)(m/sup 2/s), over the temperature range 723 K to 823 K.

  15. An assessment of magnetic effects in ferromagnetic martensitic steels for use in fusion machines

    International Nuclear Information System (INIS)

    Lechtenberg, T.; Dahms, C.; Attaya, H.

    1984-01-01

    Interest in the 9-12%Cr class of martensitic stainless steels has accelerated since these materials were included in the U.S. Alloy Development for Irradiation Performance (ADIP) task funded by the Office of Fusion Energy in 1979. This program is focused on developing structural materials for fusion reactor first wall/breeding blanket components where the neutron damage is most severe. This area of a fusion reactor will be required to tolerate damage levels on the order of 110 dpa( 1 ). As a part of ADIP, study of the martensitic steels is focused on establishing the feasibility of using these materials. The interest in martensitic steels stems from their potential to tolerate high levels of radiation damage without significant degradation of material properties. Martensitic steels have a body-centered-cubic crystal structure that, unlike face-centered-cubic structure of austenitic steels, exhibits very little swelling under neutron irradiation( 2 ). One of the outstanding issues with martensitic steels is the possible parasitic stresses associated with ferromagnetic interaction with the magnetic fields. This paper is divided into two parts, the first reviews previous work on magnetic effects to the structure and plasma; the second presents new calculations of stresses on a coolant pipe in a Starfire model assumed to be made of 12Cr-1Mo steel(HT-9)

  16. Eutectic reaction analysis between TRU-50%Zr alloy fuel and HT-9 cladding, and temperature prediction of eutectic reaction under steady-state

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok

    2001-02-01

    Blanket fuel assembly for HYPER contains a bundle of pins arrayed in triangular pitch, which has hexagonal bundle structure. The reference blanket fuel pin consists of the fuel slug of TRU-50wt%Zr alloy and the cladding material of ferritic martensite steel, HT-9. Chemical interaction between fuel slug and cladding is one of the major concerns in metallic fuel rod design. The contact of metallic fuel slug and stainless steel cladding in a fuel rod forms a complex multi-component diffusion couple at elevated temperatures. The potential problem of inter-diffusion of fuel and cladding components is essentially two-fold weakening of cladding mechanical strength due to the formation of diffusion zones in the cladding, and the formation of comparatively low melting point phases in the fuel/cladding interface to develop eutectic reaction. The main components of fuel slug are composed of zirconium alloying element in plutonium matrix, including neptunium, americium and uranium additionally. Therefore basic eutectic reaction change of Pu-Fe binary system can be assessed, while it is estimated how much other elements zirconium, uranium, americium and neptunium influence on plutonium phase stability. Afterwards it is needed that eutectic reaction is verified through experimental necessarily.

  17. Eutectic reaction analysis between TRU-50%Zr alloy fuel and HT-9 cladding, and temperature prediction of eutectic reaction under steady-state

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok

    2001-02-01

    Blanket fuel assembly for HYPER contains a bundle of pins arrayed in triangular pitch, which has hexagonal bundle structure. The reference blanket fuel pin consists of the fuel slug of TRU-50wt%Zr alloy and the cladding material of ferritic martensite steel, HT-9. Chemical interaction between fuel slug and cladding is one of the major concerns in metallic fuel rod design. The contact of metallic fuel slug and stainless steel cladding in a fuel rod forms a complex multi-component diffusion couple at elevated temperatures. The potential problem of inter-diffusion of fuel and cladding components is essentially two-fold weakening of cladding mechanical strength due to the formation of diffusion zones in the cladding, and the formation of comparatively low melting point phases in the fuel/cladding interface to develop eutectic reaction. The main components of fuel slug are composed of zirconium alloying element in plutonium matrix, including neptunium, americium and uranium additionally. Therefore basic eutectic reaction change of Pu-Fe binary system can be assessed, while it is estimated how much other elements zirconium, uranium, americium and neptunium influence on plutonium phase stability. Afterwards it is needed that eutectic reaction is verified through experimental necessarily

  18. Thermal Effects on Thin Laser-Peened Ferritic-Martensitic Samples

    International Nuclear Information System (INIS)

    Caro, M; Zalesky, T; Hosemann, P; El-dasher, B S; Halsey, W G; Stuart, B

    2007-01-01

    Laser peening is emerging as a promising technique to improve the corrosion resistance of cladding material candidates for lead-cooled fast reactors (LFRs). The challenge is in the performance capability of ∼1 mm-thick fuel-pin cladding. Ferritic-martensitic (F/M) steels are foreseen as possible candidates that stand severe conditions of high dose (150 dpa), and high temperature (∼500-600 C) under the corrosive environments of molten lead or lead-bismuth. In this paper, we present the results of experiments carried on laser peened (LP) samples of F/M steels HT9, T91, EP823, as well as the austenitic material 316 L. The samples underwent a thermal treatment in an oven at 520 C, and XRD compressive stress results indicate that the F/M samples do not retain the residual stress after 2 weeks of heat treatment. The corrosion behavior in flowing lead-bismuth eutectic (LBE) at 535 C has been investigated as well. Also, irradiation experiments of LP samples to a maximum dose of ∼10 dpa are foreseen

  19. Role of grain boundary engineering in the SCC behavior of ferritic-martensitic alloy HT-9

    International Nuclear Information System (INIS)

    Gupta, G.; Ampornrat, P.; Ren, X.; Sridharan, K.; Allen, T.R.; Was, G.S.

    2007-01-01

    This paper focuses on the role of grain boundary engineering (GBE) in stress corrosion cracking (SCC) of ferritic-martensitic (F-M) alloy HT-9 in supercritical water (SCW) at 400 deg. C and 500 deg. C. Constant extension rate tensile (CERT) tests were conducted on HT-9 in as-received (AR) and coincident site lattice enhanced (CSLE) condition. Both unirradiated and irradiated specimens (irradiated with 2 MeV protons at 400 deg. C and 500 deg. C to a dose of 7 dpa) were tested. Ferritic-martensitic steel HT-9 exhibited intergranular stress corrosion cracking when subjected to CERT tests in an environment of supercritical water at 400 deg. C and 500 deg. C and also in an inert environment of argon at 500 deg. C. CSL-enhancement reduces grain boundary carbide coarsening and cracking susceptibility in both the unirradiated and irradiated condition. Irradiation enhanced coarsening of grain boundary carbides and cracking susceptibility of HT-9 for both the AR and CSLE conditions. Intergranular (IG) cracking of HT-9 results likely from fracture of IG carbides and seems consistent with the mechanism that coarser carbides worsen cracking susceptibility. Oxidation in combination with wedging stresses is the likely cause of the observed environmental enhancement of high temperature IG cracking in HT-9

  20. Development of metallic fuel fabrication - A study on the interdiffusion behavior between ternary metallic fuel and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Soo; Seol, Kyung Won; Shon, In Jin [Chonbuk National University, Chonju (Korea)

    1999-04-01

    To study a new ternary metallic fuel for liquid metal reactor, various U-Zr-X alloys have been made by induction melting. The specimens were prepared for thermal stability tests at 630 deg. C upto 5000 hours in order to estimate the decomposition of the lamellar structure. Interdiffusion studies were carried out at 700 deg. C for 200 hours for the diffusion couples assembled with U-Zr-X ternary fuel versus austenitic stainless steel D9 and martensitic stainless steel HT9, respectively, to investigate the fuel-cladding compatibility. The ternary alloy, especially U-Zr-Mo and U-Zr-Nb alloys showed relatively good thermal stability as long as 5000hrs at 630 deg. C. From the composition profiles of the interdiffusion study, Fe penetrated deeper to the fuel side than other cladding elements such as Ni and Cr, whereas U did to the cladding side of fuel elements in the fuel/D9 couples. On the contrary, the reaction layers of Fuel/HT9 couple were thinner than that of Fuel/D9 couples and were less affected by cladding element, which was believed to be due to Zr rich layer between the fuel-cladding interface. HT9 is considered to be superior to D9 and a favorable choice as a cladding material in terms of fuel-cladding compatibility. 21 refs., 24 figs., 7 tabs. (Author)

  1. Transient bowing of core assemblies in advanced liquid metal fast reactors

    International Nuclear Information System (INIS)

    Kamal, S.A.; Orechwa, Y.

    1986-01-01

    Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety

  2. Martensitic/ferritic steels as container materials for liquid mercury target of ESS

    International Nuclear Information System (INIS)

    Dai, Y.

    1996-01-01

    In the previous report, the suitability of steels as the ESS liquid mercury target container material was discussed on the basis of the existing database on conventional austenitic and martensitic/ferritic steels, especially on their representatives, solution annealed 316 stainless steel (SA 316) and Sandvik HT-9 martensitic steel (HT-9). Compared to solution annealed austenitic stainless steels, martensitic/ferritic steels have superior properties in terms of strength, thermal conductivity, thermal expansion, mercury corrosion resistance, void swelling and irradiation creep resistance. The main limitation for conventional martensitic/ferritic steels (CMFS) is embrittlement after low temperature (≤380 degrees C) irradiation. The ductile-brittle transition temperature (DBTT) can increase as much as 250 to 300 degrees C and the upper-shelf energy (USE), at the same time, reduce more than 50%. This makes the application temperature range of CMFS is likely between 300 degrees C to 500 degrees C. For the present target design concept, the temperature at the container will be likely controlled in a temperature range between 180 degrees C to 330 degrees C. Hence, CMFS seem to be difficult to apply. However, solution annealed austenitic stainless steels are also difficult to apply as the maximum stress level at the container will be higher than the design stress. The solution to the problem is very likely to use advanced low-activation martensitic/ferritic steels (LAMS) developed by the fusion materials community though the present database on the materials is still very limited

  3. Design study of 'HIBLIC-I' reactor cavity

    International Nuclear Information System (INIS)

    Fujiie, Y.

    1984-01-01

    A preliminary conceptual design of a reactor cavity for HIBLIC-1, a heavy ion fusion reactor system, was carried out. Design efforts have been concentrated mainly on the feasibility study of the physical scenario adopted and also on the system integration of the structures and components into a compact reactor cavity. The design features of the reactor are a compact reactor cavity, maximum coolant temperature up to 500 deg C, the protection of the sacrificial wall and cavity wall from radiation, the protection of the sacrificial wall from the pressure transient due to rapid heating, the selection of a ferritic steel HT-9 as the structural material and impurity control, and tritium breeding and recovery. The purpose of this paper is to describe the outline of the reactor cavity design of HIBLIC-1. The objectives of the preliminary conceptual design were to propose the idea and concept in order to constitute the physical scenario without contradiction and to find out the critical and fundamental problems to be studied in future. The cavity configuration and dynamics, tritium breeding and radiation damage, the behavior of a structural material in liquid lithium and tritium recovery are reported. (Kako, I.)

  4. Degradation of HT9 under simultaneous ion beam irradiation and liquid metal corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Frazer, D., E-mail: davefrazer@berkeley.edu [Department of Nuclear Engineering, University of California, Berkeley, Berkeley, CA United States (United States); Qvist, S. [Department of Nuclear Engineering, University of California, Berkeley, Berkeley, CA United States (United States); Department of Physics and Astronomy, Uppsala University (Sweden); Parker, S.; Krumwiede, D.L. [Department of Nuclear Engineering, University of California, Berkeley, Berkeley, CA United States (United States); Caro, M.; Tesmer, J.; Maloy, S.A.; Wang, Y.Q. [Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, NM United States (United States); Hosemann, P. [Department of Nuclear Engineering, University of California, Berkeley, Berkeley, CA United States (United States)

    2016-10-15

    A potentially promising coolant/structural material pair for a liquid-metal-cooled fast reactors is lead bismuth eutectic (LBE) coolant with the ferritic/martensitic steel HT9. The challenge of deploying LBE, however, is the corrosive environment it creates for structural materials. This corrosion can be mitigated with precise oxygen content control in the LBE to allow for the growth of passive protective oxide layers on the surface of the steel. In this paper, results are reported from the Irradiation Corrosion Experiment II (ICE-II), which allowed the simultaneous irradiation of a sample while in contact with LBE. It was found that a characteristic multilayer structure with an outer Fe{sub 3}O{sub 4} oxide and inner FeCr{sub 2}O{sub 4} spinel was grown and the oxidation was significantly larger in the irradiated region when compared to the region that was only exposed to LBE corrosion. Possible mechanisms are discussed to help understand this irradiation enhanced corrosion behavior.

  5. Comparison of implantation-driven permeation characteristics of fusion reactor structural materials

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Struttmann, D.A.

    1986-01-01

    Implantation-driven permeation experiments have been conducted on samples of the ferritic steel HT-9, the austenitic Primary Candidate Alloy (PCA) and the vanadium alloy V-15Cr-5Ti using D 3 + ions under conditions that simulate charge-exchange neutral loading on a fusion reactor first wall. The steels all exhibited an initially intense permeation ''spike'' followed by an exponential decrease to low steady-state values. That spike was not evident in the V-15Cr-5Ti experiments. Steady-state permeation was highest in the vanadium alloy and lowest in the austenitic steel. Though permeation rates in the HT-9 were lower than those in V-15Cr-5Ti, permeation transients were much faster in HT-9 than in other materials tested. Ion-beam sputtering of the surface in the steel experiments resulted in enhanced remission at the front surface, whereas in the vanadium tests, recombination and diffusivity both appeared to diminish as the deuterium concentration rose. This may be due to a phase change in the material. We conclude that for conditions comparable to those of these experiments, tritium retention and loss in first wall structures made of steels will be less than in structures made of V-15Cr-5Ti

  6. Microstructural examination of 12% Cr martensitic stainless steel after irradiation at elevated temperatures in FFTF [Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Hsu, Chen-Yih; Gelles, D.S.; Lechtenberg, T.A.

    1986-06-01

    A remelted 12% Cr martensitic stainless steel (HT-9) has been examined by transmission electron microscopy before and after irradiation in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF). The irradiation temperatures were 365,420, 520, and 600 degree C with the fluences as high as 7.3 x 10 22 n/cm 2 (E > 0.1 MeV) or 34 dpa. The extracted precipitates from each specimen were identified using x-ray microanalysis and selected area diffraction. The precipitates in the unirradiated condition were primarily M 23 C 6 carbides, which formed at martensite lath and prior austenite grain boundaries. During irradiation at elevated temperatures, small amounts of other phases formed, which were tentatively identified as the chromium-rich α', the nickel-silicon rich G-phase, and the intermetallic Chi phase. Irradiation-induced voids were observed only in specimens irradiated at 420 degree C to a dose of 34 dpa; no voids were found for specimens irradiated at 365, 520, and 600 degree C (∼11, ∼34, and ∼34 dpa). These results are not in agreement with previous experiments in that voids have not been reported in this alloy at relatively high fluence level (∼67 dpa) following irradiation in another fast-spectrum reactor (EBR.II). This is, however, the first observation following FFTF irradiation. The present results indicate that cavities can form in HT-9 at modest fluence levels even without significant generation of helium. Hence, the cavity formation in this class of ferritic alloys is not simply caused by helium generation but rather more complex mechanisms. 12 refs., 2 figs., 3 tabs

  7. Corrosion of target and structural materials in water irradiated by an 800 MeV proton beam

    International Nuclear Information System (INIS)

    Butt, D.P.; Kanner, G.S.; Lillard, R.S.

    1996-01-01

    Radiation enhanced, aqueous corrosion of solid neutron-targets such as tungsten or tantalum, or target cladding or structural materials such as superalloys and stainless steels, is a significant concern in accelerator-driven transmutation technologies. In this paper we briefly describe our current methods for control and in situ monitoring of corrosion in accelerator cooling water loops. Using floating, electrochemical impedance spectroscopy (EIS), we have measured the corrosion rates of aluminum 6061, copper, Inconel 718, and 304L stainless steel in the flow loop of a water target irradiated by a μamp, 800 MeV proton beam. Impedance spectroscopy allows us to model the corrosion process of a material as an equivalent electrical circuit. Thus the polarization resistance, which is inversely proportional to the corrosion rate, can be extracted from the frequency response of a metal specimen. During a three month period, without the use of corrosion mitigation techniques, we observed increases of several orders of magnitude in the water conductivity and the corrosion rates. The increase in corrosion is at least partially attributed to a build up of peroxide in our pseudo-closed loop system. In this paper we also briefly describe our second generation experiments, scheduled to begin in late 1996. In these experiments we plan to measure the corrosion rates of tungsten, tantalum, Inconel 718, 316L and 304L stainless steel, HT-9 austenitic stainless steel, and aluminum 5053. Two or three electrode probes of each material are being placed directly in the proton beam, in a high neutron flux region, or a significant distance from the high radiation area. We will be measuring corrosion rates, changes in pH and conductivity, and we will be establishing parameters for filtration and mitigation of corrosion. We will also discuss our ideas for making in situ measurements of water radiolysis using optical and laser diagnostic techniques