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Sample records for 12kh18n9t steel irradiated

  1. Effect of heat treatment on the structure and the mechanical and technological properties of corrosion-resistant nitrogen-bearing 0Kh16N4AFD steel for high-strength welding constructions of railway engineering

    NASA Astrophysics Data System (ADS)

    Bannykh, O. A.; Blinov, V. M.; Kostina, M. V.; Lukin, E. I.; Blinov, E. V.; Rigina, L. G.

    2015-07-01

    The problems of applying a new nitrogen-alloyed martensitic corrosion-resistant 0Kh16N4AFD steel as a promising material for manufacturing car bodies are considered. The microstructure and properties of the steel after various heat treatments have been studied. It is shown that the steel is not behind 12Kh18N9T steel in the characteristics of ductility and corrosion resistance and significantly exceeds it in the static and cyclic strengths.

  2. Effect of Friction-Induced Deformation on the Structure, Microhardness, and Wear Resistance of Austenitic Chromium—Nickel Stainless Steel Subjected to Subsequent Oxidation

    NASA Astrophysics Data System (ADS)

    Korshunov, L. G.; Chernenko, N. L.

    2016-03-01

    The effect of plastic deformation that occurs in the zone of the sliding friction contact on structural transformations in the 12Kh18N9T austenitic steel subjected to subsequent 1-h oxidation in air at temperatures of 300-800°C, as well as on its wear resistance, has been studied. It has been shown that severe deformation induced by dry sliding friction produces the two-phase nanocrystalline γ + α structure in the surface layer of the steel ~10 μm thick. This structure has the microhardness of 5.2 GPa. Subsequent oxidation of steel at temperatures of 300-500°C leads to an additional increase in the microhardness of its deformed surface layer to the value of 7.0 GPa. This is due to the active saturation of the austenite and the strain-assisted martensite (α') with the oxygen atoms, which diffuse deep into the metal over the boundaries of the γ and α' nanocrystals with an increased rate. The concentration of oxygen in the surface layer of the steel and in wear products reaches 8 wt %. The atoms of the dissolved oxygen efficiently pin dislocations in the γ and α' phases, which enhances the strength and wear resistance of the surface of the 12Kh18N9T steel. The oxidation of steel at temperatures of 550-800°C under a light normal load (98 N) results in the formation of a large number of Fe3O4 (magnetite) nanoparticles, which increase the resistance of the steel to thermal softening and its wear resistance during dry sliding friction in a pair with 40Kh13 steel. Under a heavy normal load (196 N), the toughness of 12Kh18N9T steel and, therefore, the wear resistance of its surface layer decrease due to the presence of the brittle oxide phase.

  3. Structural factors governing steel resistance during operation in corrosive media under cavitation conditions

    SciTech Connect

    Berezovskaya, V.V.

    1988-05-01

    To evaluate the effect of structural factors on the corrosion resistance of steel under cavitation, a study was made of the effect of cavitation on chromium (95Kh18) and chromium-nickel (12Kh18N9T) corrosion-resistant steels and also on maraging steel 03Kh10N5K5M3KTYuS, specially developed for the loading conditions being studied, together with the austenitic steel 03Kh23N28M3D3T. A study of the effect of martensite content on corrosion and cavitation-corrosion resistance was carried out on steels 12Kh18N9T and 95Kh18 after different treatments. Corrosion tests were conducted in solutions of H/sub 2/SO/sub 4/ and H/sub 3/PO/sub 4/ and scanning electron microscopy was used to assess phase and corrosion behavior. Maraging steels treated for maximum hardness and overaging exhibited high cavitation-corrosion resistance in acid solutions owing to high strength and resistance to microcrack initiation and propagation. It was recommended that under cavitation conditions in corrosive media, high alloy austenitic corrosion-resistant steels are substituted by maraging steels.

  4. Comminuting irradiated ferritic steel

    DOEpatents

    Bauer, Roger E.; Straalsund, Jerry L.; Chin, Bryan A.

    1985-01-01

    Disclosed is a method of comminuting irradiated ferritic steel by placing the steel in a solution of a compound selected from the group consisting of sulfamic acid, bisulfate, and mixtures thereof. The ferritic steel is used as cladding on nuclear fuel rods or other irradiated components.

  5. Irradiation effects in ferritic steels

    NASA Astrophysics Data System (ADS)

    Lechtenberg, Thomas

    1985-08-01

    Since 1979 the Alloy Development for Irradiation Performance (ADIP) task funded by the US Department of Energy has been studying the 2-12Cr class of ferritic steels to establish the feasibility of using them in fusion reactor first wall/breeding blanket (FW/B) applications. The advantages of ferritic steels include superior swelling resistance, low thermal stresses compared to austenitic stainless steels, attractive mechanical properties up to 600°C. and service histories exceeding 100 000 h. These steels are commonly used in a range of microstructural conditions which include ferritic, martensitic. tempered martensitic, bainitic etc. Throughout this paper where the term "ferritic" is used it should be taken to mean any of these microstructures. The ADIP task is studying several candidate alloy systems including 12Cr-1MoWV (HT-9), modified 9Cr-1MoVNb, and dual-phased steels such as EM-12 and 2 {1}/{4}Cr-Mo. These materials are ferromagnetic (FM), body centered cubic (bcc), and contain chromium additions between 2 and 12 wt% and molybdenum additions usually below 2%. The perceived issues associated with the application of this class of steel to fusion reactors are the increase in the ductile-brittle transition temperature (DBTT) with neutron damage, the compatibility of these steels with liquid metals and solid breeding materials, and their weldability. The ferromagnetic character of these steels can also be important in reactor design. It is the purpose of this paper to review the current understanding of these bcc steels and the effects of irradiation. The major points of discussion will be irradiation-induced or -enhanced dimensional changes such as swelling and creep, mechanical properties such as tensile strength and various measurements of toughness, and activation by neutron interactions with structural materials.

  6. Nanoindentation on ion irradiated steels

    NASA Astrophysics Data System (ADS)

    Hosemann, P.; Vieh, C.; Greco, R. R.; Kabra, S.; Valdez, J. A.; Cappiello, M. J.; Maloy, S. A.

    2009-06-01

    Radiation induced mechanical property changes can cause major difficulties in designing systems operating in a radiation environment. Investigating these mechanical property changes in an irradiation environment is a costly and time consuming activity. Ion beam accelerator experiments have the advantage of allowing relatively fast and inexpensive materials irradiations without activating the sample but do in general not allow large beam penetration depth into the sample. In this study, the ferritic/martensitic steel HT-9 was processed and heat treated to produce one specimen with a large grained ferritic microstructure and further heat treated to form a second specimen with a fine tempered martensitic lath structure and exposed to an ion beam and tested after irradiation using nanoindentation to investigate the irradiation induced changes in mechanical properties. It is shown that the HT-9 in the ferritic heat treatment is more susceptible to irradiation hardening than HT-9 after the tempered martensitic heat treatment. Also at an irradiation temperature above 550 °C no detectable hardness increase due to irradiation was detected. The results are also compared to data from the literature gained from the fast flux test facility.

  7. Susceptibility of irradiated steels to hydrogen embrittlement

    NASA Technical Reports Server (NTRS)

    Rossin, A. D.

    1968-01-01

    Investigation determined whether irradiated pressure-vessel steels 4340 and 212-B are susceptible to hydrogen embrittlement and to catastrophic failure. Hydrogen-charging conditions which completely embrittled 4340 steel had negligible effect on 212-B steel in tensile and delayed-failure tests.

  8. Neutron irradiation creep in stainless steel alloys

    NASA Astrophysics Data System (ADS)

    Schüle, Wolfgang; Hausen, Hermann

    1994-09-01

    Irradiation creep elongations were measured in the HFR at Petten on AMCR steels, on 316 CE-reference steels, and on US-316 and US-PCA steels varying the irradiation temperature between 300°C and 500°C and the stress between 25 and 300 MPa. At the beginning of an irradiation a type of "primary" creep stage is observed for doses up to 3-5 dpa after which dose the "secondary" creep stage begins. The "primary" creep strain decreases in cold-worked steel materials with decreasing stress and decreasing irradiation temperature achieving also negative creep strains depending also on the pre-treatment of the materials. These "primary" creep strains are mainly attributed to volume changes due to the formation of radiation-induced phases, e.g. to the formation of α-ferrite below about 400°C and of carbides below about 700°C, and not to irradiation creep. The "secondary" creep stage is found for doses larger than 3 to 5 dpa and is attributed mainly to irradiation creep. The irradiation creep rate is almost independent of the irradiation temperature ( Qirr = 0.132 eV) and linearly dependent on the stress. The total creep elongations normalized to about 8 dpa are equal for almost every type of steel irradiated in the HFR at Petten or in ORR or in EBR II. The negative creep elongations are more pronounced in PCA- and in AMCR-steels and for this reason the total creep elongation is slightly smaller at 8 dpa for these two steels than for the other steels.

  9. Irradiation embrittlement of neutron-irradiated low activation ferritic steels

    NASA Astrophysics Data System (ADS)

    Kayano, H.; Kimura, A.; Narui, M.; Sasaki, Y.; Suzuki, Y.; Ohta, S.

    1988-07-01

    Effects of neutron irradiation and additions of small amounts of alloying elements on the ductile-brittle transition temperature (DBTT) of three different groups of ferritic steels were investigated by means of the Charpy impact test in order to gain an insight into the development of low-activation ferritic steels suitable for the nuclear fusion reactor. The groups of ferritic steels used in this study were (1) basic 0-5% Cr ferritic steels, (2) low-activation ferritic steels which are FeCrW steels with additions of small amounts of V, Mn, Ta, Ti, Zr, etc. and (3) FeCrMo, Nb or V ferritic steels for comparison. In Fe-0-15% Cr and FeCrMo steels, Fe-3-9% Cr steels showed minimum brittleness and provided good resistance against irradiation embrittlement. Investigations on the effects of additions of trace amounts of alloying elements on the fracture toughness of low-activation ferritic steels made clear the optimum amounts of each alloying element to obtain higher toughness and revealed that the 9Cr-2W-Ta-Ti-B ferritic steel showed the highest toughness. This may result from the refinement of crystal grains and improvement of quenching characteristics caused by the complex effect of Ti and B.

  10. Neutron Irradiation Resistance of RAFM Steels

    SciTech Connect

    Gaganidze, Ermile; Dafferner, Bernhard; Aktaa, Jarir

    2008-07-01

    The neutron irradiation resistance of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 and international reference steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) have been investigated after irradiation in the Petten High Flux Reactor up to 16.3 dpa at different irradiation temperatures (250-450 deg. C). The embrittlement behavior and hardening are investigated by instrumented Charpy-V tests with sub-size specimens. Neutron irradiation-induced embrittlement and hardening of EUROFER97 was studied under different heat treatment conditions. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement vs. hardening behavior of RAFM steels within a proper model in terms of the parameter C={delta}DBTT/{delta}{sigma} indicates hardening-dominated embrittlement at irradiation temperatures below 350 deg. C with 0.17 {<=} C {<=} 0.53 deg. C/MPa. Scattering of C at irradiation temperatures above 400 deg. C indicates non hardening embrittlement. A role of He in a process of embrittlement is investigated in EUROFER97 based steels, that are doped with different contents of natural B and the separated {sup 10}B-isotope (0.008-0.112 wt.%). Testing on small scale fracture mechanical specimens for determination of quasi-static fracture toughness will be also presented in a view of future irradiation campaigns. (authors)

  11. Irradiation hardening of reduced activation martensitic steels

    NASA Astrophysics Data System (ADS)

    Kimura, A.; Morimura, T.; Narui, M.; Matsui, H.

    1996-10-01

    Irradiation response on the tensile properties of 9Cr2W steels has been investigated following FFTF/MOTA irradiations at temperatures between 646 and 873 K up to doses between 10 and 59 dpa. The largest irradiation hardening accompanied by the largest decrease in the elongation is observed for the specimens irradiated at 646 K at doses between 10 and 15 dpa. The irradiation hardening appears to saturate at a dose of around 10 dpa at the irradiation temperature. No hardening but softening was observed in the specimens irradiated at above 703 K to doses of 40 and 59 dpa. Microstructural observation by transmission electron microscope (TEM) revealed that the dislocation loops with the a<100> type Burgers vector and small precipitates which were identified to be M 6C type carbides existed after the irradiation at below 703 K. As for the void formation, the average size of voids increased with increasing irradiation temperature from 646 to 703 K. No voids were observed above 703 K. Irradiation softening was attributed to the enhanced recovery of martensitic structure under the irradiation. Post-irradiation annealing resulted in hardening by the annealing at 673 K and softening by the annealing at 873 K.

  12. Irradiation Assisted Grain Boundary Segregation in Steels

    SciTech Connect

    Lu, Zheng; Faulkner, Roy G.

    2008-07-01

    The understanding of radiation-induced grain boundary segregation (RIS) has considerably improved over the past decade. New models have been introduced and much effort has been devoted to obtaining comprehensive information on segregation from the literature. Analytical techniques have also improved so that chemical analysis of layers 1 nm thick is almost routine. This invited paper will review the major methods used currently for RIS prediction: namely, Rate Theory, Inverse Kirkendall, and Solute Drag approaches. A summary is made of the available data on phosphorus RIS in reactor pressure vessel (RPV) steels. This will be discussed in the light of the predictions of the various models in an effort to show which models are the most reliable and easy to use for forecasting P segregation behaviour in steels. A consequence of RIS in RPV steels is a radiation induced shift in the ductile to brittle transition temperature (DBTT). It will be shown how it is possible to relate radiation-induced P segregation levels to DBTT shift. Examples of this exercise will be given for RPV steels and for ferritic steels being considered for first wall fusion applications. Cr RIS in high alloy stainless steels and associated irradiation-assisted stress corrosion cracking (IASCC) will be briefly discussed. (authors)

  13. Heavy-Section Steel Irradiation Program

    SciTech Connect

    Rosseel, T.M.

    2000-04-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established.

  14. Hydrogen retention in ion irradiated steels

    SciTech Connect

    Hunn, J.D.; Lewis, M.B.; Lee, E.H.

    1998-11-01

    In the future 1--5 MW Spallation Neutron Source, target radiation damage will be accompanied by high levels of hydrogen and helium transmutation products. The authors have recently carried out investigations using simultaneous Fe/He,H multiple-ion implantations into 316 LN stainless steel between 50 and 350 C to simulate the type of radiation damage expected in spallation neutron sources. Hydrogen and helium were injected at appropriate energy and rate, while displacement damage was introduced by nuclear stopping of 3.5 MeV Fe{sup +}, 1 {micro}m below the surface. Nanoindentation measurements showed a cumulative increase in hardness as a result of hydrogen and helium injection over and above the hardness increase due to the displacement damage alone. TEM investigation indicated the presence of small bubbles of the injected gases in the irradiated area. In the current experiment, the retention of hydrogen in irradiated steel was studied in order to better understand its contribution to the observed hardening. To achieve this, the deuterium isotope ({sup 2}H) was injected in place of natural hydrogen ({sup 1}H) during the implantation. Trapped deuterium was then profiled, at room temperature, using the high cross-section nuclear resonance reaction with {sup 3}He. Results showed a surprisingly high concentration of deuterium to be retained in the irradiated steel at low temperature, especially in the presence of helium. There is indication that hydrogen retention at spallation neutron source relevant target temperatures may reach as high as 10%.

  15. Mechanical properties of irradiated 9Cr-2WVTa steel

    SciTech Connect

    Klueh, R.L.; Alexander, D.J.; Rieth, M.

    1998-09-01

    An Fe-9Cr-2W-0.25V-0.07Ta-0.1C (9Cr-2WVTa) steel has excellent strength and impact toughness before and after irradiation in the Fast Flux Test Facility and the High Flux Reactor (HFR). The ductile-brittle transition temperature (DBTT) increased only 32 C after 28 dpa at 365 C in FFTF, compared to a shift of {approx}60 C for a 9Cr-2WV steel--the same as the 9Cr-2WVTa steel but without tantalum. This difference occurred despite the two steels having similar tensile but without tantalum. This difference occurred despite the two steels having similar tensile properties before and after irradiation. The 9Cr-2WVTa steel has a smaller prior-austenite grain size, but otherwise microstructures are similar before irradiation and show similar changes during irradiation. The irradiation behavior of the 9Cr-2WVTa steel differs from the 9Cr-2WV steel and other similar steels in two ways: (1) the shift in DBTT of the 9Cr-2WVTa steel irradiated in FFTF does not saturate with fluence by {approx}28 dpa, whereas for the 9Cr-2WV steel and most similar steels, saturation occurs at <10 dpa, and (2) the shift in DBTT for 9Cr-2WVTa steel irradiated in FFTF and HFR increased with irradiation temperature, whereas it decreased for the 9Cr-2WV steel, as it does for most similar steels. The improved properties of the 9Cr-2WVTa steel and the differences with other steels were attributed to tantalum in solution.

  16. Tensile behavior of irradiated manganese-stabilized stainless steel

    SciTech Connect

    Klueh, R.L.

    1996-10-01

    Tensile tests were conducted on seven experimental, high-manganese austenitic stainless steels after irradiation up to 44 dpa in the FFTF. An Fe-20Mn-12Cr-0.25C base composition was used, to which various combinations of Ti, W, V, B, and P were added to improve strength. Nominal amounts added were 0.1% Ti, 1% W, 0.1% V, 0.005% B, and 0.03% P. Irradiation was carried out at 420, 520, and 600{degrees}C on the steels in the solution-annealed and 20% cold-worked conditions. Tensile tests were conducted at the irradiation temperature. Results were compared with type 316 SS. Neutron irradiation hardened all of the solution-annealed steels at 420, 520, and 600{degrees}C, as measured by the increase in yield stress and ultimate tensile strength. The steel to which all five elements were added to the base composition showed the least amount of hardening. It also showed a smaller loss of ductility (uniform and total elongation) than the other steels. The total and uniform elongations of this steel after irradiation at 420{degrees}C was over four times that of the other manganese-stabilized steels and 316 SS. There was much less difference in strength and ductility at the two higher irradiation temperatures, where there was considerably less hardening, and thus, less loss of ductility. In the cold-worked condition, hardening occured only after irradiation at 420{degrees}C, and there was much less difference in the properties of the steels after irradiation. At the 420{degrees}C irradiation temperature, most of the manganese-stabilized steels maintained more ductility than the 316 SS. After irradiation at 420{degrees}C, the temperature of maximum hardening, the steel to which all five of the elements were added had the best uniform elongation.

  17. Microstructural analysis of neutron-irradiated martensitic steels

    NASA Astrophysics Data System (ADS)

    Kai, J. J.; Klueh, R. L.

    1996-06-01

    Four martensitic steels for fusion applications were examined by transmission electron microscopy after irradiation in the Fast Flux Test Facility (FFTF) at 420°C to 7.8 X 10 26 n/m 2 ( E > 0.1 MeV), about 35 dpa. There were two commercial steels, 9Cr-IMoVNb and 12Cr-1MoVW, and two experimental reduced-activation steels, 9Cr-2WV and 9Cr-2WVTa. Before irradiation, the tempered martensite microstructures of the four steels contained a high dislocation density, and the major precipitate was M 23C 6 carbide, with few MC carbides. Irradiation caused minor changes in these precipitates. Voids were found in all irradiated specimens, but swelling remained below 1%, with the 9Cr-1MoVNb having the highest void density. Although the 12Cr-IMoVW steel showed the best swelling resistance, it also contained the highest density of radiation-induced new phases, which were identified as chi-phase and possibly α'. Radiation-induced chi-phase was also observed in the 9Cr-1MoVNb steel. The two reduced-activation steels showed very stable behavior under irradiation: a high density of dislocation loops replaced the original high dislocation density; moderate void swelling occurred, and no new phase formed. The differences in microstructural evolution of the steels can explain some of the mechanical properties observations made in these steels.

  18. Weldability of neutron irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Asano, Kyoichi; Nishimura, Seiji; Saito, Yoshiaki; Sakamoto, Hiroshi; Yamada, Yuji; Kato, Takahiko; Hashimoto, Tsuneyuki

    1999-01-01

    Degradation of weldability in neutron irradiated austenitic stainless steel is an important issue to be addressed in the planning of proactive maintenance of light water reactor core internals. In this work, samples selected from reactor internal components which had been irradiated to fluence from 8.5 × 10 22 to 1.4 × 10 26 n/m 2 ( E > 1 MeV) corresponding to helium content from 0.11 to 103 appm, respectively, were subjected to tungsten inert gas arc (TIG) welding with heat input ranged 0.6-16 kJ/cm. The weld defects were characterized by penetrant test and cross-sectional metallography. The integrity of the weld was better when there were less helium and at lower heat input. Tensile properties of weld joint containing 0.6 appm of helium fulfilled the requirement for unirradiated base metal. Repeated thermal cycles were found to be very hazardous. The results showed the combination of material helium content and weld heat input where materials can be welded with little concern to invite cracking. Also, the importance of using properly selected welding procedures to minimize thermal cycling was recognized.

  19. Hydrogen transport through stainless steel under plasma irradiation

    NASA Astrophysics Data System (ADS)

    Airapetov, A. A.; Begrambekov, L. B.; Kaplevsky, A. S.; Sadovskiy, Ya A.

    2016-01-01

    The paper presents the results of investigation of gas exchange through stainless steel surface of the plasma chamber under irradiation with hydrogen atoms in oxygen atmosphere or oxygen contaminated hydrogen plasma. Dependence of this process on various irradiation parameters, such as the metal temperature, energy of irradiating ions, gas composition of plasma are studied. It is shown, that desorption from stainless steel is activated with the increase of the plasma chamber walls temperature and energy of irradiating ions. Hydrogen release occurs also under irradiation of the walls by helium and argon plasmas added with oxygen, however the amount of released hydrogen is several times lower than in the case of irradiation with oxygen contaminated deuterium plasma.

  20. Defect microstructures in neutron-irradiated copper and stainless steel

    SciTech Connect

    Zinkle, S.J.; Sindelar, R.L.

    1987-09-01

    The defect microstructures of copper and type 304L austenitic stainless steel have been examined following neutron irradiation under widely different conditions. Less than 0.2% of the defect clusters in steel irradiated at 120/sup 0/C with moderated fission neutrons were resolvable as stacking fault tetrahedra (SFT). The fraction of defect clusters identified as SFT in copper varied from approx.10% for a low-dose 14-MeV neutron irradiation at 25/sup 0/C to approx.50% for copper irradiated to 1.3 dpa in a moderated fission spectrum at 182/sup 0/C. The mean cluster size in copper was about 2.6 nm for both cases, despite the large differences in irradiation conditions. The mean defect cluster size in the irradiated steel was about 1.8 nm. The absence of SFT in stainless steel may be due to the generation of 35 appm He during the irradiation, which caused the vacancies to form helium-filled cavities instead of SFT. 20 refs.

  1. Irradiation-induced precipitation modelling of ferritic steels

    NASA Astrophysics Data System (ADS)

    Yin, You Fa; Faulkner, Roy G.; Lu, Zheng

    2009-06-01

    In high strength low alloy (HSLA) steels typically used in reactor pressure vessels (RPV), irradiation-induced microstructure changes affect the performance of the components. One such change is precipitation hardening due to the formation of solute clusters and/or precipitates which form as a result of irradiation-enhanced solute diffusion and thermodynamic stability changes. The other is irradiation-enhanced tempering which is a result of carbide coarsening due to irradiation-enhanced carbon diffusion. Both effects have been studied using a recently developed Monte Carlo based precipitation kinetics simulation technique and modelling results are compared with experimental measurements. Good agreements have been achieved.

  2. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    NASA Astrophysics Data System (ADS)

    Slugeň, Vladimír; Pecko, Stanislav; Sojak, Stanislav

    2016-01-01

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1-2 vacancies with relatively small contribution (with intensity on the level of 20-40 %) were observed in "as-received" steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2-3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  3. Evaluation of irradiation hardening of proton irradiated stainless steels by nanoindentation

    NASA Astrophysics Data System (ADS)

    Yabuuchi, Kiyohiro; Kuribayashi, Yutaka; Nogami, Shuhei; Kasada, Ryuta; Hasegawa, Akira

    2014-03-01

    Ion irradiation experiments are useful for investigating irradiation damage. However, estimating the irradiation hardening of ion-irradiated materials is challenging because of the shallow damage induced region. Therefore, the purpose of this study is to prove usefulness of nanoindentation technique for estimation of irradiation hardening for ion-irradiated materials. SUS316L austenitic stainless steel was used and it was irradiated by 1 MeV H+ ions to a nominal displacement damage of 0.1, 0.3, 1, and 8 dpa at 573 K. The irradiation hardness of the irradiated specimens were measured and analyzed by Nix-Gao model. The indentation size effect was observed in both unirradiated and irradiated specimens. The hardness of the irradiated specimens changed significantly at certain indentation depths. The depth at which the hardness varied indicated that the region deformed by the indenter had reached the boundary between the irradiated and unirradiated regions. The hardness of the irradiated region was proportional to the inverse of the indentation depth in the Nix-Gao plot. The bulk hardness of the irradiated region, H0, estimated by the Nix-Gao plot and Vickers hardness were found to be related to each other, and the relationship could be described by the equation, HV = 0.76H0. Thus, the nanoindentation technique demonstrated in this study is valuable for measuring irradiation hardening in ion-irradiated materials.

  4. Corrosion of stainless steel for HLW containers under gamma irradiation

    SciTech Connect

    Osada, K.; Muraoka, S.

    1993-12-31

    The corrosion behavior of type 304 stainless steel was studied under gamma irradiation as part of the evaluation for the long-term durability of high-level radioactive waste (HLW) disposal containers. Gamma rays, generated from fission products in high-level radioactive waste, are considered to change the environment around the canisters and overpacks. The redox potentials for NaCl solutions and corrosion potentials of stainless steel were measured to consider the effects of gamma irradiation, by using an electrochemical method. The pitting potentials of stainless steel for NaCl solutions were also measured to examine the pitting corrosion under gamma irradiation. As a result of this experiment, it is concluded that the oxidizing properties as a result of the formation of H{sub 2}O{sub 2} and H{sub 2} produced by gamma irradiation depended on the concentration of Cl{sup -}, and that the strength of oxidizing properties of 1M (mol{center_dot}dm{sup -3}) NaCl solution was particularly high, and the pitting corrosion as found for 1M NaCl solution under gamma irradiation at the dose rate of 2.6{times}10{sup 2} C/kg{center_dot}h (1.0{times}10{sup 6} R/h) at 60{degrees}C, by using an electrochemical method.

  5. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  6. The Irradiation Performance and Microstructural Evolution in 9Cr-2W Steel Under Ion Irradiation

    NASA Astrophysics Data System (ADS)

    Alsagabi, Sultan; Charit, Indrajit; Pasebani, Somayeh

    2016-02-01

    Grade 92 steel (9Cr-2W) is a ferritic-martensitic steel with good mechanical and thermal properties. It is being considered for structural applications in Generation IV reactors. Still, the irradiation performance of this alloy needs more investigation as a result of the limited available data. The irradiation performance investigation of Grade 92 steel would contribute to the understanding of engineering aspects including feasibility of application, economy, and maintenance. In this study, Grade 92 steel was irradiated by iron ion beam to 10, 50, and 100 dpa at 30 and 500 °C. In general, the samples exhibited good radiation damage resistance at these testing parameters. The radiation-induced hardening was higher at 30 °C with higher dislocation density; however, the dislocation density was less pronounced at higher temperature. Moreover, the irradiated samples at 30 °C had defect clusters and their density increased at higher doses. On the other hand, dislocation loops were found in the irradiated sample at 50 dpa and 500 °C. Further, the irradiated samples did not show any bubble or void.

  7. Characterization of Irradiated Nanostructured Ferritic Steels

    SciTech Connect

    Bentley, James; Hoelzer, David T; Tanigawa, H.; Yamamoto, T.; Odette, George R.

    2007-01-01

    The past decade has seen the development of a new class of mechanically alloyed (MA) ferritic steels with outstanding mechanical properties that come, at least in part, from the presence of high concentrations (>10{sup 23} m{sup -3}) of Ti-, Y-, and O-enriched nanoclusters (NC). From the outset, there has been much interest in their potential use for applications to fission and proposed fusion reactors, not only because of their attractive high-temperature strength, but also because the presence of NC may result in a highly radiation-resistant material by efficiently trapping point defects to enhance recombination. Of special interest for fusion applications is the potential of NC to trap transmutation-produced He in high concentrations of small cavities, rather than in fewer but larger cavities that lead to greater radiation-induced swelling and other degraded properties.

  8. Dislocation loop evolution under ion irradiation in austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Etienne, A.; Hernández-Mayoral, M.; Genevois, C.; Radiguet, B.; Pareige, P.

    2010-05-01

    A solution annealed 304 and a cold worked 316 austenitic stainless steels were irradiated from 0.36 to 5 dpa at 350 °C using 160 keV Fe ions. Irradiated microstructures were characterized by transmission electron microscopy (TEM). Observations after irradiation revealed the presence of a high number density of Frank loops. Size and number density of Frank loops have been measured. Results are in good agreement with those observed in the literature and show that ion irradiation is able to simulate dislocation loop microstructure obtained after neutron irradiation. Experimental results and data from literature were compared with predictions from the cluster dynamic model, MFVIC (Mean Field Vacancy and Interstitial Clustering). It is able to reproduce dislocation loop population for neutron irradiation. Effects of dose rate and temperature on the loop number density are simulated by the model. Calculations for ion irradiations show that simulation results are consistent with experimental observations. However, results also show the model limitations due to the lack of accurate parameters.

  9. Intergranular stress distributions in polycrystalline aggregates of irradiated stainless steel

    NASA Astrophysics Data System (ADS)

    Hure, J.; El Shawish, S.; Cizelj, L.; Tanguy, B.

    2016-08-01

    In order to predict InterGranular Stress Corrosion Cracking (IGSCC) of post-irradiated austenitic stainless steel in Light Water Reactor (LWR) environment, reliable predictions of intergranular stresses are required. Finite elements simulations have been performed on realistic polycrystalline aggregate with recently proposed physically-based crystal plasticity constitutive equations validated for neutron-irradiated austenitic stainless steel. Intergranular normal stress probability density functions are found with respect to plastic strain and irradiation level, for uniaxial loading conditions. In addition, plastic slip activity jumps at grain boundaries are also presented. Intergranular normal stress distributions describe, from a statistical point of view, the potential increase of intergranular stress with respect to the macroscopic stress due to grain-grain interactions. The distributions are shown to be well described by a master curve once rescaled by the macroscopic stress, in the range of irradiation level and strain considered in this study. The upper tail of this master curve is shown to be insensitive to free surface effect, which is relevant for IGSCC predictions, and also relatively insensitive to small perturbations in crystallographic texture, but sensitive to grain shapes.

  10. Microstructure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel

    SciTech Connect

    Ahmad Alsabbagh; Apu Sarkar; Brandon Miller; Jatuporn Burns; Leah Squires; Douglas Porter; James I. Cole; K. L. Murty

    2014-10-01

    Neutron irradiation effects on ultra-fine grain (UFG) low carbon steel prepared by equal channel angular pressing (ECAP) has been examined. Counterpart samples with conventional grain (CG) sizes have been irradiated alongside with the UFG ones for comparison. Samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 1.24 dpa. Atom probe tomography revealed manganese, silicon-enriched clusters in both ECAP and CG steel after neutron irradiation. X-ray quantitative analysis showed that dislocation density in CG increased after irradiation. However, no significant change was observed in UFG steel revealing better radiation tolerance.

  11. Evaluation of weld crack susceptibility for neutron irradiated stainless steels

    NASA Astrophysics Data System (ADS)

    Suzuki, T.; Kohyama, A.; Hirose, T.; Narui, M.

    In order to clarify the mechanisms of weld cracking, especially for heat affected zone cracking in heavily neutron irradiated stainless steels and to establish a measure to evaluate crack susceptibility, a mini-sized Varestraint (variable restraint) test machine for hot laboratory operation was designed and fabricated. This unique PIE facility was successfully applied in the hot laboratory of IMR Oarai Branch of Tohoku University. The maximum restraint applied was 4% at the surface of the specimen. Specimen surface morphology and specimen microstructures were inspected by video microscope, SEM and TEM. Under the 2% surface restraint condition, clear formation of heat affected zone (HAZ) crack was observed for the case of neutron irradiation to produce 0.5 appm He and of 2.4 kJ heat input by TIG.

  12. Stress corrosion cracking on irradiated 316 stainless steel

    NASA Astrophysics Data System (ADS)

    Furutani, Gen; Nakajima, Nobuo; Konishi, Takao; Kodama, Mitsuhiro

    2001-02-01

    Tests on irradiation-assisted stress corrosion cracking (IASCC) were carried out by using cold-worked (CW) 316 stainless steel (SS) in-core flux thimble tubes which were irradiated up to 5×10 26 n/m 2 ( E>0.1 MeV) at 310°C in a Japanese PWR. Unirradiated thimble tube was also tested for comparison with irradiated tubes. Mechanical tests such as the tensile, hardness tests and metallographic observations were performed. The susceptibility to SCC was examined by the slow strain rate test (SSRT) under PWR primary water chemistry condition and compositional analysis on the grain boundary segregation was made. Significant changes in the mechanical properties due to irradiation such as a remarkable increase of strength and hardness, and a considerable reduction of elongation were seen. SSRT results revealed that the intergranular fracture ratio (%IGSCC) increased as dissolved hydrogen (DH) increased. In addition, SSRT results in argon gas atmosphere showed a small amount of intergranular cracking. The depletion of Fe, Cr, Mo and the enrichment of Ni and Si were observed in microchemical analyses on the grain boundary.

  13. Mechanical properties of neutron-irradiated nickel-containing martensitic steels: I. Experimental study

    NASA Astrophysics Data System (ADS)

    Klueh, R. L.; Hashimoto, N.; Sokolov, M. A.; Shiba, K.; Jitsukawa, S.

    2006-10-01

    Tensile and Charpy specimens of 9Cr-1MoVNb (modified 9Cr-1Mo) and 12Cr-1MoVW (Sandvik HT9) steels and these steels doped with 2% Ni were irradiated at 300 and 400 °C in the High Flux Isotope Reactor (HFIR) up to ≈12 dpa and at 393 °C in the Fast Flux Test Facility (FFTF) to ≈15 dpa. In HFIR, a mixed-spectrum reactor, ( n, α) reactions of thermal neutrons with 58Ni produce helium in the steels. Little helium is produced during irradiation in FFTF. After HFIR irradiation, the yield stress of all steels increased, with the largest increases occurring for nickel-doped steels. The ductile-brittle transition temperature (DBTT) increased up to two times and 1.7 times more in steels with 2% Ni than in those without the nickel addition after HFIR irradiation at 300 and 400 °C, respectively. Much smaller differences occurred between these steels after irradiation in FFTF. The DBTT increases for steels with 2% Ni after HFIR irradiation were 2-4 times greater than after FFTF irradiation. Results indicated there was hardening due to helium in addition to hardening by displacement damage and irradiation-induced precipitation.

  14. Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

    SciTech Connect

    Stoller, R.E.

    1987-12-01

    The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and altered mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs.

  15. Irradiation assisted stress corrosion cracking of austenitic stainless steels

    SciTech Connect

    Was, G.S.; Atzmon, M.

    1990-06-01

    Samples of ultra high purity stainless steel have been fabricated into 2mm {times} 2mm rectangular bars and irradiated to one dpa ({approximately}l {times} 10{sup 19} p{sup +}/cm{sup 2}) using 3.4 MeV protons (>20{mu}A) while controlling the sample temperature at 400{degree}C. Samples are pressed onto a water-cooled and electrically heated copper block with a thin layer of Sn in between to improve thermal conductivity. The irradiation produced a significant prompt radiation field but sample activation was limited to {beta}-decay and this decayed rapidly in less than 48 h. Samples were hydrogen charged and strained at slow rates at {minus}30{degree}C insitu in the Auger electron spectrometer to successfully fracture several samples intergranularly for grain boundary composition analysis. An ultra-high purity (UHP) alloy of Fe-19Cr-9Ni was irradiated to 1 dpa at 400C {plus minus} 5C and 7 {times} 10{sup {minus}9} torr in the tandem accelerator of the Michigan Ion Beam Laboratory, resulting in a dislocation network density of 1.8 {times} 10{sup 9} cm{sup 2} and a dislocation loop density of 7 {times} 10{sup 16} cm{sup {minus}3} along with the dissolution of small precipitates present in the unirradiated sample. EPR experiments on the UHP irradiated alloy showed no significant increase in charge passed upon reactivation, over an unirradiated sample experiencing the same thermal history. An SCC waterloop and autoclave system has been completed and a sample has been designed to measure the susceptibility of the irradiated microstructure as compared to the unirradiated microstructure.

  16. Tensile properties of CLAM steel irradiated up to 20.1 dpa in STIP-V

    NASA Astrophysics Data System (ADS)

    Ge, Hongen; Peng, Lei; Dai, Yong; Huang, Qunying; Ye, Minyou

    2016-01-01

    Specimens of China low activation martensitic steel (CLAM) were irradiated in the fifth experiment of SINQ Target Irradiation Program (STIP-V) up to 20.1 dpa/1499 appm He/440 °C. Tensile tests were performed at room temperature (R.T) and irradiation temperatures (Tirr) in the range of 25-450 °C. The tensile results demonstrated strong effect of irradiation dose and irradiation temperature on hardening and embrittlement. With Tirr below ˜314 °C, CLAM steel specimens tested at R.T and Tirr showed similar evolution trend with irradiation dose, compared to other reduced activation ferritic/martensitic (RAFM) steels in similar irradiation conditions. At higher Tirr above ˜314 °C, it is interesting that the hardening effect decreases and the ductility seems to recover, probably due to a strong effect of high irradiation temperature.

  17. Proton irradiation creep of FM steel T91

    NASA Astrophysics Data System (ADS)

    Xu, Cheng; Was, Gary S.

    2015-04-01

    Ferritic-martensitic (FM) steel T91 was subjected to irradiation with 3 MeV protons while under load at stresses of 100-200 MPa, temperatures between 400 °C and 500 °C, and dose rates between 1.4 × 10-6 dpa/s and 5 × 10-6 dpa/s to a total dose of less than 1 dpa. Creep behavior was analyzed for parametric dependencies. The temperature dependence was found to be negligible between 400 °C and 500 °C, and the dose rate dependence was observed to be linear. Creep rate was proportional to stress at low stress values and varied with stress to the power 14 above 160 MPa. The large stress exponent of the proton irradiation creep experiments under high stress suggested that dislocation glide was driving both thermal and irradiation creep. Microstructure observations of anisotropic dislocation loops also contributed to the total creep strain. After subtracting the power law creep and anisotropic dislocation loop contributions, the remaining creep strain was accounted for by dislocation climb enabled by stress induced preferential absorption (SIPA) and preferential dislocation glide (PAG).

  18. Effects of neutron irradiation on microstructural evolution in candidate low activation ferritic steels

    NASA Astrophysics Data System (ADS)

    Kohno, Yutaka; Kohyama, Akira; Yoshino, Masahiko; Asakura, Kentaro

    1994-09-01

    Fe-(2.25-12)Cr-2W-V, Ta low activation ferritic steels (JLF series steels) were developed in the fusion materials development program of Japanese universities. Microstructural observations, including precipitation response, were performed after neutron irradiation in the FFTF/MOTA. The preirradiation microstructure was stable after irradiation at low temperature (< 683 K). Recovery of martensitic lath structure and coarsening of precipitates took place above 733 K. Precipitates observed after irradiation were the same as those in unirradiated materials in 7-9Cr steels, and no irradiation induced phase was identified. The irradiation induced shift in DBTT in the 9Cr-2W steel proved to be very small which is a reflection of stable precipitation response in these steels. A high density of fine α' precipitates was observed in the 12Cr steel which might be responsible for the large irradiation hardening found in the 12Cr steel. Void formation was observed in 7-9Cr steels irradiated at 683 K, but the amount of void swelling was very small.

  19. Effect of heat treatment and irradiation temperature on impact behavior of irradiated reduced-activation ferritic steels

    SciTech Connect

    Klueh, R.L.; Alexander, D.J.

    1998-03-01

    Charpy tests were conducted on eight normalized-and-tempered reduced-activation ferritic steels irradiated in two different normalized conditions. Irradiation was conducted in the Fast Flux Test Facility at 393 C to {approx}14 dpa on steels with 2.25, 5, 9, and 12% Cr (0.1% C) with varying amounts of W, V, and Ta. The different normalization treatments involved changing the cooling rate after austenitization. The faster cooling rate produced 100% bainite in the 2.25 Cr steels, compared to duplex structures of bainite and polygonal ferrite for the slower cooling rate. For both cooling rates, martensite formed in the 5 and 9% Cr steels, and martensite with {approx}25% {delta}-ferrite formed in the 12% Cr steel. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy. The difference in microstructure in the low-chromium steels due to the different heat treatments had little effect on properties. For the high-chromium martensitic steels, only the 5 Cr steel was affected by heat treatment. When the results at 393 C were compared with previous results at 365 C, all but a 5 Cr and a 9 Cr steel showed the expected decrease in the shift in DBTT with increasing temperature.

  20. Irradiation hardening of ODS ferritic steels under helium implantation and heavy-ion irradiation

    NASA Astrophysics Data System (ADS)

    Zhang, Hengqing; Zhang, Chonghong; Yang, Yitao; Meng, Yancheng; Jang, Jinsung; Kimura, Akihiko

    2014-12-01

    Irradiation hardening of ODS ferritic steels after multi-energy He-ion implantation, or after irradiation with energetic heavy ions including Xe and Bi-ions was investigated with nano-indentation technique. Three kinds of high-Cr ODS ferritic steels including the commercial MA956 (19Cr-3.5Al), the 16Cr-0.1Ti and the 16Cr-3.5Al-0.1Zr were used. Data of nano-hardness were analyzed with an approach based on Nix-Gao model. The depth profiles of nano-hardness can be understood by the indentation size effect (ISE) in specimens of MA956 implanted with multi-energy He-ions or irradiated with 328 MeV Xe ions, which produced a plateau damage profile in the near-surface region. However, the damage gradient overlaps the ISE in the specimens irradiated with 9.45 Bi ions. The dose dependence of the nano-hardness shows a rapid increase at low doses and a slowdown at higher doses. An 1/2-power law dependence on dpa level is obtained. The discrepancy in nano-hardness between the helium implantation and Xe-ion irradiation can be understood by using the average damage level instead of the peak dpa level. Helium-implantation to a high dose (7400 appm/0.5 dpa) causes an additional hardening, which is possibly attributed to the impediment of motion dislocations by helium bubbles formed in high concentration in specimens.

  1. A study on the microstructure and mechanical property of proton irradiated A508-3 steel

    NASA Astrophysics Data System (ADS)

    Li, Xiao-hong; Lei, Jing; Shu, Guo-gang; Wan, Qiang-mao

    2015-05-01

    Transmission electron microscopy and the nanoindentation technique were employed to study the dislocation loops and hardening induced in proton irradiated A508-3 steel. The A508-3 steel specimens were irradiated to the dose of 0.054, 0.163, 0.271 dpa at room temperature (RT), 0.163 pa at 250 °C and 0.163, 0.271 dpa at 290 °C. The effect of dose and temperature on the dislocation loops and irradiation hardening was investigated. The results indicated that the dislocation loops were formed in proton irradiated A508-3 steel. The size and number density generally increased with increasing dose at RT. When the irradiation temperature changed from RT to 290 °C, the loop size increased and the loop number density decreased. The irradiation hardening increased with dose. The effect of temperature on the irradiation induced hardening was discussed.

  2. Microstructure and nanoindentation of the CLAM steel with nanocrystalline grains under Xe irradiation

    NASA Astrophysics Data System (ADS)

    Chang, Yongqin; Zhang, Jing; Li, Xiaolin; Guo, Qiang; Wan, Farong; Long, Yi

    2014-12-01

    This work presents an early look at irradiation effects on China low activation martensitic (CLAM) steel with nanocrystalline grains (NC-CLAM steels) under 500 keV Xe-ion bombardment at room temperature to doses up to 5.3 displacements per atom (dpa). The microstructure in the topmost region of the steel is composed of nanocrystalline grains with an average diameter of 13 nm. As the samples were implanted at low dose, the nanocrystalline grains had martensite lath structure, and many dislocations and high density bubbles were introduced into the NC-CLAM steels. As the irradiation dose up to 5.3 dpa, a tangled dislocation network exists in the lath region, and the size of the bubbles increases. X-ray diffraction results show that the crystal quality decreases after irradiation, although the nanocrystals obviously coarsen. Grain growth under irradiation may be ascribed to the direct impact of the thermal spike on grain boundaries in the NC-CLAM steels. In irradiated samples, a compressive stress exists in the surface layer because of grain growth and irradiation-introduced defects, while the irradiation introduced grain-size coarsening and defects gradients from the surface to matrix result in a tensile stress in the irradiated NC-CLAM steels. Nanoindentation was used to estimate changes in mechanical properties during irradiation, and the results show that the hardness of the NC-CLAM steels increases with increasing irradiation dose, which was ascribed to the competition between the grain boundaries and the irradiation-introduced defects.

  3. Effects of hydrogen isotopes in the irradiation damage of CLAM steel

    NASA Astrophysics Data System (ADS)

    Zhao, M. Z.; Liu, P. P.; Zhu, Y. M.; Wan, F. R.; He, Z. B.; Zhan, Q.

    2015-11-01

    The isotope effect of hydrogen in irradiation damage plays an important role in the development of reduced activation Ferritic/Martensitic steels in nuclear reactors. The evolutions of microstructures and mechanical properties of China low active martensitic (CLAM) steel subjected to hydrogen and deuterium ions irradiation are studied comparatively. Under the same irradiation conditions, larger size and smaller density of dislocation loops are generated by deuterium ion than by hydrogen ion. Irradiation hardening occurs under the ion irradiation and the hardening induced by hydrogen ion is higher than by deuterium ion. Moreover, the coarsening of M23C6 precipitates is observed, which can be explained by the solute drag mechanisms. It turns out that the coarsening induced by deuterium ion irradiation is more distinct than by hydrogen ion irradiation. No distinct variations for the compositions of M23C6 precipitates are found by a large number of statistical data after hydrogen isotopes irradiation.

  4. Mechanical property changes of low activation ferritic/martensitic steels after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Kohno, Y.; Kohyama, A.; Hirose, T.; Hamilton, M. L.; Narui, M.

    Mechanical property changes of Fe- XCr-2W-0.2V,Ta ( X: 2.25-12) low activation ferritic/martensitic steels including Japanese Low Activation Ferritic/martensitic (JLF) steels and F82H after neutron irradiation were investigated with emphasis on Charpy impact property, tensile property and irradiation creep properties. Dose dependence of ductile-to-brittle transition temperature (DBTT) in JLF-1 (9Cr steel) irradiated at 646-700 K increased with irradiation up to 20 dpa and then decreased with further irradiation showing highest DBTT of 260 K at 20 dpa. F82H showed similar dose dependence in DBTT to JLF-1 with higher transition temperature than that of JLF-1 at the same displacement damage. Yield strength in JLF steels and F82H showed similar dose dependence to that of DBTT. Yield strength increased with irradiation up to 15-20 dpa and then decreased to saturate above about 40 dpa. Irradiation hardening in 7-9%Cr steels (JLF-1, JLF-3, F82H) were observed to be smaller than those in steels with 2.25%Cr (JLF-4) or 12%Cr (JLF-5). Dependences of creep strain on applied hoop stress and neutron fluence were measured to be 1.5 and 1, respectively. Temperature dependence of creep coefficient showed a maximum at about 700 K which was caused by irradiation induced void formation or irradiation enhanced creep deformation. Creep coefficient of F82H was larger than those of JLF steels above 750 K. This was considered to be caused by the differences in N and Ta concentration between F82H and JLF steels.

  5. Tensile and charpy impact properties of irradiated reduced-activation ferritic steels

    SciTech Connect

    Klueh, R.L.; Alexander, D.J.

    1996-10-01

    Tensile tests were conducted on eight reduced-activation Cr-W steels after irradiation to 15-17 and 26-29 dpa, and Charpy impact tests were conducted on the steels irradiated to 26-29 dpa. Irradiation was in the Fast Flux Test Facility at 365{degrees}C on steels containing 2.25-12% Cr, varying amounts of W, V, and Ta, and 0.1%C. Previously, tensile specimens were irradiated to 6-8 dpa and Charpy specimens to 6-8, 15-17, and 20-24 dpa. Tensile and Charpy specimens were also thermally aged to 20000 h at 365{degrees}C. Thermal aging had little effect on the tensile behavior or the ductile-brittle transition temperature (DBTT), but several steels showed a slight increase in the upper-shelf energy (USE). After {approx}7 dpa, the strength of the steels increased and then remained relatively unchanged through 26-29 dpa (i.e., the strength saturated with fluence). Post-irradiation Charpy impact tests after 26-29 dpa showed that the loss of impact toughness, as measured by an increase in DBTT and a decrease in the USE, remained relatively unchanged from the values after 20-24 dpa, which had been relatively unchanged from the earlier irradiations. As before, the two 9Cr steels were the most irradiation resistant.

  6. Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

    NASA Astrophysics Data System (ADS)

    Gelles, D. S.

    1990-05-01

    Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development.

  7. Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology Program Series 4 and 5)

    SciTech Connect

    Berggren, R.G.; McGowan, J.J.; Menke, B.H.; Nanstad, R.K.; Thoms, K.R.

    1984-01-01

    Multiple testing is done at two laboratories of typical nuclear pressure vessel materials (both irradiated and unirradiated) and statistical analyses of the test results. Multiple tests are conducted at each of several test temperatures for each material, standard deviations are determined, and results from the two laboratories are compared. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (current practice welds). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds.

  8. Embrittlement of Cr-Mo steels after low fluence irradiation in HFIR

    SciTech Connect

    Klueh, R.L.; Alexander, D.J.

    1995-04-01

    The goal of this work is the determination of the possible effect of the simultaneous formation of helium and displacement damage during irradiation on the Charpy impact behavior. Subsize Charpy impact specimens of 9Cr-1MoVNb (modified 9Cr-1Mo) and 12Cr-1MoVW (Sandvik HT9) steels and 12Cr-1MoVW with 2%Ni (12Cr-1MOVW-2Ni) were irradiated in the High Flux Isotope Reactor (HFIR) at 300 and 400{degree}C to damage levels up to 2.5 dpa. The objective was to study the effect of the simultaneous formation of displacement damage and transmutation helium on impact toghness. Despite the low fluence relative to previous irradiations of these steels, significant increases in the ductile-brittle transition temperature (DBTT) occurred. The 12Cr-1MoVW-2Ni steel irradiated at 400{degree}C had the largest increase in DBTT and displayed indications of intergranular fracture. A mechanism is proposed to explain how helium can affect the fracture behaviour of this latter steel in the present tests, and how it affected all three steels in previous experiments, where the steels were irradiated to higher fluences.

  9. Fractal characteristics of fracture morphology of steels irradiated with high-energy ions

    NASA Astrophysics Data System (ADS)

    Xian, Yongqiang; Liu, Juan; Zhang, Chonghong; Chen, Jiachao; Yang, Yitao; Zhang, Liqing; Song, Yin

    2015-06-01

    A fractal analysis of fracture surfaces of steels (a ferritic/martensitic steel and an oxide-dispersion-strengthened ferritic steel) before and after the irradiation with high-energy ions is presented. Fracture surfaces were acquired from a tensile test and a small-ball punch test (SP). Digital images of the fracture surfaces obtained from scanning electron microscopy (SEM) were used to calculate the fractal dimension (FD) by using the pixel covering method. Boundary of binary image and fractal dimension were determined with a MATLAB program. The results indicate that fractal dimension can be an effective parameter to describe the characteristics of fracture surfaces before and after irradiation. The rougher the fracture surface, the larger the fractal dimension. Correlation of the change of fractal dimension with the embrittlement of the irradiated steels is discussed.

  10. Hydrogen isotope transfer in austenitic steels and high-nickel alloy during in-core irradiation

    SciTech Connect

    Polosukhin, B.G.; Sulimov, E.M.; Zyrianov, A.P.; Kalinin, G.M.

    1995-10-01

    The transfer of protium and deuterium in austenitic chromium-nickel steels and in a high-nickel alloy was studied in a specially designed facility. The transfer parameters of protium and deuterium were found to change greatly during in-core irradiation, and the effects of irradiation increased as the temperature decreased. Thus, at temperature T<673K, the relative increase in the permeability of hydrogen isotopes under irradiation can be orders of magnitude higher in these steels. Other radiation effects were also observed, in addition to the changes from the initial values in the effects of protium and deuterium isotopic transfer. 4 refs., 3 figs., 2 tabs.

  11. Applicability of the fracture toughness master curve to irradiated highly embrittled steel and intergranular fracture

    SciTech Connect

    Nanstad, Randy K; Sokolov, Mikhail A; McCabe, Donald E

    2008-01-01

    The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory has evaluated a submerged-arc (SA) weld irradiated to a high level of embrittlement and a temper embrittled base metal that exhibits significant intergranular fracture (IGF) relative to representation by the Master Curve. The temper embrittled steel revealed that the intergranular mechanism significantly extended the transition temperature range up to 150 C above To. For the irradiated highly embrittled SA weld study, a total of 21 1T compact specimens were tested at five different temperatures and showed the Master Curve to be nonconservative relative to the results, although that observation is uncertain due to evidence of intergranular fracture.

  12. Migration and accumulation at dislocations of transmutation helium in austenitic steels upon neutron irradiation

    NASA Astrophysics Data System (ADS)

    Kozlov, A. V.; Portnykh, I. A.

    2016-04-01

    The model of the migration and accumulation at dislocations of transmutation helium and the formation of helium-vacancy pore nuclei in austenitic steels upon neutron irradiation has been proposed. As illustrations of its application, the dependences of the characteristics of pore nuclei on the temperature of neutron irradiation have been calculated. The results of the calculations have been compared with the experimental data in the literature on measuring the characteristics of radiation-induced porosity that arises upon the irradiation of shells of fuel elements of a 16Cr-19Ni-2Mo-2Mn-Si-Ti-Nb-V-B steel in a fast BN600 neutron reactor at different temperatures.

  13. Fracture properties of neutron-irradiated martensitic 9Cr-WVTa steels below room temperature

    NASA Astrophysics Data System (ADS)

    Abe, F.; Narui, M.; Kayano, H.

    1994-09-01

    Fracture properties of the reduced activation martensitic 9Cr-1WVTa and 9Cr-3WVTa steels were investigated by carrying out instrumented Charpy impact tests and tensile tests at temperatures below room temperature after irradiation in the Japan Materials Testing Reactor at 493 and 538 K. Modified 9Cr-1MoVNb steel was also examined for comparison. The irradiation-induced increase in ductile-to-brittle transition temperature was 53, 26 and 40 K for the {1}/{3} size Charpy specimens of 9Cr-1WVTa, 9Cr-3WVTa and 9Cr-1MoVNb steels, respectively, which resulted primarily from the irradiation-induced increase in yield stress. The cleavage fracture stress was 1820-1870 MPa for the three steels in unirradiated conditions, which was scarcely affected by irradiation. The deflections to the maximum load and to the brittle fracture initiation were decreased by irradiation. In the tensile test, quasi-cleavage fracture occurred at 77 K in both unirradiated and irradiated conditions. The cleavage fracture stress was 1320-1380 MPa for the tensile specimens of the three steels, which was about 1.4 times smaller than that for the Charpy specimens.

  14. Nondestructive Evaluation of Irradiation Embrittlement of SQV2A Steel by Using Magnetic Method

    SciTech Connect

    Shiwa, Mitsuharu; Cheng Weiying; Nakahigashi, Shigeo; Komura, Ichiro; Fujiwara, Koji; Takahashi, Norio

    2006-03-06

    Irradiation embrittlement of SQV2A steel was evaluated by magnetic methods. Thermal aging (TA) and electron irradiation (EI) specimens were prepared to evaluate the thermal aging and the irradiation damage effects separately. B-H loops changed with TA and EI. Higher harmonics of AC magnetization signals were sensitive to micro-structure changing of specimens. The intensity of the 3rd harmonics increased linearly with over 100 years of equivalent operation time by Larson-Miller parameter of nuclear power plants.

  15. Neutron irradiation effects on the ductile-brittle transition of ferritic/martensitic steels

    SciTech Connect

    Klueh, R.L.; Alexander, D.J.

    1997-08-01

    Ferritic/martensitic steels such as the conventional 9Cr-1MoVNb (Fe-9Cr-1Mo-0.25V-0.06Nb-0.1C) and 12Cr-1MoVW (Fe-12Cr-1Mo-0.25V-0.5W-0.5Ni-0.2C) steels have been considered potential structural materials for future fusion power plants. The major obstacle to their use is embrittlement caused by neutron irradiation. Observations on this irradiation embrittlement is reviewed. Below 425-450{degrees}C, neutron irradiation hardens the steels. Hardening reduces ductility, but the major effect is an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy, as measured by a Charpy impact test. After irradiation, DBTT values can increase to well above room temperature, thus increasing the chances of brittle rather than ductile fracture.

  16. Tensile properties of reduced activation Fe-9Cr-2W steels after FFTF irradiation

    NASA Astrophysics Data System (ADS)

    Kurishita, H.; Kayano, H.; Narui, M.; Kimura, A.; Hamilton, M. L.; Gelles, D. S.

    1994-09-01

    In order to develop radiation resistant steels with reduced activation for fusion reactor applications, the effect of fast neutron irradiation was investigated on the tensile properties of five types of Fe-9Cr-2W martensitic steel with and without small additions of boron, yttrium and aluminum. Miniature tensile specimens of the steels were irradiated to 28 dpa at 663 K and 33-35 dpa at 703, 793 and 873 K in the Fast Flux Test Facility (FFTF) and were deformed at temperatures between 300 and 873 K. The yield and ultimate tensile stresses were not significantly affected by the irradiations, but the total elongation was considerably decreased by the irradiation at 663 K. The reduction in elongation depended strongly on the test temperature with a maximum at around 673 K. The addition of yttrium alone tended to increase the high temperature strength, while the simultaneous addition of yttrium and aluminum tended to decrease the total elongation.

  17. Study on the mechanical properties evolution of A508-3 steel under proton irradiation

    NASA Astrophysics Data System (ADS)

    Lei, Jing; Ding, Hui; Shu, Guo-gang; Wan, Qiang-mao

    2014-11-01

    In an effort to study the effect of irradiation on the hardening behavior of reactor pressure vessel (RPV) steel, nanoindentation was employed to investigate the mechanical properties of A508-3 steel after an irradiation with 190 keV proton to the dose range of 0.054-0.271 displacement per atom (dpa) at room temperature. The results show that the relationship between the nanohardness and indent depth is in accordance with the Nix-Gao model. The nanohardness of A508-3 steel increases notably with the dose. In addition, the contribution of the irradiation-induced microstructural defects including matrix damage and nano clusters to the irradiation hardening is discussed.

  18. Deformation behavior in reactor pressure vessel steels as a clue to understanding irradiation hardening.

    SciTech Connect

    DiMelfi, R. J.; Alexander, D. E.; Rehn, L. E.

    1999-10-25

    In this paper, we examine the post-yield true stress vs true strain behavior of irradiated pressure vessel steels and iron-based alloys to reveal differences in strain-hardening behavior associated with different irradiating particles (neutrons and electrons) and different alloy chernky. It is important to understand the effects on mechanical properties caused by displacement producing radiation of nuclear reactor pressure steels. Critical embrittling effects, e.g. increases in the ductile-to-brittle-transition-temperature, are associated with irradiation-induced increases in yield strength. In addition, fatigue-life and loading-rate effects on fracture can be related to the post-irradiation strain-hardening behavior of the steels. All of these properties affect the expected service life of nuclear reactor pressure vessels. We address the characteristics of two general strengthening effects that we believe are relevant to the differing defect cluster characters produced by neutrons and electrons in four different alloys: two pressure vessel steels, A212B and A350, and two binary alloys, Fe-0.28 wt%Cu and Fe-0.74 wt%Ni. Our results show that there are differences in the post-irradiation mechanical behavior for the two kinds of irradiation and that the differences are related both to differences in damage produced and alloy chemistry. We find that while electron and neutron irradiations (at T {le} 60 C) of pressure vessel steels and binary iron-based model alloys produce similar increases in yield strength for the same dose level, they do not result in the same post-yield hardening behavior. For neutron irradiation, the true stress flow curves of the irradiated material can be made to superimpose on that of the unirradiated material, when the former are shifted appropriately along the strain axis. This behavior suggests that neutron irradiation hardening has the same effect as strain hardening for all of the materials analyzed. For electron irradiated steels, the

  19. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    SciTech Connect

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  20. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    SciTech Connect

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950`s are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  1. Irradiation creep of low-activation ferritic steels in FFTF/MOTA*1

    NASA Astrophysics Data System (ADS)

    Kohyama, A.; Kohno, Y.; Asakura, K.; Yoshino, M.; Namba, C.; Eiholzer, C. R.

    1994-09-01

    Irradiation creep behavior of low-activation steels, developed as structural materials for fusion reactors (JLF series steels), was investigated to obtain a fundamental understanding of these alloys under fast neutron irradiation in FFTF. (2.25-8)Cr(1-2)W bainitic steels and 12Cr-2W ferritic steels showed superior creep resistance to type-316 stainless steels under fast neutron irradiation up to 520°C. At temperatures below 460°C the creep strain increased with increasing Cr content up to 7 Cr, and further increments of Cr content up to 12% reduced the creep strain. At temperatures between 460 and 600°C, 7-8 Cr ferritic steels showed the largest creep strain. Swelling-enhanced creep, near the peak swelling temperature of 410°C, was also observed. The 9Cr-2W ferritic steel JLF-1 presented excellent properties, suggesting it as a leading candidate alloy for structural components of fusion reactors.

  2. Void swelling of Japanese candidate martensitic steels under FFTF/MOTA irradiation

    NASA Astrophysics Data System (ADS)

    Morimura, T.; Kimura, A.; Matsui, H.

    1996-12-01

    Microstructural observations of six Japanese candidate 7-9% Cr reduced activation martensitic steels were carried out after heavy neutron irradiation in order to investigate the void swelling behavior of each steel. Neutron irradiations were performed in the FFTF/MOTA up to 67 dpa at temperatures between 638 and 873 K. Transmission electron microscope observations revealed that voids were formed in all the steels irradiated to 67 dpa at 703 K, and the highest void swelling was observed in JLM-1 which was added with 30 wt.ppm of boron (0.74%), and the minimum void swelling was observed in F82H steel (0.12%). The 9% Cr martensitic steels showed the peak of void swelling at temperatures around 700 K, where void swelling gradually increased with increasing irradiation fluence to 30 dpa and increased rapidly above it. It is considered that the incubation period of void swelling of 9% Cr martensitic steels (JLM series) is about 30 dpa. JLM-1 showed the highest void swelling rate (0.045%/dpa at most). The addition of 30 wt.ppm of boron enhanced void swelling, while it was suppressed by the addition of 100 wt.ppm Ti in the 9% Cr martensitic steel. The JLF-3 steel (7.03% Cr) and F82H (7.65% Cr) showed less void swelling than JLF-I (9.04% Cr). The alloying effects on the swelling behavior of the steels were interpreted in terms of the difference in the precipitation morphology of carbides.

  3. Irradiation effects on base metal and welds of 9Cr-1Mo (EM10) martensitic steel

    SciTech Connect

    Alamo, A.; Seran, J.L.; Rabouille, O.; Brachet, J.C.; Maillard, A.; Touron, H.; Royer, J.

    1996-12-31

    9Cr martensitic steels are being developed for core components (wrapper tubes) of fast breeder reactors as well as for fusion reactor structures. Here, the effects of fast neutron irradiation on the mechanical behavior of base metal and welds of 9Cr-1Mo (EM10) martensitic steel have been studied. Two types of weldments have been produced by TIG and electron beam techniques. Half of samples have been post-weld heat treated to produce a stress-relieved structure. The irradiation has been conducted in the Phenix reactor to doses of 63--65 dpa in the temperature range 450--459 C. The characterization of the welds, before and after irradiation, includes metallographic observations, hardness measurements, tensile and Charpy tests. It is shown that the mechanical properties of the welds after irradiation are in general similar to the characteristics obtained on the base metal, which is little affected by neutron irradiation.

  4. Tensile properties and damage microstructures in ORR/HFIR-irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Wakai, E.; Hashimoto, N.; Robertson, J. P.; Jistukawa, S.; Sawai, T.; Hishinuma, A.

    2000-12-01

    The synergistic effect of displacement damage and helium generation under neutron irradiation on tensile behavior and microstructures of austenitic stainless steels was investigated. The steels were irradiated at 400°C in the spectrally-tailored (ST) Oak Ridge research reactor/high flux isotope reactor (ORR/HFIR) capsule to 17 dpa with a helium production of about 200 appm and in the HFIR target capsule to 21 and 34 dpa with 1590 and 2500 appm He, respectively. The increase of yield strength in the target irradiation was larger than that in the ST irradiation because of the high-number density of Frank loops, bubbles, voids, and carbides. Based on the theory of dispersed barrier hardening, the strengths evaluated from these clusters coincide with the measured increase of yield strengths. This analysis suggests that the main factors of radiation hardening in the ST and the target irradiation at 400°C are Frank-type loops and cavities, respectively.

  5. Microstructure and fracture behavior of F82H steel under different irradiation and tensile test conditions

    NASA Astrophysics Data System (ADS)

    Wang, K.; Dai, Y.; Spätig, P.

    2016-01-01

    Specimens of martensitic steel F82H were irradiated to doses ranging from 10.7 dpa/850 appm He to 19.6 dpa/1740 appm He at temperatures between 165 and 305 °C in the second experiment of SINQ Target Irradiation Program (STIP-II). Tensile tests were conducted at different temperatures and various fracture modes were observed. Microstructural changes including irradiation-induced defect clusters, dislocation loops and helium bubbles under different irradiation conditions were investigated using transmission electron microscopy (TEM). The deformation microstructures of tensile tested specimens were carefully examined to understand the underlying deformation mechanisms. Deformation twinning was for the first time observed in irradiated martensitic steels. A change of deformation mechanism from dislocation channeling to deformation twinning was observed when the fracture mode changed from rather ductile (quasi-cleavage) to brittle (intergranular or cleavage and intergranular mixed).

  6. Heavy-section steel irradiation program. Progress report, October 1994--March 1995

    SciTech Connect

    Corwin, W.R.

    1995-10-01

    This document is the October 1994-March 1995 Progress Report for the Heavy Section Steel Irradiation Program. The report contains a summary of activities in each of the 14 tasks of the HSSI Program, including: (1) Program management, (2) Fracture toughness shifts in high-copper weldments, (3) Fracture toughness shifts in low upper-shelf welds, (4) Irradiation effects in a commercial low upper-shelf weld, (5) Irradiation effects on weld heat-affected zone and plate materials, (6) Annealing effects in low upper-shelf welds, (7) Microstructural analysis of radiation effects, (8) In-service irradiated and aged material evaluations, (9) Japanese power development reactor vessel steel examination, (10) fracture toughness curve shift method, (11) Special technical assistance, (12) Technical assistance for JCCCNRS, (13) Correlation monitor materials, and (14) Test reactor irradiation coordination. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.

  7. Effect of cold work on tensile behavior of irradiated type 316 stainless steel

    SciTech Connect

    Klueh, R.L.; Maziasz, P.J.

    1986-01-01

    Tensile specimens were irradiated in ORR at 250, 290, 450, and 500/sup 0/C to produce a displacement damage of approx.5 dpa and 40 at. ppM He. Irradiation at 250 and 290/sup 0/C caused an increase in yield stress and ultimate tensile strength and a decrease in ductility relative to unaged and thermally aged controls. The changes were greatest for the 20%-cold-worked steel and lowest for the 50%-cold-worked steel. Irradiation at 450/sup 0/C caused a slight relative decrease in strength for all cold-worked conditions. A large decrease was observed at 500/sup 0/C, with the largest decrease occurring for the 50%-cold-worked specimen. No bubble, void, or precipitate formation was observed for specimens examined by transmission electron microscopy (TEM). The irradiation hardening was correlated with Frank-loop and ''black-dot'' loop damage. A strength decrease at 500/sup 0/C was correlated with dislocation network recovery. Comparison of tensile and TEM results from ORR-irradiated steel with those from steels irradiated in the High Flux Isotope Reactor and the Experimental Breeder Reactor indicated consistent strength and microstructure changes.

  8. Effects of helium content of microstructural development in Type 316 stainless steel under neutron irradiation

    SciTech Connect

    Maziasz, P.J.

    1985-11-01

    This work investigated the sensitivity of microstructural evolution, particularly precipitate development, to increased helium content during thermal aging and during neutron irradiation. Helium (110 at. ppM) was cold preinjected into solution annealed (SA) DO-heat type 316 stainess steel (316) via cyclotron irradiation. These specimens were then exposed side by side with uninjected samples. Continuous helium generation was increased considerably relative to EBR-II irradiation by irradiation in HFIR. Data were obtained from quantitative analytical electron microscopy (AEM) in thin foils and on extraction replicas. 480 refs., 86 figs., 19 tabs.

  9. Nanostructure evolution in ODS Eurofer steel under irradiation up to 32 dpa

    NASA Astrophysics Data System (ADS)

    Rogozhkin, S. V.; Orlov, N. N.; Aleev, A. A.; Zaluzhnyi, A. G.; Kozodaev, M. A.; Kuibeda, R. P.; Kulevoy, T. V.; Nikitin, A. A.; Chalykh, B. B.; Lindau, R.; Möslang, A.; Vladimirov, P.

    2015-01-01

    The nanostructure of the ODS Eurofer steel (9% CrWVTa + 0.5% Y2O3) has been studied after irradiation by iron ions to a damaging dose of 32 dpa. This steel in the initial state is characterized by the presence of a significant amount (˜1024 m-3) of nanosized (2-4 nm) clusters containing atoms of V, Y, O, and N. An analysis of the distribution of various chemical elements in the tested volumes has revealed variations in the composition of the matrix and of the nanosized clusters during irradiation. The data obtained were compared with the results for the ODS Eurofer steel subjected to reactor irradiation to a dose of 32 dpa.

  10. TEM characterization of dislocation loops in irradiated bcc Fe-based steels

    SciTech Connect

    Yao, Bo; Edwards, Danny J.; Kurtz, Richard J.

    2012-12-08

    In this study, we describe a methodology to examine dislocation loops in irradiated steels based on a combination of crystallographic information and g*b invisibility criteria. Dislocation loops in transmission electron microscope (TEM) images can be conveniently analyzed using this method. Through this analysis approach, dislocation loops in reduced activation ferritic/martensitic (RAFM) steels irradiated at 400 *C have been examined. The predominant types of loops found in irradiated RAFM steels were h100i{200} and 1/2h111i 111. The size, density, and density anisotropy of these two types of dislocation loops were quantified. It was observed that the h100i{200} loop density is more than twice that of 1/2h111i{111} loops. A large density anisotropy of h100i{200} loops was identified.

  11. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    SciTech Connect

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO{sub 3} and H{sub 2}O{sub 2} solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area).

  12. Void denuded zone formation for Fe-15Cr-15Ni steel and PNC316 stainless steel under neutron and electron irradiations

    NASA Astrophysics Data System (ADS)

    Sekio, Yoshihiro; Yamashita, Shinichiro; Sakaguchi, Norihito; Takahashi, Heishichiro

    2015-03-01

    Irradiation-induced void denuded zone (VDZ) formation near grain boundaries was studied to clarify the effects of minor alloying elements on vacancy diffusivity during irradiation in practical PNC316 stainless steel developed for nuclear reactor core materials. The test materials were Fe-15Cr-15Ni steel without additives and PNC316 stainless steel; the latter contains minor alloying elements to improve the void swelling resistance. These steels were neutron-irradiated in the experimental fast reactor JOYO at temperatures from 749 K to 775 K and fast neutron doses of 18-103 dpa, and electron irradiation was also carried out using 1 MeV high voltage electron microscopy at temperatures of 723 K and 773 K and doses up to 14.4 dpa. VDZ formation was analyzed by TEM microstructural observation after irradiation by considering radiation-induced segregation near the grain boundaries. VDZs were formed near random grain boundaries with higher misfit angles in both Fe-15Cr-15Ni and PNC316 steels. The VDZ widths in the PNC316 stainless steel were narrower than those for the Fe-15Cr-15Ni steel for all neutron and electron irradiations. The VDZ width analysis implied that the vacancy diffusivity was reduced in PNC316 stainless steel as a result of interaction of vacancies with minor alloying elements.

  13. Heavy-Section Steel Irradiation Program. Volume 5, No. 2, Progress report, April 1994--September 1994.

    SciTech Connect

    Corwin, W.R.

    1995-07-01

    The Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior and the fracture toughness properties of typical pressure-vessel steels as they relate to light-water RPV integrity. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into 14 tasks: (1) program management, (2) fracture toughness curve shift in high-copper weldments (Series 5 and 6), (3) K{sub lc} and K{sub la} curve shifts in low upper-shelf (LUS) welds (Series 8), (4) irradiation effects in a commercial LUS weld (Series 10), (5) irradiation effects on weld heat-affected zone and plate materials (Series 11), (6) annealing effects in LUS welds (Series 9), (7) microstructural and microfracture analysis of irradiation effects, (8) in-service irradiated and aged material evaluations, (9) Japan Power Development Reactor (JPDR) steel examination, (10) fracture toughness curve shift method, (11) special technical assistance, (12) technical assistance for Joint Coordinating Committee on Civilian Nuclear Reactor Safety (JCCCNRS) Working Groups 3 and 12, (13) correlation monitor materials, and (14) test reactor coordination. Progress on each task is reported.

  14. Effects of residual stress on irradiation hardening in stainless steels

    NASA Astrophysics Data System (ADS)

    Okubo, N.; Miwa, Y.; Kondo, K.; Kaji, Y.

    2009-04-01

    Effects of residual stress on irradiation hardening were studied in advance for predicting irradiation assisted stress corrosion cracking. The specimens of SUS316 and SUS316L with several % plastic strains, which correspond to weld residual stress, were prepared by bending and keeping deformation under irradiation. Ion irradiations of 12 MeV Ni 3+ were performed at 330, 400 and 550 oC to 45 dpa. No bended specimen was simultaneously irradiated with the bended specimen. The residual stress was estimated by X-ray residual stress measurements before and after the irradiation. The micro-hardness was measured by using nanoindenter. The residual stress did not relax even for the case of the higher temperature aging at 500 oC for the same time of irradiation. The residual stress after ion irradiation up to high dpa, however, relaxed at these experimental temperatures. The irradiation hardening of stressed specimen was obviously lower than that of un-stressed one in case of SUS316L irradiated at 300 oC to 12 dpa.

  15. Irradiation-induced microchemical changes in highly irradiated 316 stainless steel

    NASA Astrophysics Data System (ADS)

    Fujii, K.; Fukuya, K.

    2016-02-01

    Cold-worked 316 stainless steel specimens irradiated to 74 dpa in a pressurized water reactor (PWR) were analyzed by atom probe tomography (APT) to extend knowledge of solute clusters and segregation at higher doses. The analyses confirmed that those clusters mainly enriched in Ni-Si or Ni-Si-Mn were formed at high number density. The clusters were divided into three types based on their size and Mn content; small Ni-Si clusters (3-4 nm in diameter), and large Ni-Si and Ni-Si-Mn clusters (8-10 nm in diameter). The total cluster number density was 7.7 × 1023 m-3. The fraction of large clusters was almost 1/10 of the total density. The average composition (in at%) for small clusters was: Fe, 54; Cr, 12; Mn, 1; Ni, 22; Si, 11; Mo, 1, and for large clusters it was: Fe, 44; Cr, 9; Mn, 2; Ni, 29; Si, 14; Mo,1. It was likely that some of the Ni-Si clusters correspond to γ‧ phase precipitates while the Ni-Si-Mn clusters were precursors of G phase precipitates. The APT analyses at grain boundaries confirmed enrichment of Ni, Si, P and Cu and depletion of Fe, Cr, Mo and Mn. The segregation behavior was consistent with previous knowledge of radiation induced segregation.

  16. Charpy impact tests on martensitic/ferritic steels after irradiation in SINQ target-3

    NASA Astrophysics Data System (ADS)

    Dai, Yong; Marmy, Pierre

    2005-08-01

    Charpy impact tests were performed on martensitic/ferritic (MF) steels T91, F82H, Optifer-V and Optimax-A/-C irradiated in SINQ Target-3 up to 7.5 dpa and 500 appm He in a temperature range of 120-195 °C. Results demonstrate that for all the four kinds of steels, the ductile-to-brittle transition temperature (DBTT) increases with irradiation dose. The difference in the DBTT shifts (ΔDBTT) of the different steels is not significant after irradiation in the SINQ target. The ΔDBTT data from the previous small punch (Δ DBTT SP) and the present Charpy impact (ΔDBTT CVN) tests can be correlated with the expression: Δ DBTT SP = 0.4ΔDBTT CVN. All the ΔDBTT data fall into a linear band when they are plotted versus helium concentration. The results indicate that helium effects on the embrittlement of MF steels are significant, particularly at higher concentrations. It suggests that MF steels may not be very suitable for applications at low temperatures in spallation irradiation environments where helium production is high.

  17. Impact behavior of reduced-activation steels irradiated to 24 dpa

    SciTech Connect

    Klueh, R.L.; Alexander, D.J.

    1996-04-01

    Charpy impact properties of eight reduced-activation Cr-W ferritic steels were determined after irradiation to {approx}21-24 dpa in the Fast Flux Test Facility (FFTF) at 365{degree}C. Chromium concentrations in the eight steels ranged from 2.25 to 12wt% Cr (steels contained {approx}0.1%C). the 2 1/4Cr steels contained variations of tungsten and vanadium, and the steels with 5, 9, and 12% Cr, contained a combination of 2% W and 0.25% V. A 9Cr in FFTF to {approx}6-8 and {approx}15-17 dpa. Irradiation caused an increase in the DBTT and decrease in the USE, but there was little further change in the DBTT from that observed after the 15-17 dpa irradiation, indicating that the shift had essentially saturated with fluence. The results are encouraging because they indicate that the effect of irradiation on toughness can be faorably affected by changing composition and microstructure.

  18. Positron annihilation Doppler broadening spectroscopy study on Fe-ion irradiated NHS steel

    NASA Astrophysics Data System (ADS)

    Zhu, Huiping; Wang, Zhiguang; Gao, Xing; Cui, Minghuan; Li, Bingsheng; Sun, Jianrong; Yao, Cunfeng; Wei, Kongfang; Shen, Tielong; Pang, Lilong; Zhu, Yabin; Li, Yuanfei; Wang, Ji; Song, Peng; Zhang, Peng; Cao, Xingzhong

    2015-02-01

    In order to study the evolution of irradiation-induced vacancy-type defects at different irradiation fluences and temperatures, a new type of ferritic/martensitic (F/M) steel named NHS (Novel High Silicon) was irradiated by 3.25 MeV Fe-ion at room temperature and 723 K to fluences of 4.3 × 1015 and 1.7 × 1016 ions/cm2. After irradiation, vacancy-type defects were investigated with variable-energy positron beam Doppler broadening spectra. Energetic Fe-ions produced a large number of vacancy-type defects in the NHS steel, but one single main type of vacancy-type defect was observed in both unirradiated and irradiated samples. The concentration of vacancy-type defects decreased with increasing temperature. With the increase of irradiation fluence, the concentration of vacancy-type defects increased in the sample irradiated at RT, whereas for the sample irradiated at 723 K, it decreased. The enhanced recombination between vacancies and excess interstitial Fe atoms from deeper layers, and high diffusion rate of self-interstitial atoms further improved by diffusion via grain boundary and dislocations at high temperature, are thought to be the main reasons for the reversed trend of vacancy-type defects between the samples irradiated at RT and 723 K.

  19. Dependence of impact properties on irradiation temperature in reduced-activation martensitic steels

    NASA Astrophysics Data System (ADS)

    Kimura, Akihiko; Narui, Minoru; Misawa, Toshihei; Matsui, Hideki; Kohyama, Akira

    1998-10-01

    Ductile-brittle transition (DBT) behavior of 9%Cr-2%W reduced-activation martensitic (RAM) steels has been investigated following neutron irradiation in the fast flux test facility, materials open test facility (FFTF/MOTA) at different temperatures. Both the irradiations at 663 and 733 K cause an increase in DBT temperature, while the irradiation at 663 K induces the hardening and the softening at 733 K. Microstructural observation by transmission electron microscope (TEM) revealed that small dislocation loops existed in the specimen irradiated at 663 K and no such a loop, but relatively large M 6C carbides and Laves phase were formed by the irradiation at 733 K. There appears to be a linear dependence between ΔDBTT and Δ σY in neutron irradiated RAM steels when irradiation induces the hardening. Irradiation embrittlement accompanied by the softening is considered to be due to reduction of cleavage fracture stress caused by the irradiation-induced recovery of the martensitic structure, namely decrease in dislocation density and formation of large precipitates.

  20. Shear Punch Properties of Low Activation Ferritic Steels Following Irradiation in ORR

    SciTech Connect

    Ermi, Ruby M.; Hamilton, Margaret L.; Gelles, David S.; Ermi, August M.

    2001-10-01

    Shear punch post-irradiation test results are reported for a series of low activation steels containing Mn following irradiation in the Oak Ridge Reactor at 330 and 400 degrees centigrade to {approx}10 dpa. Alloy compositions included 2Cr, 9Cr and 12Cr steels with V to 1.5% and W to 1.0%. Comparison of results with tensile test results showed good correlations with previously observed trends except where disks were improperly manufactured because they were too thin or because engraving was faulty.

  1. Evaluation of stress corrosion cracking of irradiated 304L stainless steel in PWR environment using heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Gupta, J.; Hure, J.; Tanguy, B.; Laffont, L.; Lafont, M.-C.; Andrieu, E.

    2016-08-01

    IASCC has been a major concern regarding the structural and functional integrity of core internals of PWR's, especially baffle-to-former bolts. Despite numerous studies over the past few decades, additional evaluation of the parameters influencing IASCC is still needed for an accurate understanding and modeling of this phenomenon. In this study, Fe irradiation at 450 °C was used to study the cracking susceptibility of 304 L austenitic stainless steel. After 10 MeV Fe irradiation to 5 dpa, irradiation-induced damage in the microstructure was characterized and quantified along with nano-hardness measurements. After 4% plastic strain in a PWR environment, quantitative information on the degree of strain localization, as determined by slip-line spacing, was obtained using SEM. Fe-irradiated material strained to 4% in a PWR environment exhibited crack initiation sites that were similar to those that occur in neutron- and proton-irradiated materials, which suggests that Fe irradiation may be a representative means for studying IASCC susceptibility. Fe-irradiated material subjected to 4% plastic strain in an inert argon environment did not exhibit any cracking, which suggests that localized deformation is not in itself sufficient for initiating cracking for the irradiation conditions used in this study.

  2. He and H irradiation effects on the nanoindentation hardness of CLAM steel

    NASA Astrophysics Data System (ADS)

    Jiang, Siben; Peng, Lei; Ge, Hongen; Huang, Qunying; Xin, Jingping; Zhao, Ziqiang

    2014-12-01

    In this study, He and H ion irradiation induced hardening behavior of China Low Activation Martensitic (CLAM) steel was investigated, and the influence of Si on irradiation hardening was also examined. CLAM steel with different Si contents, Heat 0912 and Heat 0408D, were irradiated with single He (He concentration range from 0 to 2150 appm) ion beam and He/H dual ion beams. Then nanoindentation tests were applied to evaluate the ion irradiation induced hardening effect. The result of Heat 0912 showed hardening effect would be more serious with higher He concentration, and the trend saturated when He concentration reach 1000 appm. Comparing the result of Heat 0912 and Heat 0408D, higher Si content might improve the resistance to hardening.

  3. Effect of boron on post irradiation tensile properties of reduced activation ferritic steel (F-82H) irradiated in HFIR

    SciTech Connect

    Shiba, Kiyoyuki; Suzuki, Masahide; Hishinuma, Akimichi; Pawel, J.E.

    1994-12-31

    Reduced activation ferritic/martensitic steel, F-82H (Fe-8Cr-2W-V-Ta), was irradiated in the High Flux Isotope Reactor (HFIR) to doses between 11 and 34 dpa at 400 and 500 C. Post irradiation tensile tests were performed at the nominal irradiation temperature in vacuum. Some specimens included {sup 10}B or natural boron (nB) to estimate the helium effect on tensile properties. Tensile properties including the 0.2% offset yield stress, the ultimate tensile strength, the uniform elongation and the total elongation were measured. The tensile properties were not dependent on helium content in specimens irradiated to 34 dpa, however {sup 10}B-doped specimens with the highest levels of helium showed slightly higher yield strength and less ductility than boron-free specimens. Strength appears to go through a peak, and ductility through a trough at about 11 dpa. The irradiation to more than 21 dpa reduced the strength and increased the elongation to the unirradiated levels. Ferritic steels are one of the candidate alloys for nuclear fusion reactors because of their good thermophysical properties, their superior swelling resistance, and the low corrosion rate in contact with potential breeder and coolant materials.

  4. TEM, XRD and nanoindentation characterization of Xenon ion irradiation damage in austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Huang, H. F.; Li, J. J.; Li, D. H.; Liu, R. D.; Lei, G. H.; Huang, Q.; Yan, L.

    2014-11-01

    Cross-sectional and bulk specimens of a 20% cold-worked 316 austenitic stainless steel (CW 316 SS) has been characterized by TEM, XRD and nanoindentation to determine the microstructural evolution and mechanical property changes of 316 SS after irradiation with 7 MeV Xe26+ ions. TEM results reveal the presence of dislocation loops with a number density of approximately 3 × 1022 m-3 and sizes between 3 to 10 nm due to the collapse of vacancy rich cores inside displacement cascades. Peak broadening observed in XRD diffraction patters reveal systematic changes to lattice parameters due to irradiation. The calculated indentation values in irradiated 316 SS were found to be much higher in comparison to the unirradiated specimen, indicating the dose dependent effect of irradiation on hardness. The relationship between irradiation induced microstructural evolution and the changes to the mechanical properties of CW 316 SS are discussed in the context of fluence and irradiation temperature.

  5. The microstructure of neutron irradiated type-348 stainless steel and its relation to creep and hardening

    NASA Astrophysics Data System (ADS)

    Thomas, L. E.; Beeston, J. M.

    1982-06-01

    Annealed type-348 stainless steel specimens irradiated to 33 to 39 dpa at 350°C were examined by transmission electron microscopy to determine the cause of pronounced irradiation creep and hardening. The irradiation produced very high densities of 1-2 nm diameter helium bubbles, 2-20 nm diameter faulted (Frank) dislocation loops and 10 nm diameter precipitate particles. These defects account for the observed irradiation hardening but do not explain the creep strains. Too few point defects survive as faulted dislocation loops for significant creep by the stress-induced preferential absorption (SIPA) mechanism and there are not enough unfaulted dislocations for creep by climb-induced glide. Also, the irradiation-induced precipitates are face-centred cubic G-phase (a niobium nickel suicide), and cannot cause creep. It is suggested that the irradiation creep occurs by a grain-boundary movement mechanism such as diffusion accomodated grain-boundary sliding.

  6. Fatigue behavior of type 316 stainless steel following neutron irradiation inducing helium

    SciTech Connect

    Grossbeck, M.L.; Liu, K.C.

    1980-01-01

    Since a tokamak fusion reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially the first wall and blanket. Type 316 stainless steel in the 20% cold-worked condition has been irradiated in the HFIR in order to introduce helium as well as displacement damage. A miniature hourglass specimen was developed for the reactor irradiations and subsequent fully reversed low cycle fatigue testing. For material irradiated and tested at 430/sup 0/C in vacuum to a damage level of 7 to 15 dpa and containing 200 to 1000 appm He, a reduction in life by a factor of 3 to 10 was observed. An attempt was made to predict irradiated fatigue life by fitting data from irradiated material to a power law equation similar to the universal slopes equation and using ductility ratios from tensile tests to modify the equation for irradiated material.

  7. Temperature dependence of the deformation behavior of 316 stainless steel after low temperature neutron irradiation

    SciTech Connect

    Pawel-Robertson, J.E.; Rowcliffe, A.F.; Grossbeck, M.L.

    1996-10-01

    The effects of low temperature neutron irradiation on the tensile behavior of 316 stainless steel have been investigated. A single heat of solution annealed 316 was irradiated to 7 and 18 dpa at 60, 200, 330, and 400{degrees}C. The tensile properties as a function of dose and as a function of temperature were examined. Large changes in yield strength, deformation mode, strain to necking, and strain hardening capacity were seen in this irradiation experiment. The magnitudes of the changes are dependent on both irradiation temperature and neutron dose. Irradiation can more than triple the yield strength over the unirradiated value and decrease the strain to necking (STN) to less than 0.5% under certain conditions. A maximum increase in yield strength and a minimum in the STN occur after irradiation at 330{degrees}C but the failure mode remains ductile.

  8. Evolution of microstructure after irradiation creep in several austenitic steels irradiated up to 120 dpa at 320 °C

    NASA Astrophysics Data System (ADS)

    Renault-Laborne, A.; Garnier, J.; Malaplate, J.; Gavoille, P.; Sefta, F.; Tanguy, B.

    2016-07-01

    Irradiation creep was investigated in different austenitic steels. Pressurized tubes with stresses of 127-220 MPa were irradiated in BOR-60 at 320 °C to 120 dpa. Creep behavior was dependent on both chemical composition and metallurgical state of steels. Different steels irradiated with and without stress were examined by TEM. Without stress, the irradiation produced high densities of dislocation lines and Frank loops and, depending on the type of steels, precipitates. Stress induced an increase of the precipitate mean size and density and, for some grades, an increase of the mean loop size and a decrease of their density. An anisotropy of Frank loop density or size induced by stress was not observed systematically. Dislocation line microstructure seems not to be different between the stressed and unstressed specimens. No cavities were detectable in these specimens. By comparing with the data from this work, the main irradiation creep models are discussed.

  9. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    DOE PAGESBeta

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; Pakarinen, Janne; Wu, Yaqiao; Allen, Todd; Yang, Yong

    2015-08-08

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019more » n/cm2 E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinod,al decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. Lastly, The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.« less

  10. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    SciTech Connect

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; Pakarinen, Janne; Wu, Yaqiao; Allen, Todd; Yang, Yong

    2015-08-08

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019 n/cm2 E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinod,al decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. Lastly, The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

  11. Comparison of irradiated and hydrogen implanted German RPV steels using PAS technique

    NASA Astrophysics Data System (ADS)

    Pecko, Stanislav; Sojak, Stanislav; Slugeň, Vladimír

    2015-12-01

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This spectroscopic method is a really effective tool for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to irradiation. German commercial reactor pressure vessel steels, originally from CARISMA program, were used in our study. The German experimental reactor VAK was selected as the proper irradiation facility in the 1980s. A specimen in as-received state and 2 different irradiated cuts from the same material were measured by PALS and size of defects with their intensity was indentified. Afterwards there was prepared an experiment with concern in simulation of neutron irradiation by hydrogen ion implantation on a linear accelerator with energy of 100 keV. Results are concerning on comparison between defects caused by neutron irradiation and hydrogen implantation. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to hydrogen ions implantation.

  12. Damage structure of austenitic stainless steel 316LN irradiated at low temperature in HFIR

    SciTech Connect

    Hashimoto, N.; Robertson, J.P.; Grossbeck, M.L.; Rowcliffe, A.F.; Wakai, E.

    1998-03-01

    TEM disk specimens of austenitic stainless steel 316LN irradiated to damage levels of about 3 dpa at irradiation temperatures of either about 90 C or 250 C have been investigated by using transmission electron microscopy. The irradiation at 90 C and 250 C induced a dislocation loop density of 3.5 {times} 10{sup 22} m{sup {minus}3} and 6.5 {times} 10{sup 22} m{sup {minus}3}, a black dot density of 2.2 {times} 10{sup 23} m{sup {minus}3} and 1.6 {times} 10{sup 23} m{sup {minus}3}, respectively, in the steels, and a high density (<1 {times} 10{sup 22} m{sup {minus}3}) of precipitates in matrix. Cavities could be observed in the specimens after the irradiation. It is suggested that the dislocation loops, the black dots, and the precipitates cause irradiation hardening, an increase in the yield strength and a decrease in the uniform elongation, in the 316LN steel irradiated at low temperature.

  13. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    NASA Astrophysics Data System (ADS)

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; Pakarinen, Janne; Wu, Yaqiao; Allen, Todd; Yang, Yong

    2015-11-01

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ∼315 °C to 0.08 dpa (5.6 × 1019 n/cm2, E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinodal decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

  14. Microstructural evolution of RPV steels under proton and ion irradiation studied by positron annihilation spectroscopy

    NASA Astrophysics Data System (ADS)

    Jiang, J.; Wu, Y. C.; Liu, X. B.; Wang, R. S.; Nagai, Y.; Inoue, K.; Shimizu, Y.; Toyama, T.

    2015-03-01

    The microstructural evolution of reactor pressure vessel (RPV) steels induced by proton and heavy ion irradiation at low temperature (∼373 K) has been investigated using positron annihilation spectroscopy (PAS), atom probe tomography (APT), transmission electron microscopy (TEM) and nanoindentation. The PAS results indicated that both proton and heavy ion irradiation produce a large number of matrix defects, which contain small-size defects such as vacancies, vacancy-solute complexes, dislocation loops, and large-size vacancy clusters. In proton irradiated RPV steels, the size and number density of vacancy cluster defects increased rapidly with increasing dose due to the migration and agglomeration of vacancies. In contrast, for Fe ion irradiated steels, high density, larger size vacancy clusters can be easily induced at low dose, showing saturation in PAS response with increasing dose. No clear precipitates, solute-enriched clusters or other forms of solute segregation were observed by APT. Furthermore, dislocation loops were observed by TEM after 1.0 dpa, 240 keV proton irradiation, and an increase of the average nanoindentation hardness was found. It is suggested that ion irradiation produces many point defects and vacancy cluster defects, which induce the formation of dislocation loops and the increase of nanoindentation hardness.

  15. Crack initiation behavior of neutron irradiated model and commercial stainless steels in high temperature water

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale J.; Was, Gary S.

    2014-01-01

    The objective of this study was to isolate key factors affecting the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of eleven neutron-irradiated austenitic stainless steel alloys. Four commercial purity and seven high purity stainless steels were fabricated with specific changes in composition and microstructure, and irradiated in a fast reactor spectrum at 320 °C to doses between 4.4 and 47.5 dpa. Constant extension rate tensile (CERT) tests were performed in normal water chemistry (NWC), hydrogen water chemistry (HWC), or primary water (PW) environments to isolate the effects of environment, elemental solute addition, alloy purity, alloy heat, alloy type, cold work, and irradiation dose. The irradiated alloys showed a wide variation in IASCC susceptibility, as measured by the relative changes in mechanical properties and crack morphology. Cracking susceptibility measured by %IG was enhanced in oxidizing environments, although testing in the lowest potential environment caused an increase in surface crack density. Alloys containing solute addition of Ni or Ni + Cr exhibited no IASCC. Susceptibility was reduced in materials cold worked prior to irradiation, and increased with increasing irradiation dose. Irradiation-induced hardening was accounted for by the dislocation loop microstructure, however no relation between crack initiation and radiation hardening was found.

  16. Microstructural development under irradiation in European ODS ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Schäublin, R.; Ramar, A.; Baluc, N.; de Castro, V.; Monge, M. A.; Leguey, T.; Schmid, N.; Bonjour, C.

    2006-06-01

    Oxide dispersion strengthened steels based on the ferritic/martensitic steel EUROFER97 are promising candidates for a fusion reactor because of their improved high temperature mechanical properties and their potential higher radiation resistance relative to the base material. Several EUROFER97 based ODS F/M steels are investigated in this study. There are the Plansee ODS steels containing 0.3 wt% yttria, and the CRPP ODS steels, whose production route is described in detail. The reinforcing particles represent 0.3-0.5% weight and are composed of yttria. The effect of 0.3 wt% Ti addition is studied. ODS steel samples have been irradiated with 590 MeV protons to 0.3 and 1.0 dpa at room temperature and 350 °C. Microstructure is investigated by transmission electron microscopy and mechanical properties are assessed by tensile and Charpy tests. While the Plansee ODS presents a ferritic structure, the CRPP ODS material presents a tempered martensitic microstructure and a uniform distribution of the yttria particles. Both materials provide a yield stress higher than the base material, but with reduced elongation and brittle behaviour. Ti additions improve elongation at high temperatures. After irradiation, mechanical properties of the material are only slightly altered with an increase in the yield strength, but without significant decrease in the total elongation, relative to the base material. Samples irradiated at room temperature present radiation induced defects in the form of blacks dots with a size range from 2 to 3 nm, while after irradiation at 350 °C irradiation induced a0<1 0 0>{1 0 0} dislocation loops are clearly visible along with nanocavities. The dispersed yttria particles with an average size of 6-8 nm are found to be stable for all irradiation conditions. The density of the defects and the dispersoid are measured and found to be about 2.3 × 10 22 m -3 and 6.2 × 10 22 m -3, respectively. The weak impact of irradiation on mechanical properties of ODS F

  17. Temperature dependence of fracture toughness in HT9 steel neutron-irradiated up to 145 dpa

    SciTech Connect

    Baek, Jong-Hyuk; Byun, Thak Sang; Maloy, S; Toloczko, M

    2014-01-01

    The temperature dependence of fracture toughness in HT9 steel irradiated to high doses was investigated using miniature three-point bend (TPB) fracture specimens. These specimens were from the ACO-3 fuel duct wall of the Fast Flux Test Facility (FFTF), in which irradiation doses were in the range of 3.2 144.8 dpa and irradiation temperatures in the range of 380.4 502.6 oC. A miniature specimen reuse technique has been established for this investigation: the specimens used were the tested halves of miniature Charpy impact specimens (~13 3 4 mm) with diamond-saw cut in the middle. The fatigue precracking for specimens and fracture resistance (J-R) tests were carried out in a MTS servo-hydraulic testing machine with a vacuum furnace following the standard procedure described in the ASTM Standard E 1820-09. For each of five irradiated and one archive conditions, 7 to 9 J-R tests were performed at selected temperatures ranging from 22 C to 600 C. The fracture toughness of the irradiated HT9 steel was strongly dependent on irradiation temperatures rather than irradiation dose. When the irradiation temperature was below about 430 C, the fracture toughness of irradiated HT9 increased with test temperature, reached an upper shelf of 180 200 MPa m at 350 450 C and then decreased with test temperature. When the irradiation temperature 430 C, the fracture toughness was nearly unchanged until about 450 C and decreased with test temperature in higher temperature range. Similar test temperature dependence was observed for the archive material although the highest toughness values are lower after irradiation. Ductile stable crack growth occurred except for a few cases where both the irradiation temperature and test temperature are relatively low.

  18. Small punch test evaluation of neutron-irradiation-induced embrittlement of a Cr-Mo low-alloy steel

    SciTech Connect

    Song, S.-H. . E-mail: shsonguk@yahoo.co.uk; Faulkner, R.G.; Flewitt, P.E.J.; Marmy, P.; Weng, L.-Q.

    2004-09-15

    Neutron-irradiation-induced embrittlement of a 2.25Cr1Mo steel is investigated by means of small punch testing along with scanning electron microscopy. There is an apparent irradiation-induced embrittlement effect after the steel is irradiated at about 400 deg. C for 86 days with a neutron dose rate of 1.75x10{sup -8} dpa/s. The embrittlement is mainly nonhardening embrittlement caused by impurity grain boundary segregation.

  19. Initial tensile test results from J316 stainless steel irradiated in the HFIR spectrally tailored experiment

    SciTech Connect

    Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.

    1995-04-01

    The objective of this work is to determine the effects of neutron irradiation on the mechanical properties of austenitic stainless steel alloys. In this experiment, the spectrum has been tailored to reduce the thermal neutron flux and achieve a He/dpa level near that expected in a fusion reactor.

  20. Irradiation-induced grain growth in nanocrystalline reduced activation ferrite/martensite steel

    SciTech Connect

    Liu, W. B.; Chen, L. Q.; Zhang, C. Yang, Z. G.; Ji, Y. Z.; Zang, H.; Shen, T. L.

    2014-09-22

    In this work, we investigate the microstructure evolution of surface-nanocrystallized reduced activation ferrite/martensite steels upon high-dose helium ion irradiation (24.3 dpa). We report a significant irradiation-induced grain growth in the irradiated buried layer at a depth of 300–500 nm, rather than at the peak damage region (at a depth of ∼840 nm). This phenomenon can be explained by the thermal spike model: minimization of the grain boundary (GB) curvature resulting from atomic diffusion in the cascade center near GBs.

  1. Radiation hardening and deformation behavior of irradiated ferritic-martensitic steels

    SciTech Connect

    Robertson, J.P.; Klueh, R.L.; Rowcliffe, A.F.; Shiba, K.

    1998-03-01

    Tensile data from several 8--12% Cr alloys irradiated in the High Flux Isotope Reactor (HFIR) to doses up to 34 dpa at temperatures ranging from 90 to 600 C are discussed in this paper. One of the critical questions surrounding the use of ferritic-martensitic steels in a fusion environment concerns the loss of uniform elongation after irradiation at low temperatures. Irradiation and testing at temperatures below 200--300 C results in uniform elongations less than 1% and stress-strain curves in which plastic instability immediately follows yielding, implying dislocation channeling and flow localization. Reductions in area and total elongations, however, remain high.

  2. Hardness of Carburized Surfaces in 316LN Stainless Steel after Low Temperature Neutron Irradiation

    SciTech Connect

    Byun, TS

    2005-01-31

    A proprietary surface carburization treatment is being considered to minimize possible cavitation pitting of the inner surfaces of the stainless steel target vessel of the SNS. The treatment gives a large supersaturation of carbon in the surface layers and causes substantial hardening of the surface. To answer the question of whether such a hardened layer will remain hard and stable during neutron irradiation, specimens of the candidate materials were irradiated in the High Flux Isotope Reactor (HFIR) to an atomic displacement level of 1 dpa. Considerable radiation hardening occurred in annealed 316LN stainless steel and 20% cold rolled 316LN stainless steel, and lesser radiation hardening in Kolsterised layers on these materials. These observations coupled with optical microscopy examinations indicate that the carbon-supersaturated layers did not suffer radiation-induced decomposition and softening.

  3. The effects of neutron irradiation on fracture toughness of austenitic stainless steels.

    SciTech Connect

    Chopra, O. K.; Gruber, E. E.; Shack, W. J.

    1999-05-21

    Austenitic stainless steels are used extensively as structural alloys in reactor pressure vessel internal components because of their superior fracture toughness properties. However, exposure to high levels of neutron irradiation for extended periods leads to significant reduction in the fracture resistance of these steels. This paper presents results of fracture toughness J-R curve tests on four heats of Type 304 stainless steel that were irradiated to fluence levels of {approx}0.3 and 0.9 x 10{sup 21} n cm{sup {minus}2} (E >1 MeV) at {approx}288 C in a helium environment in the Halden heavy water boiling reactor. The tests were performed on 1/4-T compact tension specimens in air at 288 C; crack extensions were determined by both DC potential and elastic unloading compliance techniques.

  4. Microstructure and microhardness of CLAM steel irradiated up to 20.8 dpa in STIP-V

    NASA Astrophysics Data System (ADS)

    Peng, Lei; Ge, Hongen; Dai, Yong; Huang, Qunying; Ye, Minyou

    2016-01-01

    Specimens of China low activation martensitic (CLAM) steel were irradiated in the fifth experiment of SINQ target irradiation program (STIP-V) up to 20.8 dpa/1564 appm He. Microhardness measurements and transmission electron microscope (TEM) observations have been performed to investigate irradiation induced hardening effects. The results of CLAM steel specimens show similar trend in microhardness and microstructure changes with irradiation dose, compared to F82H/Optimax-A steels irradiated in STIP-I/II. Defects and helium bubbles were observed in all specimens, even at a very low dose of 5.4 dpa. For defects and bubbles, the mean size and number density increased with increasing irradiation dose to 13 dpa, and then the mean size increased and number density decreased with the increasing irradiation dose to 20.8 dpa.

  5. Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

    SciTech Connect

    Chopra, O. K.; Shack, W. J.

    2008-01-21

    In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heat-affected-zone materials, that were irradiated to fluence levels as high as {approx} 2x 10{sup 21} n/cm{sup 2} (E > 1 MeV) ({approx} 3 dpa) in a light water reactor at 288-300 C. The results are compared with the data available in the literature. The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined. A superposition model was used to represent the cyclic CGRs of austenitic SSs. The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that bounds the existing data has been defined. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated.

  6. Analytical description of true stress-true strain curves for neutron-irradiated stainless austenitic steels

    SciTech Connect

    Gussev, Maxim N; Byun, Thak Sang; Busby, Jeremy T

    2012-01-01

    This paper summarizes the results of an investigation for the deformation hardening behaviors of neutron-irradiated stainless steels in terms of true stress( ) true strain( ) curves. It is commonly accepted that the - curves are more informative for describing plastic flow, but there are few papers devoted to using the true curves for describing constitutive behaviors of materials. This study uses the true curves obtained from stainless steel samples irradiated to doses in the range of 0 55 dpa by various means: finite element calculation, optic extensomentry, and recalculation of engineering curves. It is shown that for the strain range 0 0.6 the true curves can be well described by the Swift equation: =k ( - 0)0.5. The influence of irradiation on the parameters of the Swift equation is investigated in detail. It is found that in most cases the k-parameter of this equation is not changed significantly by irradiation. Since large data scattering was observed for the 0-parameter, a modified Swift equation =k*( - 0 2/k2)0.5 was proposed and evaluated. This equation is based on the concept of zero stress, which is, in general, close to yield stress. The relationships among k, 0, and damage dose are discussed in detail, so as to more accurately describe the true curves for irradiated stainless steels.

  7. Structure and composition of phases occurring in austenitic stainless steels in thermal and irradiation environments

    SciTech Connect

    Lee, E.H.; Maziasz, P.J.; Rowcliffe, A.F.

    1980-01-01

    Transmission electron diffraction techniques coupled with quantitative x-ray energy dispersive spectroscopy have been used to characterize the phases which develop in austenitic stainless steels during exposure to thermal and to irradiation environments. In AISI 316 and Ti-modified stainless steels some thirteen phases have been identified and characterized in terms of their crystal structure and chemical composition. Irradiation does not produce any completely new phases. However, as a result of radiation-induced segregation principally of Ni and Si, and of enhanced diffusion rates, several major changes in phase relationships occur during irradiation. Firstly, phases characteristic of remote regions of the phase diagram appear unexpectedly and dissolve during postirradiation annealing (radiation-induced phases). Secondly, some phases develop with their compositions significantly altered by the incorporation of Ni or Si (radiation-modified phases).

  8. Response of 9Cr-ODS Steel to Proton Irradiation at 400 °C

    SciTech Connect

    Jianchao He; Farong Wan; Kumar Sridharan; Todd R. Allen; A. Certain; Y. Q. Wu

    2014-09-01

    The stability of Y–Ti–O nanoclusters, dislocation structure, and grain boundary segregation in 9Cr-ODS steels has been investigated following proton irradiation at 400 °C with damage levels up to 3.7 dpa. A slight coarsening and a decrease in number density of nanoclusters were observed as a result of irradiation. The composition of nanoclusters was also observed to change with a slight increase of Y and Cr concentration in the nanoclusters following irradiation. Size, density, and composition of the nanoclusters were investigated as a function of nanocluster size, specifically classified to three groups. In addition to the changes in nanoclusters, dislocation loops were observed after irradiation. Finally, radiation-induced enrichment of Cr and depletion of W were observed at grain boundaries after irradiation.

  9. Characterization of ion beam irradiated 304 stainless steel utilizing nanoindentation and Laue microdiffraction

    NASA Astrophysics Data System (ADS)

    Lupinacci, A.; Chen, K.; Li, Y.; Kunz, M.; Jiao, Z.; Was, G. S.; Abad, M. D.; Minor, A. M.; Hosemann, P.

    2015-03-01

    Characterizing irradiation damage in materials utilized in light water reactors is critical for both material development and application reliability. Here we use both nanoindentation and Laue microdiffraction to characterize both the mechanical response and microstructure evolution due to irradiation. Two different irradiation conditions were considered in 304 stainless steel: 1 dpa and 10 dpa. In addition, an annealed condition of the 10 dpa specimen for 1 h at 500 °C was evaluated. Nanoindentation revealed an increase in hardness due to irradiation and also revealed that hardness saturated in the 10 dpa case. Broadening using Laue microdiffraction peaks indicates a significant plastic deformation in the irradiated area that is in good agreement with both the SRIM calculations and the nanoindentation results.

  10. Post-irradiation annealing effect on helium diffusivity in austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Katsura, R.; Morisawa, J.; Kawano, S.; Oliver, B. M.

    2004-08-01

    As an experimental basis for helium induced weld cracking of neutron irradiated austenitic stainless steels, helium diffusivity has been evaluated by measuring helium release at high temperature. Isochronal and isothermal experiments were performed at temperatures between 700 and 1300 °C for 304 and 316L stainless steels. In 1 h isochronal experiments, helium was released beginning at ˜900 °C and reaching almost 100% at 1300 °C. No apparent differences in helium release were observed between the two stainless steel types. At temperatures between 900 and 1300 °C, the diffusion rate was calculated from the time dependence of the helium release rate to be: D0=4.91 cm 2/s, E=289 kJ/mol. The observed activation energy suggests that the release of helium from the steels is associated with the removal of helium from helium bubbles and/or from vacancy diffusion.

  11. Post-Irradiation Annealing Effect on Helium Diffusivity in Austenitic Stainless Steels

    SciTech Connect

    Katsura, Ryoei; Morisawa, J; Kawano, S; Oliver, Brian M.

    2004-08-01

    As an experimental basis for helium induced weld cracking of neutron irradiated austenitic stainless steels, helium diffusivity has been evaluated by measuring helium release rates at high temperature. Isochronal and isothermal experiment were performed at temperatures between 700 and 1300 for Type 304 and 316L stainless steels. In 1 hour isochronal experiments, helium was released beginning at {approx}900 and reaching near 100% at 1300. No apparent differences in helium release rate were observed between Type 304 and 316L stainless steels. At temperatures between 1100 and 1300, the diffusion rate was calculated from the time dependence of the helium release rate to be:?D0=3.42?104 cm2/s, E=173.2 kJ/mol. The observed activation energy suggests that the release of helium from the steels is associated with the removal of helium from helium bubbles.

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

    SciTech Connect

    Grossbeck, M.L.; Gibson, L.T.; 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.

  13. High temperature deformation behavior, thermal stability and irradiation performance in Grade 92 steel

    NASA Astrophysics Data System (ADS)

    Alsagabi, Sultan

    The 9Cr-2W ferritic-martensitic steel (i.e. Grade 92 steel) possesses excellent mechanical and thermophysical properties; therefore, it has been considered to suit more challenging applications where high temperature strength and creep-rupture properties are required. The high temperature deformation mechanism was investigated through a set of tensile testing at elevated temperatures. Hence, the threshold stress concept was applied to elucidate the operating high temperature deformation mechanism. It was identified as the high temperature climb of edge dislocations due to the particle-dislocation interactions and the appropriate constitutive equation was developed. In addition, the microstructural evolution at room and elevated temperatures was investigated. For instance, the microstructural evolution under loading was more pronounced and carbide precipitation showed more coarsening tendency. The growth of these carbide precipitates, by removing W and Mo from matrix, significantly deteriorates the solid solution strengthening. The MX type carbonitrides exhibited better coarsening resistance. To better understand the thermal microstructural stability, long tempering schedules up to 1000 hours was conducted at 560, 660 and 760°C after normalizing the steel. Still, the coarsening rate of M23C 6 carbides was higher than the MX-type particles. Moreover, the Laves phase particles were detected after tempering the steel for long periods before they dissolve back into the matrix at high temperature (i.e. 720°C). The influence of the tempering temperature and time was studied for Grade 92 steel via Hollomon-Jaffe parameter. Finally, the irradiation performance of Grade 92 steel was evaluated to examine the feasibility of its eventual reactor use. To that end, Grade 92 steel was irradiated with iron (Fe2+) ions to 10, 50 and 100 dpa at 30 and 500°C. Overall, the irradiated samples showed some irradiation-induced hardening which was more noticeable at 30°C. Additionally

  14. Monitoring microstructural evolution in irradiated steel with second harmonic generation

    SciTech Connect

    Matlack, Kathryn H.; Kim, Jin-Yeon; Jacobs, Laurence J.; Wall, James J.; Qu, Jianmin

    2015-03-31

    Material damage in structural components is driven by microstructural evolution that occurs at low length scales and begins early in component life. In metals, these microstructural features are known to cause measurable changes in the acoustic nonlinearity parameter. Physically, the interaction of a monochromatic ultrasonic wave with microstructural features such as dislocations, precipitates, and vacancies, generates a second harmonic wave that is proportional to the acoustic nonlinearity parameter. These nonlinear ultrasonic techniques thus have the capability to evaluate initial material damage, particularly before crack initiation and propagation occur. This paper discusses how the nonlinear ultrasonic technique of second harmonic generation can be used as a nondestructive evaluation tool to monitor microstructural changes in steel, focusing on characterizing neutron radiation embrittlement in nuclear reactor pressure vessel steels. Current experimental evidence and analytical models linking microstructural evolution with changes in the acoustic nonlinearity parameter are summarized.

  15. Study of irradiation effects in China low activation martensitic steel CLAM

    NASA Astrophysics Data System (ADS)

    Huang, Qunying; Li, Jiangang; Chen, Yixue

    2004-08-01

    Reduced activation ferritic/martensitic steels (RAFM steels) are presently considered as the primary structural materials for a demonstration (DEMO) fusion plant and the first fusion power reactors because of their attractive properties. Studies on various properties of China low activation martensitic steel (CLAM) are underway. The activation level of CLAM steel was calculated with the widely used inventory code FISPACT with the latest data library FENDL/A-2 based on the first wall (FW) neutron spectrum of the fusion-driven subcritical system (FDS) from the Monte Carlo transport code MCNP/4C calculation with FENDL-2 data library. The results were compared with the activation levels of other RAFM steels, such as EUROFER97, F82H, JLF-1 and 9Cr-2WVTa etc., under the same irradiation conditions. Furthermore, the dominant nuclides to γ-ray dose rate of CLAM steel were analyzed. The required control levels of impurities in CLAM steel will soon be implemented based on the hands-on and remote recycling dose rate limits.

  16. IRRADIATION CREEP AND SWELLING OF RUSSIAN FERRITIC-MARTENSITIC STEELS IRRADIATED TO VERY HIGH EXPOSURES IN THE BN-350 FAST REACTOR AT 305-335 DEGREES C

    SciTech Connect

    Konobeev, Yu V.; Dvoraishin, A. M.; Porollo, S. I.; Shulepin, S. V.; Budylkin, N. I.; Mironova, E. G.; Garner, Francis A.; Toloczko, Mychailo B.

    2003-09-03

    Russian ferritic martensitic (F(slash)M) steels EP(dash)450, EP(dash)852 and EP(dash)823 were irradiated in the BN(dash)350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb(dash)Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP(dash)450 and EP(dash) 823 at temperatures between 390 and 520 degrees C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP(dash)450 and EP(dash)852 at temperatures between 305 and 335 degrees C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, it appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation related densification. These irradiation creep studies confirm that the creep compliance of F(slash)M steels is about one half that of austenitic steels.

  17. Heavy-Section Steel Irradiation Program: Volume 3, Progress report, October 1991--September 1992

    SciTech Connect

    Corwin, W.R.

    1995-02-01

    The primary goal of the Heavy-Section Steel Irradiation Program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into 10 tasks: (1) program management, (2) K{sub Ic} curve shift in high-copper welds, (3) K{sub Ia} curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K{sub Ic} and K{sub Ia} curve shifts in low upper-shelf welds, (6) irradiation effects in a commercial low upper-shelf weld, (7) microstructural analysis of irradiation effects, (8) in-service aged material evaluations, (9) correlation monitor materials, and (10) special technical assistance. This report provides an overview of the activities within each of these tasks from October 1991 to September 1992.

  18. Microstructural analysis of ferritic-martensitic steels irradiated at low temperature in HFIR

    SciTech Connect

    Hashimoto, N.; Robertson, J.P.; Rowcliffe, A.F.; Wakai, E.

    1998-09-01

    Disk specimens of ferritic-martensitic steel, HT9 and F82H, irradiated to damage levels of {approximately}3 dpa at irradiation temperatures of either {approximately}90 C or {approximately}250 C have been investigated by using transmission electron microscopy. Before irradiation, tempered HT9 contained only M{sub 23}C{sub 6} carbide. Irradiation at 90 C and 250 C induced a dislocation loop density of 1 {times} 10{sup 22} m{sup {minus}3} and 8 {times} 10{sup 21} m{sup {minus}3}, respectively. in the HT9 irradiated at 250 C, a radiation-induced phase, tentatively identified as {alpha}{prime}, was observed with a number density of less than 1 {times} 10{sup 20} m{sup {minus}3}. On the other hand, the tempered F82H contained M{sub 23}C{sub 6} and a few MC carbides; irradiation at 250 C to 3 dpa caused minor changes in these precipitates and induced a dislocation loop density of 2 {times} 10{sup 22} m{sup {minus}3}. Difference in the radiation-induced phase and the loop microstructure may be related to differences in the post-yield deformation behavior of the two steels.

  19. Heavy-section steel irradiation program. Semiannual progress report, October 1996--March 1997

    SciTech Connect

    Rosseel, T.M.

    1998-02-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV`s fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established. Its primary goal is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior and, in particular, the fracture toughness properties of typical pressure-vessel steels as they relate to light-water RPV integrity. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into eight tasks: (1) program management, (2) irradiation effects in engineering materials, (3) annealing, (4) microstructural analysis of radiation effects, (5) in-service irradiated and aged material evaluations, (6) fracture toughness curve shift method, (7) special technical assistance, and (8) foreign research interactions. The work is performed by the Oak Ridge National Laboratory.

  20. Positron study of steel NF 709 after irradiation and thermal strain

    NASA Astrophysics Data System (ADS)

    Veternikova, J.; Degmova, J.; Simko, F.; Pekarcikova, M.; Sojak, S.; Slugen, V.

    2015-12-01

    New improved austenitic steel NF 709 was studied in term of thermal and radiation stability in consideration of its application as structural material for the newest generation of nuclear reactors - Generation IV. Samples of steel NF 709 were exposed to two strains: annealing at 1000 °C in argon atmosphere and simulated irradiation performed by helium ion implantation. Changes of the microstructure after the experimental strains were observed by positron annihilation techniques. The microstructure after both treatments indicated growing of vacancy defects; although these changes were small or in the range of error bar. Thus, material NF 709 can be considered as well resistant to these applied strains.

  1. Cracking behavior of thermally aged and irradiated CF-8 cast austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Chen, Y.; Alexandreanu, B.; Chen, W.-Y.; Natesan, K.; Li, Z.; Yang, Y.; Rao, A. S.

    2015-11-01

    To assess the combined effect of thermal aging and neutron irradiation on the cracking behavior of CF-8 cast austenitic stainless steel, crack growth rate (CGR) and fracture toughness J-R curve tests were carried out on compact-tension specimens in high-purity water with low dissolved oxygen. Both unaged and thermally aged specimens were irradiated at ∼320 °C to 0.08 dpa. Thermal aging at 400 °C for 10,000 h apparently had no effect on the corrosion fatigue and stress corrosion cracking behavior in the test environment. The cracking susceptibility of CF-8 was not elevated significantly by neutron irradiation at 0.08 dpa. Transgranular cleavage-like cracking was the main fracture mode during the CGR tests, and a brittle morphology of delta ferrite was often seen on the fracture surfaces at the end of CGR tests. The fracture toughness J-R curve tests showed that both thermal aging and neutron irradiation can induce significant embrittlement. The loss of fracture toughness due to neutron irradiation was more pronounced in the unaged than aged specimens. After neutron irradiation, the fracture toughness values of the unaged and aged specimens were reduced to a similar level. G-phase precipitates were observed in the aged and irradiated specimens with or without prior aging. The similar microstructural changes resulting from thermal aging and irradiation suggest a common microstructural mechanism of inducing embrittlement in CF-8.

  2. Re-weldability of neutron irradiated Type 304 and 316L stainless steels

    NASA Astrophysics Data System (ADS)

    Morishima, Y.; Koshiishi, M.; Kashiwakura, K.; Hashimoto, T.; Kawano, S.

    2004-08-01

    Weldability of irradiated stainless steel (SS) has been studied to develop the technical guideline regarding the repair-welding of reactor internals. Type 304 and 316L SSs were irradiated at ambient temperature in the US Advanced Test Reactor. The multi-pass bead-on-plate TIG (GTA) and YAG laser welding with heat input levels less than 1 MJ/m were performed on specimens containing helium up to 18 appm. In this paper, results of cross-sectional micrograph observations of the heat affected zone were considered in light of helium bubble properties. The tendency for weld crack formation of irradiated Type 316L SS was compared with that of irradiated Type 304 SS.

  3. Effects of silicon, carbon and molybdenum additions on IASCC of neutron irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Nakano, J.; Miwa, Y.; Kohya, T.; Tsukada, T.

    2004-08-01

    To study the effects of minor elements on irradiation assisted stress corrosion cracking (IASCC), high purity type 304 and 316 stainless steels (SSs) were fabricated and minor elements, Si or C were added. After neutron irradiation to 3.5 × 10 25 n/m 2 ( E>1 MeV), slow strain rate tests (SSRTs) of irradiated specimens were conducted in oxygenated high purity water at 561 K. Specimen fractured surfaces were examined using a scanning electron microscope (SEM) after the SSRTs. The fraction of intergranular stress corrosion cracking (IGSCC) on the fractured surface after the SSRTs increased with neutron fluence. In high purity SS with added C, the fraction of IGSCC was the smallest in the all SSs, although irradiation hardening level was the largest of all the SSs. Addition of C suppressed the susceptibility to IGSCC.

  4. ATR-A1 irradiation experiment on vanadium alloys and low activation steels

    SciTech Connect

    Tasi, H.; Strain, R.V.; Gomes, I.; Hins, A.G.; Smith, D.L.

    1996-04-01

    To study the mechanical properties of vanadium alloys under neutron irradiation at low temperatures, an experiment was designed and constructed for irradiation in the Advanced Test Reactor (ATR). The experiment contained Charpy, tensile, compact tension, TEM, and creep specimens of vanadium alloys. It also contained limited low-activation ferritic steel specimens as part of the collaborative agreement with Monbusho of Japan. The design irradiation temperatures for the vanadium alloy specimens in the experiment are {approx}200 and 300{degrees}C, achieved with passive gap-gap sizing and fill gas blending. To mitigate vanadium-to-chromium transmutation from the thermal neutron flux, the test specimens are contained inside gadolinium flux filters. All specimens are lithium-bonded. The irradiation started in Cycle 108A (December 3, 1995) and is expected to have a duration of three ATR cycles and a peak influence of 4.4 dpa.

  5. Effect of irradiation temperature on void swelling of China Low Activation Martensitic steel (CLAM)

    SciTech Connect

    Zhao Fei; Qiao Jiansheng; Huang Yina; Wan Farong Ohnuki, Soumei

    2008-03-15

    CLAM is one composition of a Reduced Activation Ferritic/Martensitic steel (RAFM), which is being studied in a number of institutes and universities in China. The effect of electron-beam irradiation temperature on irradiation swelling of CLAM was investigated by using a 1250 kV High Voltage Electron Microscope (HVEM). In-situ microstructural observations indicated that voids formed at each experimental temperature - 723 K, 773 K and 823 K. The size and number density of voids increased with increasing irradiation dose at each temperature. The results show that CLAM has good swelling resistance. The maximum void swelling was produced at 723 K; the swelling was about 0.3% when the irradiation damage was 13.8 dpa.

  6. Effects of proton irradiation on nanocluster precipitation in ferritic steel containing fcc alloying additions

    SciTech Connect

    Zhang, Zhongwu; Liu, C T; Wang, Xun-Li; Miller, Michael K; Ma, Dong; Chen, Guang; Williams, J R; Chin, Bryan

    2012-01-01

    Newly-developed precipitate-strengthened ferritic steels with and without pre-existing nanoscale precipitates were irradiated with 4 MeV protons to a dose of ~5 mdpa at 50 C and subsequently examined by nanoindentation and atom probe tomography (APT). Irradiation-enhanced precipitation and coarsening of pre-existing nanoscale precipitates were observed. Copper partitions to the precipitate core along with a segregation of Ni, Al and Mn to the precipitate/matrix interface after both thermal aging and proton irradiation. Proton irradiation induces the precipitation reaction and coarsening of pre-existing nanoscale precipitates, and these results are similar to a thermal aging process. The precipitation and coarsening of nanoscale precipitates are responsible for the changes in hardness. The observation of the radiation-induced softening is essentially due to the coarsening of the pre-existing Cu-rich nanoscale precipitates. The implication of the precipitation on the embrittlement of reactor-pressure-vessel steels after irradiation is discussed.

  7. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    SciTech Connect

    Tsai, H.; Strain, R.V.; Gomes, I.; Chung, H.; Smith, D.L.

    1997-04-01

    The ATR-A1 irradiation experiment in the Advanced Test Reactor (ATR) was a collaborative U.S./Japan effort to study at low temperatures the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation was completed on May 5, 1996, as planned, after achieving an estimated neutron damage of 4.7 dpa in vanadium. The capsule has since been kept in the ATR water canal for the required radioactivity cool-down. Planning is underway for disassembly of the capsule and test specimen retrieval.

  8. Helium behavior in ferritic/martensitic steels irradiated in spallation target

    NASA Astrophysics Data System (ADS)

    Krsjak, Vladimir; Kuriplach, Jan; Shen, Tielong; Sabelova, Veronika; Sato, Koichi; Dai, Yong

    2015-01-01

    Two positron annihilation spectroscopy (PAS) techniques have been used for the investigation of helium behavior in STIP samples. Positron lifetime measurements and coincidence Doppler broadening spectroscopy have been employed together in a complex PAS characterization of RAFM steel irradiated in a mixed neutron-proton spectrum up to 20 dpa and 1800 appm He. Both techniques show an increase of the He-to-dpa ratio up to ∼10 dpa. At higher irradiation loads, the ratio is decreasing, which was attributed to the formation and growth of helium bubbles.

  9. Heavy-section steel technology and irradiation programs-retrospective and prospective views

    SciTech Connect

    Nanstad, Randy K; Bass, Bennett Richard; Rosseel, Thomas M; Merkle, John Graham; Sokolov, Mikhail A

    2007-01-01

    In 1965, the Atomic Energy Commission (AEC), at the advice of the Advisory Committee on Reactor Safeguards (ACRS), initiated the process that resulted in the establishment of the Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratory (ORNL). In 1989, the Heavy-Section Steel Irradiation (HSSI) Program, formerly the HSST task on irradiation effects, was formed as a separate program, and, in 2007, the HSST/HSSI Programs, sponsored by the U.S. Nuclear Regulatory Commission (USNRC), celebrated 40 years of continuous research oriented toward the safety of light-water nuclear reactor pressure vessels (RPV). This paper presents a summary of results from those programs with a view to future activities.

  10. Properties of copper?stainless steel HIP joints before and after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Tähtinen, S.; Laukkanen, A.; Singh, B. N.; Toft, P.

    2002-12-01

    The tensile and fracture behaviour of CuCrZr and CuAl25 IG0 alloys joint to 316L(N) stainless steel by hot isostatic pressing (HIP) have been determined in unirradiated and neutron-irradiated conditions. The tensile and fracture behaviour of copper alloy HIP joint specimens are dominated by the properties of the copper alloys, and particularly, by the strength mismatch and mismatch in strain hardening capacities between copper alloys and stainless steel. The test temperature, neutron irradiation and thermal cycles primarily affect the copper alloy HIP joint properties through changing the strength mismatch between the base alloys. Changes in the loading conditions i.e. tensile, bend ( JI) and mixed-mode bend ( JI/ JII) lead to different fracture modes in the copper alloy HIP joint specimens.

  11. Welding-induced microstructure in austenitic stainless steels before and after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Stoenescu, R.; Schäublin, R.; Gavillet, D.; Baluc, N.

    2007-02-01

    The effects of neutron irradiation on the microstructure of welded joints made of austenitic stainless steels have been investigated. The materials were welded AISI 304 and AISI 347, so-called test weld materials, and irradiated with neutrons at 300 °C to 0.3 and 1.0 dpa. In addition, an AISI 304 type from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 300 °C, was investigated. The microstructure of heat-affected zones and base materials was analysed before and after irradiation, using transmission electron microscopy. Neutron diffraction was performed for internal stress measurements. It was found that the heat-affected zone contains, relative to the base material, a higher dislocation density, which relates well to a higher residual stress level and, after irradiation, a higher irradiation-induced defect density. In both materials, the irradiation-induced defects are of the same type, consisting in black dots and Frank dislocation loops. Careful analysis of the irradiation-induced defect contrast was performed and it is explained why no stacking fault tetrahedra could be identified.

  12. Tensile properties of a titanium modified austenitic stainless steel and the weld joints after neutron irradiation

    SciTech Connect

    Shiba, Kiyoyuki; Ioka, Ikuo; Jitsukawa, Shiro; Hamada, Shozo; Hishinuma, Atkinichi; Robertson, J.P.

    1999-10-01

    Tensile specimens of a titanium modified austenitic stainless steel and its weldments fabricated with Tungsten Inert Gas (TIG) and Electron Beam (EB) welding techniques were irradiated to a peak dose of 19 dpa and a peak helium level of 250 appm in the temperature range between 200 and 400 C in spectrally tailored capsules in the Oak Ridge Research Reactor (ORR) and the High Flux Isotope Reactor (HFIR). The He/dpa ratio of about 13 appm/dpa is similar to the typical helium/.dpa ratio of a fusion reactor environment. The tensile tests were carried out at the irradiation temperature in vacuum. The irradiation caused an increase in yield stress to levels between 670 and 800 MPa depending on the irradiation temperature. Total elongation was reduced to less than 10%, however the specimens failed in a ductile manner. The results were compared with those of the specimens irradiated using irradiation capsules producing larger amount of He. Although the He/dpa ratio affected the microstructural change, the impact on the post irradiation tensile behavior was rather small not only for base metal specimens but also for the weld joint and the weld metal specimens.

  13. Analysis of Tensile Deformation and Failure in Austenitic Stainless Steels: Part II- Irradiation Dose Dependence

    SciTech Connect

    Kim, Jin Weon; Byun, Thak Sang

    2010-01-01

    Irradiation effects on stable and unstable deformations and fracture behaviors in irradiated austenitic stainless steels (SSs) have been studied in detail based on the equivalent true stress versus true strain curves. An iterative technique in finite element simulation was used to obtain the equivalent true stress-true strain data from experimental tensile curves. It was shown that the strain hardening rate was retained at a high level on unstable deformation after significant irradiation and was independent of the irradiation dose up to the initiation of a localized necking. The equivalent fracture stress was nearly independent of irradiation dose before the damage (embrittlement) mechanism changed. In low dose range (< ~ 2dpa), the fracture strain and tensile fracture energy decreased rapidly with dose and at higher doses they decreased gradually to saturated levels, which were still high for irradiated materials. It was also found that the fracture properties for EC316LN SS were less sensitive to irradiation dose than those for 316 SS, although their uniform tensile properties showed almost the same dose dependencies. It was confirmed that the dose dependence of tensile fracture properties evaluated by the linear approximation model for nominal stress was accurate enough for practical use without elaborate calculations.

  14. Microstructural stability of a self-ion irradiated lanthana-bearing nanostructured ferritic steel

    SciTech Connect

    Pasebani, Somayeh; Charit, Indrajit; Burns, Jatuporn; Alsagabi, Sultan; Butt, Darryl P.; Cole, James I.; Price, Lloyd M.; Shao, Lin

    2015-07-01

    Thermally stable nanofeatures with high number density are expected to impart excellent high temperature strength and irradiation stability in nanostructured ferritic steels (NFSs) which have potential applications in advanced nuclear reactors. A lanthana-bearing NFS (14LMT) developed via mechanical alloying and spark plasma sintering was used in this study. The sintered samples were irradiated by Fe2+ ions to 10, 50 and 100 dpa at 30 °C and 500 °C. Microstructural and mechanical characteristics of the irradiated samples were studied using different microscopy techniques and nanoindentation, respectively. Overall morphology and number density of the nanofeatures remained unchanged after irradiation. Average radius of nanofeatures in the irradiated sample (100 dpa at 500 °C) was slightly reduced. A notable level of irradiation hardening and enhanced dislocation activity occurred after ion irradiation except at 30 °C and ≥50 dpa. Other microstructural features like grain boundaries and high density of dislocations also provided defect sinks to assist in defect removal.

  15. Microstructural stability of a self-ion irradiated lanthana-bearing nanostructured ferritic steel

    NASA Astrophysics Data System (ADS)

    Pasebani, Somayeh; Charit, Indrajit; Burns, Jatuporn; Alsagabi, Sultan; Butt, Darryl P.; Cole, James I.; Price, Lloyd M.; Shao, Lin

    2015-07-01

    Thermally stable nanofeatures with high number density are expected to impart excellent high temperature strength and irradiation stability in nanostructured ferritic steels (NFSs) which have potential applications in advanced nuclear reactors. A lanthana-bearing NFS (14LMT) developed via mechanical alloying and spark plasma sintering was used in this study. The sintered samples were irradiated by Fe2+ ions to 10, 50 and 100 dpa at 30 °C and 500 °C. Microstructural and mechanical characteristics of the irradiated samples were studied using different microscopy techniques and nanoindentation, respectively. Overall morphology and number density of the nanofeatures remained unchanged after irradiation. Average radius of nanofeatures in the irradiated sample (100 dpa at 500 °C) was slightly reduced. A notable level of irradiation hardening and enhanced dislocation activity occurred after ion irradiation except at 30 °C and ⩾50 dpa. Other microstructural features like grain boundaries and high density of dislocations also provided defect sinks to assist in defect removal.

  16. Tensile properties of a titanium modified austenitic stainless steel and the weld joints after neutron irradiation

    SciTech Connect

    Shiba, K.; Ioka, I.; Jitsukawa, S.; Hamada, A.; Hishinuma, A.

    1996-10-01

    Tensile specimens of a titanium modified austenitic stainless steel and its weldments fabricated with Tungsten Inert Gas (TIG) and Electron Beam (EB) welding techniques were irradiated to a peak dose of 19 dpa and a peak helium level of 250 appm in the temperature range between 200 and 400{degrees}C in spectrally tailored capsules in the Oak Ridge Research Reactor (ORR) and the High Flux Isotope Reactor (HFIR). The He/dpa ratio of about 13 appm/dpa is similar to the typical helium/dpa ratio of a fusion reactor environment. The tensile tests were carried out at the irradiation temperature in vacuum. The irradiation caused an increase in yield stress to levels between 670 and 800 MPa depending on the irradiation temperature. Total elongation was reduced to less than 10%, however the specimens failed in a ductile manner. The results were compared with those of the specimens irradiated using irradiation capsules producing larger amount of He. Although the He/dpa ratio affected the microstructural change, the impact on the post irradiation tensile behavior was rather small for not only base metal specimens but also for the weld joint and the weld metal specimens.

  17. Thermal annealing recovery of fracture toughness in HT9 steel after irradiation to high doses

    NASA Astrophysics Data System (ADS)

    Byun, Thak Sang; Baek, Jong-Hyuk; Anderoglu, Osman; Maloy, Stuart A.; Toloczko, Mychailo B.

    2014-06-01

    The HT9 ferritic/martensitic steel with a nominal chemistry of Fe(bal.)12%Cr1%MoVW has been used as a primary core material for fast fission reactors such as FFTF because of its high resistance to radiation-induced swelling and embrittlement. Both static and dynamic fracture test results have shown that the HT9 steel can become brittle when it is exposed to high dose irradiation at a relatively low temperature (<430 °C). This article aims at a comprehensive discussion on the thermal annealing recovery of fracture toughness in the HT9 steel after irradiation up to 3148 dpa at 378504 °C. A specimen reuse technique has been established and applied to this study: the fracture specimens were tested Charpy specimens or broken halves of Charpy bars (13 × 3 × 4 mm). The post-anneal fracture test results indicated that much of the radiation-induced damage can be recovered by a simple thermal annealing schedule: the fracture toughness was incompletely recovered by 550 °C annealing, while nearly complete or complete recovery occurred after 650 °C annealing. This indicates that thermal annealing is a feasible damage mitigation technique for the reactor components made of HT9 steel. The partial recovery is probably due to the non-removable microstructural damages such as void or gas bubble formation, elemental segregation and precipitation.

  18. Effect of recrystallization on ion-irradiation hardening and microstructural changes in 15Cr-ODS steel

    NASA Astrophysics Data System (ADS)

    Ha, Yoosung; Kimura, Akihiko

    2015-12-01

    The effects of recrystallization on ion-irradiation hardening and microstructural changes were investigated for a 15Cr-ODS ferritic steel. Dual ion-irradiation experiments were performed at 470 °C using 6.4 MeV Fe3+ ions simultaneously with energy-degraded 1 MeV He+ ions. The displacement of damage at 600 nm depth from the specimen surface was 30 dpa. Nano-indentation test with Berkovich type indentation tip was measured by constant stiffness measurement (CSM) technique. Results from nano-indentation tests indicate irradiation hardening in ODS steels even at 470 °C, while it wasn't observed in reduced activation ferritic steel. Recrystallized ODS steel shows a larger irradiation hardening, which is considered to be due to the reduction of grain boundaries and interfaces of matrix/oxide particles. In 20% cold rolled ODS steel after recrystallization, both the hardening and bubble number density were lower than those of recrystallized ODS steel, suggesting that dislocations generated by cold rolling suppress bubble formation. Based on the estimation of irradiation hardening from TEM observation results, it is considered that the bubbles are not the main factor controlling ion-irradiation hardening.

  19. Infrared nanosecond pulsed laser irradiation of stainless steel: micro iron-oxide zones generation.

    PubMed

    Ortiz-Morales, M; Frausto-Reyes, C; Soto-Bernal, J J; Acosta-Ortiz, S E; Gonzalez-Mota, R; Rosales-Candelas, I

    2014-07-15

    Nanosecond-pulsed, infrared (1064 nm) laser irradiation was used to create periodic metal oxide coatings on the surface of two samples of commercial stainless steel at ambient conditions. A pattern of four different metal oxide zones was created using a galvanometer scanning head and a focused laser beam over each sample. This pattern is related to traverse direction of the laser beam scanning. Energy-dispersive X-ray spectroscopy (EDS) was used to find the elemental composition and Raman spectroscopy to characterize each oxide zone. Pulsed laser irradiation modified the composition of the stainless steel samples, affecting the concentration of the main components within each heat affected zone. The Raman spectra of the generated oxides have different intensity profiles, which suggest different oxide phases such as magnetite and maghemite. In addition, these oxides are not sensible to the laser power of the Raman system, as are the iron oxide powders reported in the literature. These experiments show that it is possible to generate periodic patterns of various iron oxide zones by laser irradiation, of stainless steel at ambient conditions, and that Raman spectroscopy is a useful punctual technique for the analysis and inspection of small oxide areas. PMID:24699286

  20. Infrared nanosecond pulsed laser irradiation of stainless steel: Micro iron-oxide zones generation

    NASA Astrophysics Data System (ADS)

    Ortiz-Morales, M.; Frausto-Reyes, C.; Soto-Bernal, J. J.; Acosta-Ortiz, S. E.; Gonzalez-Mota, R.; Rosales-Candelas, I.

    2014-07-01

    Nanosecond-pulsed, infrared (1064 nm) laser irradiation was used to create periodic metal oxide coatings on the surface of two samples of commercial stainless steel at ambient conditions. A pattern of four different metal oxide zones was created using a galvanometer scanning head and a focused laser beam over each sample. This pattern is related to traverse direction of the laser beam scanning. Energy-dispersive X-ray spectroscopy (EDS) was used to find the elemental composition and Raman spectroscopy to characterize each oxide zone. Pulsed laser irradiation modified the composition of the stainless steel samples, affecting the concentration of the main components within each heat affected zone. The Raman spectra of the generated oxides have different intensity profiles, which suggest different oxide phases such as magnetite and maghemite. In addition, these oxides are not sensible to the laser power of the Raman system, as are the iron oxide powders reported in the literature. These experiments show that it is possible to generate periodic patterns of various iron oxide zones by laser irradiation, of stainless steel at ambient conditions, and that Raman spectroscopy is a useful punctual technique for the analysis and inspection of small oxide areas.

  1. Predictive Reactor Pressure Vessel Steel Irradiation Embrittlement Models: Issues and Opportunities

    SciTech Connect

    Odette, George Robert; Nanstad, Randy K

    2009-01-01

    Nuclear plant life extension to 80 years will require accurate predictions of neutron irradiation-induced increases in the ductile-brittle transition temperature ( T) of reactor pressure vessel (RPV) steels at high fluence conditions that are far outside the existing database. Remarkable progress in mechanistic understanding of irradiation embrittlement has led to physically motivated T correlation models that provide excellent statistical fi ts to the existing surveillance database. However, an important challenge is developing advanced embrittlement models for low fl ux-high fl uence conditions pertinent to extended life. These new models must also provide better treatment of key variables and variable combinations and account for possible delayed formation of late blooming phases in low copper steels. Other issues include uncertainties in the compositions of actual vessel steels, methods to predict T attenuation away from the reactor core, verifi cation of the master curve method to directly measure the fracture toughness with small specimens and predicting T for vessel annealing remediation and re-irradiation cycles.

  2. Positron annihilation study of neutron irradiated model alloys and of a reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Lambrecht, M.; Almazouzi, A.

    2009-03-01

    The hardening and embrittlement of reactor pressure vessel steels are of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation-induced damage, and positron annihilation spectroscopy has been shown to be a suitable method for analysing most of these defects. In this paper, this technique (both positron annihilation lifetime spectroscopy and coincidence Doppler broadening) has been used to investigate neutron irradiated model alloys, with increasing chemical complexity and a reactor pressure vessel steel. It is found that the clustering of copper takes place at the very early stages of irradiation using coincidence Doppler broadening, when this element is present in the alloy. On the other hand, considerations based on positron annihilation spectroscopy analyses suggest that the main objects causing hardening are most probably self-interstitial clusters decorated with manganese in Cu-free alloys. In low-Cu reactor pressure vessel steels and in (Fe, Mn, Ni, Cu) alloys, the main effect is still due to Cu-rich precipitates at low doses, but the role of manganese-related features becomes pre-dominant at high doses.

  3. Fracture properties of a neutron-irradiated stainless steel submerged arc weld cladding overlay

    SciTech Connect

    Corwin, W.R.; Berggren, R.G.; Nanstad, R.K.

    1984-01-01

    The ability of stainless steel cladding to increase the resistance of an operating nuclear reactor pressure vessel to extension of surface flaws depends greatly on the properties of the irradiated cladding. Therefore, weld overlay cladding irradiated at temperatures and fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the submerged arc, single-wire, oscillating-electrode method. Three layers of cladding provided a thickness adequate for fabrication of test specimens. The first layer was type 309, and the upper two layers were type 308 stainless steel. The type 309 was diluted considerably by excessive melting of the base plate. Specimens were taken from near the base plate-cladding interface and also from the upper layers. Charpy V-notch and tensile specimens were irradiated at 288/sup 0/C to a fluence of 2 x 10/sup 23/ neutrons/m/sup 2/ (>1 MeV). 10 refs., 16 figs., 4 tabs.

  4. Deformation localization and dislocation channel dynamics in neutron-irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Gussev, Maxim N.; Field, Kevin G.; Busby, Jeremy T.

    2015-05-01

    The dynamics of deformation localization and dislocation channel formation were investigated in situ in a neutron-irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction, and transmission electron microscopy (TEM). Channel formation was observed at ∼70% of the polycrystalline yield stress of the irradiated materials (σ0.2). It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the σ0.2, channels often formed near the middle of the grain boundary. For a single grain, the role of elastic stiffness value (Young's modulus) in channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in "soft" grains with a high Schmid factor located near "stiff" grains with high elastic stiffness. The spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one-third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. In the AISI 304 steel, channels in grains oriented close to <0 0 1>||TA (tensile axis) and <1 0 1>||TA were twin free and grain with <1 1 1>||TA and grains oriented close to a Schmid factor maximum contained deformation twins.

  5. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    SciTech Connect

    Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

    2010-06-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (≤3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC

  6. Mechanical properties of 1950's vintage 304 stainless steel weldment components after low temperature neutron irradiation

    SciTech Connect

    Sindelar, R.L.; Caskey, G.R. Jr.; Thomas, J.K. ); Hawthorne, J.R.; Hiser, A.L. ); Lott, R.A.; Begley, J.A.; Shogan, R.P. . Science and Technology Center)

    1991-01-01

    The reactor vessels of the nuclear production reactors at the Savannah River Site (SRS) were constructed in the 1950's from Type 304 stainless steel plates welded with Type 308 stainless steel filler using the multipass metal inert gas process. An irradiated mechanical properties database has been developed for the vessel with materials from archival primary coolant system piping irradiated at low temperatures (75 to 150{degrees}C) in the State University of New York at Buffalo reactor (UBR) and the High Flux Isotope Reactor (HFIR) to doses of 0.065 to 2.1 dpa. Fracture toughness, tensile, and Charpy-V impact properties of the weldment components (base, weld, and weld heat-affected-zone (HAZ)) have been measured at temperatures of 25{degrees}C and 125{degrees}C in the L-C and C-L orientations for materials in both the irradiated and unirradiated conditions for companion specimens. Fracture toughness and tensile properties of specimens cut from an SRS reactor vessel sidewall with doses of 0.1 and 0.5 dpa were also measured at temperatures of 25 and 125{degrees}C. The irradiated materials exhibit hardening with loss of work hardenability and a reduction in toughness relative to the unirradiated materials. The HFIR-irradiated materials show an increase in yield strength between about 20% and 190% with a concomitant tensile strength increase between about 15% to 30%. The elastic-plastic fracture toughness parameters and Charpy-V energy absorption both decrease and show only a slight sensitivity to dose. The irradiation-induced decrease in the elastic-plastic fracture toughness (J{sub def} at 1 mm crack extension) is between 20% to 65%; the range of J{sub 1C} values are 72.8 to 366 kJ/m{sup 2} for the irradiated materials. Similarly, Charpy V-notch results show a 40% to 60% decrease in impact energies.

  7. Mechanical properties of 1950`s vintage 304 stainless steel weldment components after low temperature neutron irradiation

    SciTech Connect

    Sindelar, R.L.; Caskey, G.R. Jr.; Thomas, J.K.; Hawthorne, J.R.; Hiser, A.L.; Lott, R.A.; Begley, J.A.; Shogan, R.P.

    1991-12-31

    The reactor vessels of the nuclear production reactors at the Savannah River Site (SRS) were constructed in the 1950`s from Type 304 stainless steel plates welded with Type 308 stainless steel filler using the multipass metal inert gas process. An irradiated mechanical properties database has been developed for the vessel with materials from archival primary coolant system piping irradiated at low temperatures (75 to 150{degrees}C) in the State University of New York at Buffalo reactor (UBR) and the High Flux Isotope Reactor (HFIR) to doses of 0.065 to 2.1 dpa. Fracture toughness, tensile, and Charpy-V impact properties of the weldment components (base, weld, and weld heat-affected-zone (HAZ)) have been measured at temperatures of 25{degrees}C and 125{degrees}C in the L-C and C-L orientations for materials in both the irradiated and unirradiated conditions for companion specimens. Fracture toughness and tensile properties of specimens cut from an SRS reactor vessel sidewall with doses of 0.1 and 0.5 dpa were also measured at temperatures of 25 and 125{degrees}C. The irradiated materials exhibit hardening with loss of work hardenability and a reduction in toughness relative to the unirradiated materials. The HFIR-irradiated materials show an increase in yield strength between about 20% and 190% with a concomitant tensile strength increase between about 15% to 30%. The elastic-plastic fracture toughness parameters and Charpy-V energy absorption both decrease and show only a slight sensitivity to dose. The irradiation-induced decrease in the elastic-plastic fracture toughness (J{sub def} at 1 mm crack extension) is between 20% to 65%; the range of J{sub 1C} values are 72.8 to 366 kJ/m{sup 2} for the irradiated materials. Similarly, Charpy V-notch results show a 40% to 60% decrease in impact energies.

  8. Effect of heat treatment and irradiation temperature on impact properties of Cr-W-V ferritic steels

    NASA Astrophysics Data System (ADS)

    Klueh, R. L.; Alexander, D. J.

    Charpy impact tests were conducted on eight normalized-and-tempered ferritic and martensitic steels irradiated in two different normalized conditions. Irradiation was conducted in the Fast Flux Test Facility (FFTF) at 393°C to ≈14 dpa on eight steels with 2.25%, 5%, 9%, and 12% Cr (0.1% C) with varying amounts of W, V, and Ta. The different normalization treatments involved changing the cooling rate after austenitization. The faster cooling rate produced 100% bainite in the 2.25Cr steels, compared to duplex structures of bainite and polygonal ferrite for the slower cooling rate. For both cooling rates, martensite formed in the 5% and 9% Cr steels, and martensite with ≈25% δ-ferrite formed in the 12% Cr steel. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy (USE). The difference in microstructure in the low-chromium steels due to the different heat treatments had little effect on properties. For the high-chromium martensitic steels, only the 5Cr steel was affected by heat treatment. When the results at 393°C were compared with previous results at 365°C, all but a 5Cr and a 9Cr steel showed the expected decrease in the shift in DBTT with increasing temperature.

  9. Degradation of mechanical properties of stainless steel cladding due to neutron irradiation and thermal aging

    SciTech Connect

    Haggag, F.M.

    1994-09-01

    Thermal aging of three-wire series-arc stainless steel weld overlay cladding at 288{degrees}C for 1605 h resulted in an appreciable decrease (16%) in the Charpy V-notch (CVN) upper-shelf energy (USE), but the effect on the 41-J transition temperature shift was very small (3{degrees}C). The combined effect following neutron irradiation at 288{degrees}C to a fluence of 5 X 10{sup 19} neutrons/cm{sup 2} (>1 MeV) was a 22% reduction in the USE and a 29{degrees}C shift in the 41-J transition temperature. The effect of thermal aging on tensile properties was very small. However, the combined effect of irradiation and aging was an increase in the yield strength (6 to 34% at test temperatures from 288 to -125{degrees}C) and no apparent change in ultimate tensile strength or total elongation. Neutron irradiation reduced the initiation fracture toughness (J{sub {kappa}}) much more than did thermal aging alone. However, irradiation slightly decreased the tearing modulus but no reduction was caused by thermal aging alone. The effects of long-term thermal exposure times (20,000 and 50,000 h) will be investigated when the specimens become available. Also, long-term thermal exposure of the three-wire cladding as well as type 308 stainless steel weld materials at 343{degrees}C is in progress.

  10. Alloying effect of Ni and Cr on irradiated microstructural evolution of type 304 stainless steels

    NASA Astrophysics Data System (ADS)

    Tan, L.; Busby, J. T.

    2013-11-01

    Life extension of the existing nuclear power plants imposes significant challenges to core structural materials that suffer increased fluences. This paper presents the microstructural evolution of a type 304 stainless steel and its variants alloyed with extra Ni and Cr under neutron irradiation at ˜320 °C for up to 10.2 dpa. Similar to the reported data of type 304 variants, a large amount of Frank loops, ultrafine G-phase/M23C6 particles, and limited amount of cavities were observed in the irradiated samples. The irradiation promoted the growth of pre-existing M23C6 at grain boundaries and resulted in some phase transformation to CrC in the alloy with both extra Ni and Cr. A new type of ultrafine precipitates, possibly (Ti,Cr)N, was observed in all the samples, and its amount was increased by the irradiation. Additionally, α-ferrite was observed in the type 304 steel but not in the Ni or Ni + Cr alloyed variants. The effect of Ni and Cr alloying on the microstructural evolution is discussed.

  11. Analysis of tensile deformation and failure in austenitic stainless steels: Part II - Irradiation dose dependence

    NASA Astrophysics Data System (ADS)

    Kim, Jin Weon; Byun, Thak Sang

    2010-01-01

    Irradiation effects on the stable and unstable deformation and fracture behavior of austenitic stainless steels (SSs) have been studied in detail based on the equivalent true stress versus true strain curves. An iterative finite element simulation technique was used to obtain the equivalent true stress-true strain data from experimental tensile curves. The simulation result showed that the austenitic stainless steels retained high strain hardening rate during unstable deformation even after significant irradiation. The strain hardening rate was independent of irradiation dose up to the initiation of a localized necking. Similarly, the equivalent fracture stress was nearly independent of dose before the damage (embrittlement) mechanism changed. The fracture strain and tensile fracture energy decreased with dose mostly in the low dose range <˜2 dpa and reached nearly saturation values at higher doses. It was also found that the fracture properties for EC316LN SS were less sensitive to irradiation than those for 316 SS, although their uniform tensile properties showed almost the same dose dependencies. It was confirmed that the dose dependence of tensile fracture properties evaluated by the linear approximation model for nominal stress was accurate enough for practical use without elaborate calculations.

  12. Hardening and microstructural evolution of A533b steels irradiated with Fe ions and electrons

    NASA Astrophysics Data System (ADS)

    Watanabe, H.; Arase, S.; Yamamoto, T.; Wells, P.; Onishi, T.; Odette, G. R.

    2016-04-01

    Radiation hardening and embrittlement of A533B steels is heavily dependent on the Cu content. In this study, to investigate the effect of copper on the microstructural evolution of these materials, A533B steels with different Cu levels were irradiated with 2.4 MeV Fe ions and 1.0 MeV electrons. Ion irradiation was performed from room temperature (RT) to 350 °C with doses up to 1 dpa. At RT and 290 °C, low dose (<0.1 dpa) hardening trend corresponded with ΔH ∝ (dpa)n, with n initially approximately 0.5 and consistent with a barrier hardening mechanism, but saturating at ≈0.1 dpa. At higher dose levels, the radiation-induced hardening exhibited a strong Cu content dependence at 290 °C, but not at 350 °C. Electron irradiation using high-voltage electron microscopy revealed the growth of interstitial-type dislocation loops and enrichment of Ni, Mn, and Si in the vicinities of pre-existing dislocations at doses for which the radiation-induced hardness due to ion irradiation was prominent.

  13. Heavy-section steel irradiation program. Semiannual progress report, October 1995--March 1996

    SciTech Connect

    Corwin, W.R.

    1997-04-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents which have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPVs fracture resistance which occurs during service, since without that radiation damage, it is virtually impossible to postulate a realistic scenario that would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior and, in particular, the fracture toughness properties of typical pressure-vessel steels as they relate to light-water RPV integrity. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties.

  14. Correlation between locally deformed structure and oxide film properties in austenitic stainless steel irradiated with neutrons

    NASA Astrophysics Data System (ADS)

    Chimi, Yasuhiro; Kitsunai, Yuji; Kasahara, Shigeki; Chatani, Kazuhiro; Koshiishi, Masato; Nishiyama, Yutaka

    2016-07-01

    To elucidate the mechanism of irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water for neutron-irradiated austenitic stainless steels (SSs), the locally deformed structures, the oxide films formed on the deformed areas, and their correlation were investigated. Tensile specimens made of irradiated 316L SSs were strained 0.1%-2% at room temperature or at 563 K, and the surface structures and crystal misorientation among grains were evaluated. The strained specimens were immersed in high-temperature water, and the microstructures of the oxide films on the locally deformed areas were observed. The appearance of visible step structures on the specimens' surface depended on the neutron dose and the applied strain. The surface oxides were observed to be prone to increase in thickness around grain boundaries (GBs) with increasing neutron dose and increasing local strain at the GBs. No penetrative oxidation was observed along GBs or along surface steps.

  15. Deformation Microstructure of a Reduced-Activation Ferritic/Martensitic Steel Irradiated in HFIR

    SciTech Connect

    Hashimoto, N.; Klueh, R.L.; Ando, M.; Tanigawa, H.; Sawai, T.; Shiba, K.

    2003-09-15

    In order to determine the contributions of different microstructural features to strength and to deformation mode, microstructure of deformed flat tensile specimens of irradiated reduced activation F82H (IEA heat) base metal (BM) and its tungsten inert-gas (TIG) weldments (weld metal and weld joint) were investigated by transmission electron microscopy (TEM), following fracture surface examination by scanning electron microscopy (SEM). After irradiation, the fracture surfaces of F82H BM and TIG weldment showed a martensitic mixed quasi-cleavage and ductile-dimple fracture. The microstructure of the deformed region of irradiated F82H BM contained dislocation channels. This suggests that dislocation channeling could be the dominant deformation mechanism in this steel, resulting in the loss of strain-hardening capacity. While, the necked region of the irradiated F82H TIG, where showed less hardening than F82H BM, showed deformation bands only. From these results, it is suggested that the pre-irradiation microstructure, especially the dislocation density, could affect the post-irradiation deformation mode.

  16. Microstructure of HFIR-irradiated 12-Cr 1 MoVW ferritic steel

    SciTech Connect

    Vitek, J.M.; Klueh, R.L.

    1983-01-01

    As part of the fusion materials development program in the United States, a 12 Cr-1 MoVW ferritic steel was irradiated in the High Flux Isotope Reactor (HFIR) to a damage level of 36 dpa at 300, 400, 500, and 600/sup 0/C. During irradiation in HFIR, a transmutation reaction of nickel results in the production of helium, to a level of 99 at. ppM in the present experiment. The microstructures were evaluated after irradiation and the results are presented. Cavities were found at all temperatures. Small cavities (3 to 9 nm) were observed after irradiation at 300, 500 and 600/sup 0/C. At 500 and 600/sup 0/C, the cavities were found preferentially at dislocations, lath boundaries, and prior austenite grain boundaries. After irradiation at 400/sup 0/C, larger cavities (4 to 30 nm) were observed homogeneously distributed throughout the tempered martensite structure. The maximum swelling was 0.07% after irradiation at 400/sup 0/C. Comparision of the results with other studies in which helium was not present at such high levels indicated helium enhances the swelling of 12 Cr-1 MoVW.

  17. Irradiation testing of 316L(N)-IG austenitic stainless steel for ITER

    NASA Astrophysics Data System (ADS)

    van Osch, E. V.; Horsten, M. G.; de Vries, M. I.

    1998-10-01

    In the frame work of the European Fusion Technology Programme and the International Thermonuclear Experimental Reactor (ITER), ECN is investigating the irradiation behaviour of the structural materials for ITER. The main structural material for ITER is austenitic stainless steel Type 316L(N)-IG. The operating temperatures of (parts of) the components are envisaged to range between 350 and 700 K. A significant part of the dose-temperature domain of irradiation conditions relevant for ITER has already been explored, there is, however, very little data at about 600 K. Available data tend to indicate a maximum in the degradation of the mechanical properties after irradiation at this temperature, e.g. a minimum in ductility and a maximum of hardening. Therefore an irradiation program for plate material 316L(N)-IG, its Electron Beam (EB) weld and Tungsten Inert Gas (TIG) weld metal, and also including Hot Isostatically Pressed (HIP) 316L(N) powder and solid-solid joints, was set up in 1995. Irradiations have been carried out in the High Flux Reactor (HFR) in Petten at a temperature of 600 K, at dose levels from 1 to 10 dpa. The paper presents the currently available post-irradiation test results. Next to tensile and fracture toughness data on plate, EB and TIG welds, first results of powder HIP material are included.

  18. Heavy-section steel irradiation program. Progress report, April 1996--September 1996

    SciTech Connect

    Corwin, W.R.

    1997-09-01

    The Heavy-Section Steel Irradiation Program was established to quantitatively assess the effects of neutron irradiation on the material behavior of typical reactor pressure vessel (RPV) steels. During this period, fracture mechanics testing of specimens of the irradiated low upper shelf (LUS) weld were completed and analyses performed. Heat treatment of five RPV plate materials was initiated to examine phosphorus segregation effects on the fracture toughness of the heat affected zone of welds. Initial results show that all five materials exhibited very large prior austenite grain sizes as a consequence of the initial heat treatment. Irradiated and annealed specimens of LUS weld material were tested and analyzed. Four sets of Charpy V-notch (CVN) specimens were aged at various temperatures and tested to examine the reason for overrecovery of upper shelf energy that has been observed. Molecular dynamics cascade simulations were extended to 40 keV and have provided information representative of most of the fast neutron spectrum. Investigations of the correlation between microstructural changes and hardness changes in irradiated model alloys was also completed. Preliminary planning for test specimen machining for the Japan Power Development Reactor was completed. A database of Charpy impact and fracture toughness data for RPV materials that have been tested in the unirradiated and irradiated conditions is being assembled and analyzed. Weld metal appears to have similar CVN and fracture toughness transition temperature shifts, whereas the fracture toughness shifts are greater than CVN shifts for base metals. Draft subcontractor reports on precracked cylindrical tensile specimens were completed, reviewed, and are being revised. Testing on precracked CVN specimens, both quasi-static and dynamic, was evaluated. Additionally, testing of compact specimens was initiated as an experimental comparison of constraint limitations. 16 figs., 2 tabs.

  19. Heavy-section steel irradiation program. Semiannual progress report, September 1993--March 1994

    SciTech Connect

    Corwin, W.R.

    1995-04-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only component in the primary pressure boundary for which, if it should rupture, the engineering safety systems cannot assure protection from core damage. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. In particular, ft is vital to fully understand the degree of irradiation-induced degradation of the RPV`s fracture resistance that occurs during service. The Heavy-Section Steel (HSS) Irradiation Program has been established; its primary goal is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties of typical pressure-vessel steels, as they relate to light-water RPV integrity. The program includes the direct continuation of irradiation studies previously conducted within the HSS Technology Program augmented by enhanced examinations of the accompanying microstructural changes. During this period, the report on the duplex-type crack-arrest specimen tests from Phase 11 of the K{sub la} program was issued, and final preparations for testing the large, irradiated crack-arrest specimens from the Italian Committee for Research and Development of Nuclear Energy and Alternative Energies were completed. Tests on undersize Charpy V-notch (CVN) energy specimens in the irradiated and annealed weld 73W were completed. The results are described in detail in a draft NUREG report. In addition, the ORNL investigation of the embrittlement of the High Flux Isotope RPV indicated that an unusually large ratio of the high-energy gamma-ray flux to fast-neutron flux is most likely responsible for the apparently accelerated embrittlement.

  20. Removal of Metal-Oxide Layers Formed on Stainless and Carbon Steel Surfaces by Excimer Laser Irradiation in Various Atmospheres

    SciTech Connect

    Kameo, Yutaka; Nakashima, Mikio; Hirabayashi, Takakuni

    2002-02-15

    To apply the laser ablation technique for decontamination of metal wastes contaminated with radioactive nuclides, the effect of irradiation atmospheres on removal of oxide layers on steel surfaces by laser ablation was studied. Based on the assumption that the absorption of laser light follows the Lambert-Beer law, ablation parameters, such as absorption length and threshold fluence for ablation, of sintered Fe{sub 2}O{sub 3} and stainless and carbon steels were measured in He, O{sub 2}, Kr, or SF{sub 6} atmospheres. The results indicated that SF{sub 6} was the most effective gas of all irradiation atmospheres studied for the exclusive removal of oxide layers formed on stainless and carbon steel samples in high-temperature pressurized water. Secondary ion mass spectroscopic measurement and scanning electron microscopic observation confirmed that no oxide layer existed on the steel samples after the exclusive removal with laser irradiation.

  1. Evolution of Nickel-Manganese-Silicon Dominated Phases in Highly Irradiated Reactor Pressure Vessel Steels

    SciTech Connect

    Peter B Wells; Yuan Wu; Tim Milot; G. Robert Odette; Takuya Yamamoto; Brandon Miller; James Cole

    2014-11-01

    Formation of a high density of Ni-Mn-Si nm-scale precipitates in irradiated reactor pressure vessel steels, both with and without Cu, could lead to severe embrittlement. Models long ago predicted that these precipitates, which are not treated in current embrittlement regulations, would emerge only at high fluence. However, the mechanisms and variables that control Ni-Mn- Si precipitate formation, and their detailed characteristics, have not been well understood. High flux irradiations of six steels with systematic variations in Cu and Ni were carried out at ˜ 295±5°C to high and very high neutron fluences of ˜ 1.3x1020 and 1.1x1021 n/cm2. Atom probe tomography (APT) shows that significant mole fractions of these precipitates form in the Cu bearing steels at ˜ 1.3x1020 n/cm2, while they are only beginning to develop in Cu-free steels. However, large mole fractions, far in excess of those found in previous studies, are observed at 1.1x1021 n/cm2 at all Cu levels. The precipitates diffract, and in one case are compositionally and structurally consistent with the Mn6Ni16Si7 G-phase. At the highest fluence, the large precipitate mole fractions primarily depend on the steel Ni content, rather than Cu, and lead to enormous strength increases up to about 700 MPa. The implications of these results to light water reactor life extension are discussed briefly.

  2. Analysis of stress-induced Burgers vector anisotropy in pressurized tube specimens of irradiated ferritic-martensitic steel: JLF-1

    SciTech Connect

    Gelles, D.S.; Shibayama, T.

    1998-09-01

    A procedure for determining the Burgers vector anisotropy in irradiated ferritic steels allowing identification of all a<100> and all a/2<111> dislocations in a region of interest is applied to a pressurized tube specimen of JLF-1 irradiated at 430 C to 14.3 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) or 61 dpa. Analysis of micrographs indicates large anisotropy in Burgers vector populations develop during irradiation creep.

  3. Deformation localization and dislocation channel dynamics in neutron-irradiated austenitic stainless steels

    DOE PAGESBeta

    Gussev, Maxim N.; Field, Kevin G.; Busby, Jeremy T.

    2015-02-24

    We investigated dynamics of deformation localization and dislocation channel formation in situ in a neutron irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction and transmission electron microscopy. Channel formation was observed at 70% of the formal tensile yield stress for both alloys. It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the yield stress, channels often formed near the middle of the grain boundary. For amore » single grain, the role of elastic stiffness value (Young modulus) in the channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in soft grains with a high Schmid factor located near stiff grains with high elastic stiffness. Moreover, the spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. Finally, it was shown that in the AISI 304 steel, channels were twin-free in grains oriented close to [001] and [101] of standard unit triangle; [111]-grains and grains oriented close to Schmid factor maximum contained deformation twins.« less

  4. Deformation localization and dislocation channel dynamics in neutron-irradiated austenitic stainless steels

    SciTech Connect

    Gussev, Maxim N.; Field, Kevin G.; Busby, Jeremy T.

    2015-02-24

    We investigated dynamics of deformation localization and dislocation channel formation in situ in a neutron irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction and transmission electron microscopy. Channel formation was observed at 70% of the formal tensile yield stress for both alloys. It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the yield stress, channels often formed near the middle of the grain boundary. For a single grain, the role of elastic stiffness value (Young modulus) in the channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in soft grains with a high Schmid factor located near stiff grains with high elastic stiffness. Moreover, the spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. Finally, it was shown that in the AISI 304 steel, channels were twin-free in grains oriented close to [001] and [101] of standard unit triangle; [111]-grains and grains oriented close to Schmid factor maximum contained deformation twins.

  5. Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

    SciTech Connect

    Byun, Thak Sang; Toloczko, M; Maloy, S

    2013-01-01

    Static fracture toughness tests have been performed for high dose HT9 steel using miniature disk compact tension (DCT) specimens to expand the knowledge base for fast reactor core materials. The HT9 steel DCT specimens were from the ACO-3 duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3 148 dpa at 378 504oC. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa m occurred in room temperature tests when irradiation temperature was below 400 C, while ductile fracture with stable crack growth was observed in all tests at higher irradiation temperatures. No fracture toughness less than 100 MPa m was measured when the irradiation temperature was above 430 C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the dose range 3 148 dpa. A post upper-shelf behavior was observed for the non-irradiated and high temperature (>430 C) irradiation cases, which indicates that the ductile-brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.

  6. Heavy-Section Steel Irradiation Program. Semiannual progress report, April--September 1990: Volume 1, No. 2

    SciTech Connect

    Corwin, W.R.

    1993-11-01

    The primary goal of the Heavy-Section Steel Irradiation (HSSI) program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. The program includes the direct continuation of irradiation studies previously conducted within the Heavy-Section Steel Technology program augmented by enhanced examinations of the accompanying microstructural changes. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and post-irradiation annealing are being examined on a wide range of fracture properties. During this period detailed statistical analyses of the fracture data on K{sub lc} shift of high-copper welds revealed greater shifts in fracture toughness than in Charpy transition temperatures. Testing of the duplex specimens from the second phase of the irradiated crack arrest testing on high-copper welds was initiated. Short-term aging studies were conducted on stainless steel weld-overlay cladding. Additional determinations were made of chemistry and unirradiated RT{sub NDT}s of the low upper-shelf weld metal from the Midland reactor and fracture toughness testing begun. An initial model describing the evolution of radiation-induced self-defect/solute clusters and other microstructures was developed and experiments initiated to examine the effects of low-energy, low-temperature neutron irradiations.

  7. Dual beam irradiation of nanostructured FeCrAl oxide dispersion strengthened steel

    NASA Astrophysics Data System (ADS)

    Chen, C.-L.; Richter, A.; Kögler, R.; Talut, G.

    2011-05-01

    Nanostructured ferritic oxide dispersion strengthened (ODS) alloy is an ideal candidate for fission/fusion power plant materials, particularly in the use of a first-wall and blanket structure of a next generation reactor. These steels usually contain a high density of Y-Ti-O and Y-Al-O nanoparticles, high dislocation densities and fine grains. The material contains nanoparticles with an average diameter of 21 nm and was treated by several cold rolling procedures, which modify the dislocation density. Structural analysis with HRTEM shows that the chemical composition of the initial Y 2O 3 oxide is modified to perovskite YAlO 3 (YAP) and Y 2Al 5O 12 garnet (YAG). Irradiation of these alloys was performed with a dual beam irradiation of 2.5 MeV Fe +/31 dpa and 350 keV He +/18 appm/dpa. Irradiation causes atomic displacements resulting in vacancy and self-interstitial lattice defects and dislocation loops. Extended SRIM calculations for ODS steel indicate a clear spatial separation between the excess vacancy distribution close to the surface and the excess interstitials in deeper layers of the material surface. The helium atoms are supposed to accumulate mainly in the vacancies. Additionally to structural changes, the effect of the irradiation generated defects on the mechanical properties of the ODS is investigated by nanoindentation. A clear hardness increase in the irradiated area is observed, which reaches a maximum at a close surface region. This feature is attributed to synergistic effects between the displacement damage and He implantation resulting in He filled vacancies. Fine He cavities with diameters of a few nanometers were identified in TEM images.

  8. Crack growth rates of irradiated austenitic stainless steel weld heat affected zone in BWR environments.

    SciTech Connect

    Chopra, O. K.; Alexandreanu, B.; Gruber, E. E.; Daum, R. S.; Shack, W. J.; Energy Technology

    2006-01-31

    Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of reactor pressure vessels because of their superior fracture toughness. However, exposure to high levels of neutron irradiation for extended periods can exacerbate the corrosion fatigue and stress corrosion cracking (SCC) behavior of these steels by affecting the material microchemistry, material microstructure, and water chemistry. Experimental data are presented on crack growth rates of the heat affected zone (HAZ) in Types 304L and 304 SS weld specimens before and after they were irradiated to a fluence of 5.0 x 10{sup 20} n/cm{sup 2} (E > 1 MeV) ({approx} 0.75 dpa) at {approx}288 C. Crack growth tests were conducted under cycling loading and long hold time trapezoidal loading in simulated boiling water reactor environments on Type 304L SS HAZ of the H5 weld from the Grand Gulf reactor core shroud and on Type 304 SS HAZ of a laboratory-prepared weld. The effects of material composition, irradiation, and water chemistry on growth rates are discussed.

  9. Technical Letter Report on the Cracking of Irradiated Cast Stainless Steels with Low Ferrite Content

    SciTech Connect

    Chen, Y.; Alexandreanu, B.; Natesan, K.

    2014-11-01

    Crack growth rate and fracture toughness J-R curve tests were performed on CF-3 and CF-8 cast austenite stainless steels (CASS) with 13-14% of ferrite. The tests were conducted at ~320°C in either high-purity water with low dissolved oxygen or in simulated PWR water. The cyclic crack growth rates of CF-8 were higher than that of CF-3, and the differences between the aged and unaged specimens were small. No elevated SCC susceptibility was observed among these samples, and the SCC CGRs of these materials were comparable to those of CASS alloys with >23% ferrite. The fracture toughness values of unirradiated CF-3 were similar between unaged and aged specimens, and neutron irradiation decreased the fracture toughness significantly. The fracture toughness of CF-8 was reduced after thermal aging, and declined further after irradiation. It appears that while lowering ferrite content may help reduce the tendency of thermal aging embrittlement, it is not very effective to mitigate irradiation-induced embrittlement. Under a combined condition of thermal aging and irradiation, neutron irradiation plays a dominant role in causing embrittlement in CASS alloys.

  10. Welding-induced mechanical properties in austenitic stainless steels before and after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Stoenescu, R.; Schäublin, R.; Gavillet, D.; Baluc, N.

    2007-03-01

    The effects of neutron irradiation on the mechanical properties of welded joints made of austenitic stainless steels have been investigated. The materials are welded AISI 304 and AISI 347, so-called test weld materials, irradiated with neutrons at 573 K to doses of 0.3 and 1.0 dpa. In addition, an AISI 304 from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 573 K, was investigated. The mechanical properties of heat-affected zones and base materials were analysed before and after irradiation. Tensile parameters were determined at room temperature and at 573 K, for all materials and irradiation conditions. In the test weld materials it is found that radiation hardening is lower and loss of ductility is higher in the heat-affected zone than in the base material. In the in-service material radiation hardening is about the same in heat-affected zone and base material. After irradiation, deformation takes place by stacking faults and twins, at both room temperature and high temperature, contrary to unirradiated materials, where deformation takes place by twinning at room temperature and by dislocation cells at high temperature. No defect free channels are observed.

  11. Properties of precipitation hardened steel irradiated at 323 K in the Japan materials testing reactor

    NASA Astrophysics Data System (ADS)

    Niimi, M.; Matsui, Y.; Jitsukawa, S.; Hoshiya, T.; Tsukada, T.; Ohmi, M.; Mimura, H.; Ooka, N.; Hide, K.

    A precipitation hardening type 630 stainless steel was irradiated in the Japan Materials Testing Reactor (JMTR) in contact with the reactor primary coolant. The temperature of the irradiated specimens was about 330 K. The fast neutron ( E > 1 MeV) fluence for the specimens ranged from 10 24 to 10 26 m -2. Tension tests and fracture toughness tests were carried out at room temperature, while Charpy impact tests were done at temperatures of 273-453 K. Tensile strength data showed a peak of 1600 MPa at around 7 × 10 24 m -2, then gradually decreased to about 1500 MPa at 1.2 × 10 26 m -2. The elongation decreased with irradiation from 12% for unirradiated material to 6% at 1.2 × 10 26 m -2. The fractography after the tension test revealed that the fracture was ductile. Fracture toughness decreased to about a half of the value for unirradiated material with irradiation. The cleavage fracture was dominant on the fractured surface. Charpy impact tests showed an increase of ductile-brittle transition temperature (DBTT) by 60 K with irradiation.

  12. Irradiation-assisted stress corrosion cracking behavior of austenitic stainless steels applicable to LWR core internals.

    SciTech Connect

    Chung, H. M.; Shack, W. J.; Energy Technology

    2006-01-31

    This report summarizes work performed at Argonne National Laboratory on irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels that were irradiated in the Halden reactor in simulation of irradiation-induced degradation of boiling water reactor (BWR) core internal components. Slow-strain-rate tensile tests in BWR-like oxidizing water were conducted on 27 austenitic stainless steel alloys that were irradiated at 288 C in helium to 0.4, 1.3, and 3.0 dpa. Fractographic analysis was conducted to determine the fracture surface morphology. Microchemical analysis by Auger electron spectroscopy was performed on BWR neutron absorber tubes to characterize grain-boundary segregation of important elements under BWR conditions. At 0.4 and 1.4 dpa, transgranular fracture was mixed with intergranular fracture. At 3 dpa, transgranular cracking was negligible, and fracture surface was either dominantly intergranular, as in field-cracked core internals, or dominantly ductile or mixed. This behavior indicates that percent intergranular stress corrosion cracking determined at {approx}3 dpa is a good measure of IASCC susceptibility. At {approx}1.4 dpa, a beneficial effect of a high concentration of Si (0.8-1.5 wt.%) was observed. At {approx}3 dpa, however, such effect was obscured by a deleterious effect of S. Excellent resistance to IASCC was observed up to {approx}3 dpa for eight heats of Types 304, 316, and 348 steel that contain very low concentrations of S. Susceptibility of Types 304 and 316 steels that contain >0.003 wt.% S increased drastically. This indicates that a sulfur related critical phenomenon plays an important role in IASCC. A sulfur content of <0.002 wt.% is the primary material factor necessary to ensure good resistance to IASCC. However, for Types 304L and 316L steel and their high-purity counterparts, a sulfur content of <0.002 wt.% alone is not a sufficient condition to ensure good resistance to IASCC. This is in distinct contrast to

  13. IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

    SciTech Connect

    Brumovsky, M.; Steele, L.E.

    1997-02-01

    In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracture mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.

  14. Effect of neutron irradiation at low temperature on the embrittlement of the reduced-activation ferritic steels

    NASA Astrophysics Data System (ADS)

    Rybin, V. V.; Kursevich, I. P.; Lapin, A. N.

    1998-10-01

    Effects of neutron irradiation to fluence of 2.0 × 10 24 n/m 2 ( E > 0.5 MeV) in temperature range 70-300°C on mechanical properties and structure of the experimental reduced-activation ferritic 0.1%C-(2.5-12)%Cr-(1-2)%W-(0.2-0.7)%V alloys were investigated. The steels were studied in different initial structural conditions obtained by changing the modes of heat treatments. Effect of neutron irradiation estimated by a shift in ductile-brittle transition temperature (ΔDBTT) and reduction of upper shelf energy (ΔUSE) highly depends on both irradiation condition and steel chemical composition and structure. For the steel with optimum chemical composition (9Cr-1.5WV) after irradiation to 2 × 10 24 n/m 2 ( E ⩾ 0.5 MeV) at 280°C the ΔDBTT does not exceed 25°C. The shift in DBTT increased from 35°C to 110°C for the 8Cr-1.5WV steel at a decrease in irradiation temperature from 300°C to 70°C. The CCT diagrams are presented for several reduced-activated steels.

  15. A review of irradiation effects on LWR core internal materials - IASCC susceptibility and crack growth rates of austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Chopra, O. K.; Rao, A. S.

    2011-02-01

    Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods changes the microstructure (radiation hardening) and microchemistry (radiation-induced segregation) of these steels, and degrades their fracture properties. Irradiation-assisted stress corrosion cracking (IASCC) is another degradation process that affects LWR internal components exposed to neutron radiation. The existing data on irradiated austenitic SSs were reviewed to evaluate the effects of key parameters such as material composition, irradiation dose, and water chemistry on IASCC susceptibility and crack growth rates of these materials in LWR environments. The significance of microstructural and microchemistry changes in the material on IASCC susceptibility is also discussed. The results are used to determine (a) the threshold fluence for IASCC and (b) the disposition curves for cyclic and IASCC growth rates for irradiated SSs in LWR environments.

  16. Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

    SciTech Connect

    Byun, Thak Sang; Toloczko, Mychailo B.; Saleh, Tarik A.; Maloy, Stuart A.

    2013-01-14

    To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3–148 dpa at 378–504 C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa pm occurred in room temperature tests when irradiation temperature was below 400 C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa pm was measured when the irradiation temperature was above 430 C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3–148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 *C) irradiation cases, which indicates that the ductile–brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.

  17. Internal stress distribution for generating closure domains in laser-irradiated Fe–3%Si(110) steels

    SciTech Connect

    Iwata, Keiji; Imafuku, Muneyuki; Orihara, Hideto; Sakai, Yusuke; Ohya, Shin-Ichi; Suzuki, Tamaki; Shobu, Takahisa; Akita, Koichi; Ishiyama, Kazushi

    2015-05-07

    Internal stress distribution for generating closure domains occurring in laser-irradiated Fe–3%Si(110) steels was investigated using high-energy X-ray analysis and domain theory based on the variational principle. The measured triaxial stresses inside the specimen were compressive and the stress in the rolling direction became more dominant than stresses in the other directions. The calculations based on the variational principle of magnetic energy for closure domains showed that the measured triaxial stresses made the closure domains more stable than the basic domain without closure domains. The experimental and calculation results reveal that the laser-introduced internal stresses result in the occurrence of the closure domains.

  18. Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale J.; Was, Gary S.

    2015-01-01

    The objective of this study was to compare the microstructures, microchemistry, hardening, susceptibility to IASCC initiation, and deformation behavior resulting from proton or reactor irradiation. Two commercial purity and six high purity austenitic stainless steels with various solute element additions were compared. Samples of each alloy were irradiated in the BOR-60 fast reactor at 320 °C to doses between approximately 4 and 12 dpa or by a 3.2 MeV proton beam at 360 °C to a dose of 5.5 dpa. Irradiated microstructures consisted mainly of dislocation loops, which were similar in size but lower in density after proton irradiation. Both irradiation types resulted in the formation of Ni-Si rich precipitates in a high purity alloy with added Si, but several other high purity neutron irradiated alloys showed precipitation that was not observed after proton irradiation, likely due to their higher irradiation dose. Low densities of small voids were observed in several high purity proton irradiated alloys, and even lower densities in neutron irradiated alloys, implying void nucleation was in process. Elemental segregation at grain boundaries was very similar after each irradiation type. Constant extension rate tensile experiments on the alloys in simulated light water reactor environments showed excellent agreement in terms of the relative amounts of intergranular cracking, and an analysis of localized deformation after straining showed a similar response of cracking to surface step height after both irradiation types. Overall, excellent agreement was observed after proton and reactor irradiation, providing additional evidence that proton irradiation is a useful tool for accelerated testing of irradiation effects in austenitic stainless steel.

  19. Microstructural Aspects of Irradiation Damage in A508 Gr 4N Forging Steel: Composition and Flux Effects

    SciTech Connect

    M.G. Burke; R.J. Stofanak; J.M. Hyde; C.A. English; W.L. Server

    2002-10-09

    Neutron irradiation can promote significant changes in the microstructure and associated mechanical properties of low alloy steels. In particular, irradiation can induce the formation of non-equilibrium phases and segregation, which may lead to a degradation in toughness. In this study, the microstructural changes caused by neutron irradiation have been characterized in A508 Grade (Gr) 4N-type steels ({approx}3.5% Ni) using a variety of state-of-the-art analytical techniques including 3D-Atom Probe Field-Ion Microscopy and Small Angle Neutron Scattering, along with post-irradiation annealing studies combining Positron Annihilation Lineshape Analysis and hardness measurements. Important differences between conventional and ''superclean'' A508 Gr 4N steel have been identified in this investigation. The data indicate that Ni is not the controlling factor in the irradiation damage behavior of these materials; rather, the Mn content of the steel is a dominant factor in the irradiation-induced microstructural development of solute-related hardening features.

  20. Microstructural evolution of type 304 and 316 stainless steels under neutron irradiation at LWR relevant conditions

    SciTech Connect

    Tan, Lizhen; Stoller, Roger E.; Field, Kevin G.; Yang, Ying; Morgan, Dane; Wirth, Brian D.; Gussev, Maxim N.; Busby, Jeremy T.; Nam, H.

    2015-12-11

    Extension of light water reactors' useful life will expose austenitic internal core components to irradiation damage levels beyond 100 displacements per atom (dpa), which will lead to profound microstructural evolution and consequent degradation of macroscopic properties. Microstructural evolution, including Frank loops, cavities, precipitates, and segregation at boundaries and the resultant radiation hardening in type 304 and 316 stainless steel (SS) variants, were studied in this work via experimental characterization and multiple simulation methods. Experimental data for up to 40 heats of type 304SS and 316SS variants irradiated in different reactors to 0.6–120 dpa at 275–375°C were either generated from this work or collected from literature reports. These experimental data were then combined with models of Frank loop and cavity evolution, computational thermodynamics and precipitation, and ab initio and rate theory integrated radiation-induced segregation models to provide insights into microstructural evolution and degradation at higher radiation doses.

  1. Grain boundary segregation in neutron-irradiated 304 stainless steel studied by atom probe tomography

    NASA Astrophysics Data System (ADS)

    Toyama, T.; Nozawa, Y.; Van Renterghem, W.; Matsukawa, Y.; Hatakeyama, M.; Nagai, Y.; Al Mazouzi, A.; Van Dyck, S.

    2012-06-01

    Radiation-induced segregation (RIS) of solute atoms at a grain boundary (GB) in 304 stainless steel (SS), neutron-irradiated to a dose of 24 dpa at 300 °C in the fuel wrapper plates of a commercial pressurized water reactor, was investigated using laser-assisted atom probe tomography (APT). Ni, Si, and P enrichment and Cr and Fe depletion at the GB were evident. The full-width at half-maximum of the RIS region was ˜3 nm for the concentration profile peaks of Ni and Si. The atomic percentages of Ni, Si, and Cr at the GB were ˜19%, ˜7%, and ˜14%, respectively, in agreement with previously-reported values for neutron-irradiated SS. A high number density of intra-granular Ni-Si rich precipitates formed in the matrix. A precipitate-denuded zone with a width of ˜10 nm appeared on both sides of the GB.

  2. Influence of crystal orientation on hardness and nanoindentation deformation in ion-irradiated stainless steels

    NASA Astrophysics Data System (ADS)

    Miura, Terumitsu; Fujii, Katsuhiko; Fukuya, Koji; Takashima, Keisuke

    2011-10-01

    The influence of crystal orientation on hardness and the range of plastic deformation caused by nanoindentation was investigated in a solution annealed type 316 stainless steel irradiated with Fe 2+ ions. The hardness was a function of grain orientation and was correlated with the Taylor factor averaged over three normal directions of the contact surface of the Berkovich indenter. The transmission electron microscope observations of the deformation microstructure under the indentations showed that the range of plastic deformation reached up to 10 times the indent depth for unirradiated material and depended on the orientation relation between the contact surface of the indenter and the slip directions. The range of plastic deformation decreased as the damage structure developed in ion irradiation.

  3. High-temperature fatigue life of type 316 stainless steel containing irradiation induced helium

    SciTech Connect

    Grossbeck, M.L.; Liu, K.C.

    1981-01-01

    Specimens of 20%-cold-worked AISI type 316 stainless steel were irradiated in the High Flux Isotope Reactor (HFIR) at 550/sup 0/C to a maximum damage level of 15 dpa and a transmutation produced helium level of 820 at. ppM. Fully reversed strain controlled fatigue tests were performed in a vacuum at 550/sup 0/C. No significant effect of the irradiation on low-cycle fatigue life was observed; however, the strain range of the 10/sup 7/ cycle endurance limit decreased from 0.35 to 0.30%. The relation between total strain range and number of cycles to failure was found to be ..delta..epsilon/sub T/ = 0.02N/sub f//sup -0/ /sup 12/ + N/sub f//sup -0/ /sup 6/ for N/sub f/ < 10/sup 7/ cycles.

  4. Microstructural Evolution of Type 304 and 316 Stainless Steels Under Neutron Irradiation at LWR Relevant Conditions

    NASA Astrophysics Data System (ADS)

    Tan, L.; Stoller, R. E.; Field, K. G.; Yang, Y.; Nam, H.; Morgan, D.; Wirth, B. D.; Gussev, M. N.; Busby, J. T.

    2016-02-01

    Life extension of light water reactors will expose austenitic internal core components to irradiation damage levels beyond 100 displacements per atom (dpa), leading to profound microstructural evolution and consequent degradation of macroscopic properties. Microstructural evolution, including Frank loops, cavities, precipitates, and segregation at boundaries and the resultant radiation hardening in type 304 and 316 stainless steel (SS) variants were studied in this work via experimental characterization and multiple simulation methods. Experimental data for up to 40 heats of type 304SS and 316SS variants irradiated in different reactors to 0.6-120 dpa at 275-375°C were generated from this work or collected from literature reports. These experimental data were then combined with models of Frank loop and cavity evolution, computational thermodynamics and precipitation, and ab initio and rate theory integrated radiation-induced segregation models to provide insights into microstructural evolution and degradation at higher doses.

  5. A Physically-Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels1

    SciTech Connect

    Eason, Ernest D.; Odette, George Robert; Nanstad, Randy K; Yamamoto, Takuya

    2013-01-01

    This paper presents a physically-based, empirically calibrated model for estimating irradiation-induced transition temperature shifts in reactor pressure vessel steels, based on a broader database and more complete understanding of embrittlement mechanisms than was available for earlier models. Brief descriptions of the underlying radiation damage mechanisms and the database are included, but the emphasis is on the model and the quality of its fit to U.S. power reactor surveillance data. The model is compared to a random sample of surveillance data that were set aside and not used in fitting and to selected independent data from test reactor irradiations, in both cases showing good ability to predict data that were not used for calibration. The model is a good fit to the surveillance data, with no significant residual error trends for variables included in the model or additional variables that could be included.

  6. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    SciTech Connect

    Tsai, H.; Strain, R.V.; Gomes, I.; Smith, D.L.; Matsui, H.

    1996-10-01

    The ATR-A1 irradiation experiment was a collaborative U.S./Japan effort to study at low temperature the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation started in the Advanced Test Reactor (ATR) on November 30, 1995, and ended as planned on May 5, 1996. Total exposure was 132.9 effective full power days (EFPDs) and estimated neutron damage in the vanadium was 4.7 dpa. The vehicle has been discharged from the ATR core and is scheduled to be disassembled in the next reporting period.

  7. Intergranular stress corrosion cracking susceptibility of neutron-irradiated, thermally sensitized type 304 stainless steel

    SciTech Connect

    Onchi, T.; Hide, K.; Mayuzumi, M.; Hoshiya, T.

    2000-05-01

    Austenitic stainless steels (SS) have been used as core component materials for light water reactors. As reactors age, however, the material tends to suffer from degradation primarily resulting from irradiation-assisted stress corrosion cracking (IASCC) as well as intergranular stress corrosion cracking (IGSCC). Neutron-irradiated, thermally sensitized Type 304 (UNS S30400) stainless steels (SS) were examined by slow strain rate (SSR) stress corrosion cracking (SCC) tests in 290 C water of 0.2 ppm dissolved oxygen concentration (DO) and by SSR tensile tests in 290 C inert gas environment. Neutron fluences ranged from 4 x 10{sup 22} n/m{sup 2} to 3 x 10{sup 25} n/m{sup 2} (energy [E] > 1 MeV). percent intergranular (%IG) cracking, which has been used as an intergranular (IG) cracking susceptibility indicator in the SSR SCC tests, changes anomalously with neutron fluence in spite of the strain-to-failure rate decreasing with an increase of neutron fluence. Apparently, %IG is a misleading indicator for the irradiated, thermally sensitized Type 304 SS and for the irradiated, nonsensitized SS when IG cracking susceptibility is compared at different neutron fluences, test temperatures, DO, and strain rates. These test parameters may affect deformation and fracture behaviors of the irradiated SS during the SSR SCC tests, resulting in changing %IG, which is given by the ratio of the total IG cracking area to the entire fracture surface area. It is suggested that strain-to-IG crack initiation for the irradiated, thermally sensitized SS and for the irradiated, nonsensitized SS is the alternative indicator in the SSR SCC tests. An engineering expedient to determine the IG crack initiation strain is given by a deviating point on superposed stress-strain curves in inert gas and in oxygenated water. The strain-to-IG crack initiation becomes smaller with an increase of neutron fluence and DO. The SSR tensile tests in inert gas are needed to obtain strain-to-IG crack initiation in

  8. MECHANICAL PROPERTIES AND MICROSTRUCTURE IN LOW ACTIVATION MARTENSITIC STEELS F82H AND OPTIMAX AFTER 800 MEV PROTON IRRADIATION

    SciTech Connect

    Y. DAI; ET AL

    1999-10-01

    Low-activation martensitic steels, F82H (mod.) and Optimax-A, have been irradiated with 800-MeV protons up to 5.9 dpa. The tensile properties and microstructure have been studied. The results show that radiation hardening increases continuously with irradiation dose. F82H has lesser irradiation hardening as compared to Optimax-A in the present work and DIN1.4926 from a previous study. The irradiation embrittlement effects are evident in the materials since the uniform elongation is reduced sharply to less than 2%. However, all the irradiated samples ruptured in a ductile-fracture mode. Defect clusters have been observed. The size and the density of defect clusters increase with the irradiation dose. Precipitates are amorphous after irradiation.

  9. Stability of the strengthening nanoprecipitates in reduced activation ferritic steels under Fe2+ ion irradiation

    NASA Astrophysics Data System (ADS)

    Tan, L.; Katoh, Y.; Snead, L. L.

    2014-02-01

    The stability of MX-type precipitates is critical to retain mechanical properties of both reduced activation ferritic-martensitic (RAFM) and conventional FM steels at elevated temperatures. Radiation resistance of TaC, TaN, and VN nanoprecipitates irradiated up to ∼49 dpa at 500 °C using Fe2+ is investigated in this work. Transmission electron microscopy (TEM) utilized in standard and scanning mode (STEM) reveals the non-stoichiometric nature of the nanoprecipitates. Irradiation did not alter their crystalline nature. The radiation resistance of these precipitates, in an order of reduced resistance, is TaC, VN, and TaN. Particle dissolution, growth, and reprecipitation were the modes of irradiation-induced instability. Irradiation also facilitated formation of Fe2W type Laves phase limited to the VN and TaN bearing alloys. This result suggests that nitrogen level should be controlled to a minimal level in alloys to gain greater radiation resistance of the MX-type precipitates at similar temperatures as well as postpone the formation and subsequent coarsening of Laves phase.

  10. In situ and tomographic observations of defect free channel formation in ion irradiated stainless steels.

    PubMed

    Kacher, J; Liu, G S; Robertson, I M

    2012-11-01

    The effects of heavy-ion irradiation on dislocation processes in stainless steels were investigated using in situ irradiation and deformation in the transmission electron microscope as well as post mortem electron tomography to retrieve information on the three-dimensional dislocation state. Irradiation-induced defects were found to pose a strong collective barrier to dislocation motion, leading to dislocation pileups forming in grain interiors and at grain boundaries. The passage of multiple dislocations along the same slip plane removes the irradiation defects and leads to the eventual formation of a defect-free channel. These channels are composed of densely tangled dislocation networks which line the channel-matrix walls as well as residual dislocation debris in the channel interiors. The structures of the dislocation tangles were found to be similar to those encountered in later stages of deformation in unirradiated materials, with the exception that they developed earlier in the deformation process and were confined to the defect free channels. Also, defect free channels were found to widen through both source widening as well as complex cross-slip mechanisms. PMID:22365051

  11. Effect of heavy ion irradiation on microstructural evolution in CF8 cast austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Chen, Wei-Ying; Li, Meimei; Kirk, Marquis A.; Baldo, Peter M.; Lian, Tiangan

    2016-04-01

    The microstructural evolution in ferrite and austenitic in cast austenitic stainless steel (CASS) CF8, as received or thermally aged at 400 °C for 10,000 h, was followed under TEM with in situ irradiation of 1 MeV Kr ions at 300 and 350 °C to a fluence of 1.9 × 1015 ions/cm2 (∼3 dpa) at the IVEM-Tandem Facility. For the unaged CF8, the irradiation-induced dislocation loops appeared at a much lower dose in the austenite than in the ferrite. At the end dose, the austenite formed a well-developed dislocation network microstructure, while the ferrite exhibited an extended dislocation structure as line segments. Compared to the unaged CF8, the aged specimen appeared to have lower rate of damage accumulation. The rate of microstructural evolution under irradiation in the ferrite was significantly lower in the aged specimen than in the unaged. This difference is attributed to the different initial microstructures in the unaged and aged specimens, which implies that thermal aging and irradiation are not independent but interconnected damage processes.

  12. Strain hardening during mechanical twining and dislocation channeling in irradiated 316 stainless steels

    SciTech Connect

    Byun, Thak Sang; Hashimoto, Naoyuki

    2007-01-01

    Localized deformation mechanisms and strain-hardening behaviors in irradiated 316 and 316LN stainless steels were investigated, and a theoretical model was proposed to explain the linear strain-hardening behavior during the localized deformation. After low temperature irradiation to significant doses the deformation microstructure changed from dislocation tangles to channels or to mechanical twins. It was also observed that irradiation hardening straightened gliding dislocations and increased the tendency for forming pileups. Regardless of these microstructural changes, the strain-hardening behavior was relatively insensitive to the irradiation. This dose-independent strain-hardening rate resulted in dose independence of the true stress parameters such as the plastic instability stress and true fracture stress. In the proposed model, the long-range back stress was formulated as a function of the number of pileup dislocations per slip band and the number of slip bands in a grain. The calculation results confirmed the experimental observation that strain-hardening rate was insensitive to the change in deformation mechanism because the long-range back stress hardening became as high as the hardening by tangled dislocations.

  13. Void Swelling and Microstructure of Austenitic Stainless Steels Irradiated in the BOR - 60 Reactor

    SciTech Connect

    Chen, Y.; Yang, Yong; Huang, Yina; Allen, T.; Alexandreanu, B.; Natesan, K.

    2012-11-01

    As nuclear power plants age and neutron fluence increases, detrimental effects resulting from radiation damage have become an increasingly important issue for the operational safety and structural integrity of core internal components. In this study, irradiated specimens of reactor core internal components were characterized by transmission electron microscopy. The specimens had been irradiated to 5.5-45 dpa in the BOR-60 reactor at a dose rate close to 10-6 dpa/s and temperature of about 320°C. No voids were observed in the austenitic stainless steels and nickel alloys at all doses. Despite the possibility that fine voids below the TEM resolution limit may be present, it was clear that void swelling was insignificant in all examined alloys up to 45 dpa. Irradiated microstructures of the studied alloys were dominated by a high density of Frank loops. The mean size and density of the Frank loops varied from one material to another, but saturated with increasing dose above ~10 dpa. While no irradiation-induced precipitations were present below 24.5 dpa, fine precipitates were evident in several alloys at 45 dpa.

  14. Fatigue behavior of irradiated helium-containing ferritic steels for fusion reactor applications*1

    NASA Astrophysics Data System (ADS)

    Grossbeck, M. L.; Vitek, J. M.; Liu, K. C.

    1986-11-01

    The martensitic alloys 12Cr-1MoVW and 9Cr-1MoVNb have been irradiated in the High Flux Isotope Reactor (HFIR) and subsequently tested in fatigue. In order to achieve helium levels characteristic of fusion reactors, the 12Cr-1MoVW was doped with 1 and 2% Ni, resulting in helium levels of 210 and 410 appm at damage levels of 25 dpa. The 9Cr-1MoVNb was irradiated to a damage level of 3 dpa and contained less than 5 appm He. Irradiations were carried out at 55°C and testing at 22°C. No significant changes were found in 9Cr-1MoVNb upon irradiation at this damage level, but effects that could possibly be attributed to helium were found in 12Cr-1MoVW. Levels of 210 and 410 appm He produced cyclic strengthening of 29 and 34% over unirradiated nickel-doped materials, respectively. This cyclic hardening, attributable largely to helium, resulted in degradation of the cyclic life. However, the fatigue life remained comparable to or better than unirradiated 20%-cold-worked 316 stainless steel.

  15. Irradiation-induced sensitization of austenitic stainless steel in-core components

    SciTech Connect

    Chung, H.M.; Sanecki, J.E.; Ruther, W.E.; Kassner, T.F.

    1990-10-01

    High- and commercial-purity specimens of Type 304 SS from BWR absorber rod tubes, irradiated during service to fluence levels of 6 {times} 10{sup 20} to 2 {times} 10{sup 21} n{center dot}cm{sup {minus}2} (E > 1 MeV) in two reactors, were examined by Auger electron spectroscopy to characterize irradiation-induced grain boundary segregation and depletion of alloying and impurity elements, which have been associated with irradiation-assisted stress corrosion cracking (IASCC) of the steel. Ductile and intergranular fracture surfaces were produced by bending of hydrogen-charged specimens in the ultra-high vacuum of Auger microscope. The intergranular fracture surfaces in high-fluence commercial-purity material were characterized by relatively high levels of Si, P, and In segregation. An Auger energy peak at 59 eV indicated either segregation of an unidentified element or formation of an unidentified compound on the grain boundary. In contrast to the commercial-purity material, segregation of the impurity elements and intergranular failure in the high-purity material were negligible for a similar fluence level. However, grain boundary depletion of Cr was more significant in high-purity material than in commercial-purity material, which indicates that irradiation-induced segregation of impurity elements and depletion of alloying elements are interdependent. 7 refs., 10 figs., 2 tabs.

  16. Tensile stress corrosion cracking of type 304 stainless steel irradiated to very high dose

    SciTech Connect

    Chung, H. M.; Ruther, W. E.; Strain, R. V.; Shack, W. J.

    2001-09-01

    Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs), accumulate very high levels of irradiation damage (20--100 displacement per atom or dpa) by the end of life. The data bases and mechanistic understanding of, the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high dose, i.e., is it purely mechanical failure or is it stress-commotion cracking? In this work, hot-cell tests and microstructural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-11 reactor after irradiation to {approximately}50 dpa at {approximately}370 C. Slow-strain-rate tensile tests were conducted at 289 C in air and in water at several levels of electrochemical potential (ECP), and microstructural characteristics were analyzed by scanning and transmission electron microcopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, but was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating that the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at low ECP, and this susceptibility led to poor work-hardening capability and low ductility.

  17. Heavy-Section Steel Irradiation Program: Progress report for April--September 1995. Volume 6, Number 2

    SciTech Connect

    Corwin, W.R.

    1996-08-01

    The goal of the Heavy-Section Steel Irradiation Program is to provide a thorough, quantitative assessment of effects of neutron irradiation on material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and post-irradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into 14 tasks: (1) program management, (2) fracture toughness (K{sub Ic}) curve shift in high-copper welds, (3) crack-arrest toughness (K{sub Ia}) curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K{sub Ic} and K{sub Ia} curve shifts in low upper-shelf welds, (6) annealing effects in low upper-shelf welds, (7) irradiation effects in a commercial low upper-shelf weld, (8) microstructural analysis of irradiation effects, (9) in-service aged material evaluations, (10) correlation monitor materials, (11) special technical assistance, (12) JPDR steel examination, (13) technical assistance for JCCCNRS Working Groups 3 and 12, and (14) additional requirements for materials. This report provides an overview of the activities within each of these task from April through September 1995.

  18. Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

    SciTech Connect

    Chen, Y.; Chopra, O. K.; Soppet, W. K.; Shack, W. J.; Yang, Y.; Allen, T. R.; Univ. of Wisconsin at Madison

    2010-02-16

    Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

  19. Slow positron beam and nanoindentation study of irradiation-related defects in reactor vessel steels

    NASA Astrophysics Data System (ADS)

    Liu, Xiangbing; Wang, Rongshan; Jiang, Jing; Wu, Yichu; Zhang, Chonghong; Ren, Ai; Xu, Chaoliang; Qian, Wangjie

    2014-08-01

    In order to understand the nature of the hardening after radiation in reactor vessel steels, China A508-3 steels were implanted by proton with an energy of 240 keV up to 2.5 × 1016, 5.5 × 1016, 1.1 × 1017, and 2.5 × 1017 ions cm-2, respectively. Vacancy type defects were detected by energy-variable positron beam Doppler broadening technique and then nanoindentation measurements were performed to investigate proton-induced hardening effects. The results showed that S-parameter increased as a function of positron incident energy after irradiation, and the increasing rate of the S-parameter near the surface was larger than that in the bulk due to radiation damage. The size of vacancy type defects increased with dose. Irradiation induced hardening was shown that the average hardness increased with dose. Moreover a direct correlation between positron annihilation parameter and hardness was found based on Kasada method.

  20. True stress-strain curve acquisition for irradiated stainless steel including the range exceeding necking strain

    NASA Astrophysics Data System (ADS)

    Kamaya, Masayuki; Kitsunai, Yuji; Koshiishi, Masato

    2015-10-01

    True stress-strain curves were obtained for irradiated 316L stainless steel by a tensile test and by a curve estimation procedure. In the tensile test, the digital image correlation technique together with iterative finite element analysis was applied in order to identify curves for strain larger than the necking strain. The true stress-strain curves were successfully obtained for the strain of more than 0.4 whereas the necking strain was about 0.2 in the minimum case. The obtained true stress-strain curves were approximated well with the Swift-type equation including the post-necking strain even if the exponential constant n was fixed to 0.5. Then, the true stress-strain curves were estimated by a curve estimation procedure, which was referred to as the K-fit method. Material properties required for the K-fit method were the yield and ultimate strengths or only the yield strength. Some modifications were made for the K-fit method in order to improve estimation accuracy for irradiated stainless steels.

  1. Hydrogen and oxygen trapping and retention in stainless steel and graphite materials irradiated in plasma

    NASA Astrophysics Data System (ADS)

    Begrambekov, L.; Ayrapetov, A.; Ermakov, V.; Kaplevsky, A.; Sadovsky, Ya.; Shigin, P.

    2013-11-01

    The paper presents the results of experimental investigation of energy and flux dependences of hydrogen isotopes and oxygen trapping in carbon materials (carbon fiber composite and pyrolitic graphite), and metals (stainless steel and nickel) under irradiation in the deuterium gas discharge plasma with and without oxygen addition. The dependence of hydrogen trapping on ion energy, ion current density, oxygen addition in deuterium plasma are presented and analyzed. The sorbed molecules, containing hydrogen atoms from the residual gas and deuterium atoms of the working gas are shown to be the important source of hydrogen trapping in both carbon based materials and stainless steel. Irradiation of the SS vacuum vessel with the neutrals or/and ions of (D2 + O2) plasma initiate the hydrogen diffusion from the vessel wall and H2, HD, D2O, HDO, H2O molecule formation on the wall surface. Trapping of the low energy plasma particles and the particles from the sorbed molecules as well as modification of working gas composition are considered as the processes provided at the expense of the potential energy of plasma particles with respect to the surface and occurred through their inelastic collisions with the surface. The hydrogen trapping occurred due to “potential” processes was named as “potential”, and in contrast the trapping of fast particles due to their kinetic energy was labeled as “kinetic”.

  2. Microstructural characterization of deformation localization at small strains in a neutron-irradiated 304 stainless steel

    NASA Astrophysics Data System (ADS)

    Field, Kevin G.; Gussev, Maxim N.; Busby, Jeremy T.

    2014-09-01

    A specific phenomenon - highly localized regions of deformation - was found and investigated at the free surface and near-surface layer of a neutron irradiated AISI 304 stainless steel bend specimen deformed to a maximum surface strain of 0.8%. It was shown that local plastic deformation near the surface might reach significant levels being localized at specific spots even when the maximum free surface strain remains below 1%. The effect was not observed in non-irradiated steel of the same composition at similar strain levels. Cross-sectional EBSD analysis demonstrated that the local misorientation level was highest near the free surface and diminished with increasing depth in these regions. (S)TEM indicated that the local density of dislocation channels might vary up to an order of magnitude. These channels may contain twins or may be twin free depending on grain orientation and local strain levels. BCC-phase (α-martensite) formation associated with channel-grain boundary intersection points was observed using EBSD and STEM in the near-surface layer.

  3. Microstructural evolution of HFIR-irradiated low activation F82H and F82H-{sup 10}B steels

    SciTech Connect

    Wakai, E.; Shiba, K.; Sawai, T.; Hashimoto, N.; Robertson, J.P.; Klueh, R.L.

    1998-03-01

    Microstructures of reduced-activation F82H (8Cr-2W-0.2V-0.04Ta) and the F82H steels doped with {sup 10}B, irradiated at 250 and 300 C to 3 and 57 dpa in the High Flux Isotope Reactor (HFIR), were examined by TEM. In the F82H irradiated at 250 C to 3 dpa, dislocation loops, small unidentified defect clusters with a high number density, and a few MC precipitates were observed in the matrix. The defect microstructure after 300 C irradiation to 57 dpa is dominated by the loops, and the number density of loops was lower than that of the F82H-{sup 10}B steel. Cavities were observed in the F82H-{sup 10}B steels, but the swelling value is insignificant. Small particles of M{sub 6}C formed on the M{sub 23}C{sub 6} carbides that were present in both steels before the irradiation at 300 C to 57 dpa. A low number density of MC precipitate particles formed in the matrix during irradiation at 300 C to 57 dpa.

  4. Heavy-section steel technology and irradiation programs-retrospective and prospective views

    SciTech Connect

    Nanstad, Randy K; Bass, Bennett Richard; Rosseel, Thomas M; Merkle, John Graham; Sokolov, Mikhail A

    2007-01-01

    In 1965, the Atomic Energy Commission (AEC), at the advice of the Advisory Committee on Reactor Safeguards (ACRS), initiated the process that resulted in the establishment of the Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratory (ORNL). Dr. Spencer H. Bush of Battelle Northwest Laboratory, the man being honored by this symposium, representing the ACRS, was one of the Staff Advisors for the program and helped to guide its technical direction. In 1989, the Heavy-Section Steel Irradiation (HSSI) Program, formerly the HSST task on irradiation effects, was formed as a separate program, and this year the HSST/HSSI Programs, sponsored by the U.S. Nuclear Regulatory Commission (USNRC), celebrate 40 years of continuous research oriented toward the safety of light-water nuclear reactor pressure vessels. This paper presents a summary of results from those programs with a view to future activities. The HSST Program was established in 1967 and initially included extensive investigations of heavy-section low-alloy steel plates, forgings, and welds, including metallurgical studies, mechanical properties, fracture toughness (quasi-static and dynamic), fatigue crack-growth, and crack arrest toughness. Also included were irradiation effects studies, thermal shock analyses, testing of thick-section tensile and fracture specimens, and non-destructive testing. In the subsequent decades, the HSST Program conducted extensive large-scale experiments with intermediate-size vessels (with varying size flaws) pressurized to failure, similar experiments under conditions of thermal shock and even pressurized thermal shock (PTS), wide-plate crack arrest tests, and biaxial tests with cruciform-shaped specimens. Extensive analytical and numerical studies accompanied these experiments, including the development of computer codes such as the recent Fracture Analysis of Vessels Oak Ridge (FAVOR) code currently being used for PTS evaluations. In the absence of radiation

  5. Evolution of structure and properties of VVER-1000 RPV steels under accelerated irradiation up to beyond design fluences

    NASA Astrophysics Data System (ADS)

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Maltsev, D.; Frolov, A.; Zabusov, O.; Erak, D.; Zhurko, D.

    2015-01-01

    In this paper comprehensive studies of structure and properties of VVER-1000 RPV steels after the accelerated irradiation to fluences corresponding to extended lifetime up to 60 years or more as well as comparative studies of materials irradiated with different fluxes were carried out. The significant flux effect is confirmed for the weld metal (nickel concentration ⩾1.35%) which is mainly due to development of reversible temper brittleness. The rate of radiation embrittlement of VVER-1000 RPV steels under operation up to 60 years and more (based on the results of accelerated irradiation considering flux effect for weld metal) is expected not to differ significantly from the observed rate under irradiation within surveillance specimens.

  6. Influence of neutron irradiation on mechanical and dimensional stability of irradiated stainless steels, and its possible impact on spent fuel storage

    SciTech Connect

    Garner, Francis A.

    2007-04-27

    Stainless steels used as cladding and structural materials in nuclear reactors undergo very pronounced changes in physical and mechanical properties during irradiation at elevated temperatures, often quickly leading to an increased tendency toward embrittlement. On a somewhat longer time scale there arise very significant changes in component volume and relative dimensions due to void swelling and irradiation creep. Irradiation creep is an inherently undamaging process but once swelling exceeds the 5-10% range austenitic steels become exceptionally brittle. Other processes also contribute to embrittlement and thereby contribute to difficulty in storing and handling of spent fuel assemblies removed from decommissioned fast reactors. In light water reactors other forms of embrittlement develop prior to reaching significant levels of void swelling. A review is presented of our current understanding of the radiation-induced changes in physical and mechanical properties that contgribute to embrittlement.

  7. Low cycle fatigue properties of reduced activation ferritic/martensitic steels after high-dose neutron irradiation

    NASA Astrophysics Data System (ADS)

    Gaganidze, E.; Petersen, C.; Aktaa, J.; Povstyanko, A.; Prokhorov, V.; Diegele, E.; Lässer, R.

    2011-08-01

    This paper focuses on the low cycle fatigue (LCF) behaviour of reduced activation ferritic/martensitic steels irradiated to a displacement damage dose of up to 70 dpa at 330-337 °C in the BOR 60 reactor within the ARBOR 2 irradiation programme. The influence of neutron irradiation on the fatigue behaviour was determined for the as-received EUROFER97, pre-irradiation heat-treated EUROFER97 HT and F82H-mod steels. Strain-controlled push-pull loading was performed using miniaturized cylindrical specimens at a constant temperature of 330 °C with total strain ranges between 0.8% and 1.1%. Comparison of the LCF behaviour of irradiated and reference unirradiated specimens was performed for both the adequate total and inelastic strains. Neutron irradiation-induced hardening may have various effects on the fatigue behaviour of the steels. The reduction of inelastic strain in the irradiated state compared with the reference unirradiated state at common total strain amplitudes may increase fatigue lifetime. The increase in the stress at the adequate inelastic strain, by contrast, may accelerate fatigue damage accumulation. Depending on which of the two effects mentioned dominates, neutron irradiation may either extend or reduce the fatigue lifetime compared with the reference unirradiated state. The results obtained for EUROFER97 and EUROFER97 HT confirm these considerations. Most of the irradiated specimens show fatigue lifetimes comparable to those of the reference unirradiated state at adequate inelastic strains. Some irradiated specimens, however, show lifetime reduction or increase in comparison with the reference state at adequate inelastic strains.

  8. Manufacturing and STA-investigation of witness-samples for the temperature monitoring of structural steels under irradiation

    NASA Astrophysics Data System (ADS)

    Sevryukov, O. N.; Fedotov, V. T.; Polyansky, A. A.; Pokrovski, S. A.; Kuzmin, R. S.

    2016-04-01

    The object of investigations was alloys based on lead and cadmium used as fuse monitors to control the maximum irradiation temperature (fuse temperature monitors, FTM) of samples from structural steels under irradiation in a research reactor IR-8. The result of the work was selected and tested initial materials for production of alloys. A technological scheme of the production of alloys for FTM has been developed and experimental studies of the properties of these alloys have been carried out.

  9. Structural Transformations in Austenitic Stainless Steel Induced by Deuterium Implantation: Irradiation at 295 K

    NASA Astrophysics Data System (ADS)

    Morozov, Oleksandr; Zhurba, Volodymir; Neklyudov, Ivan; Mats, Oleksandr; Progolaieva, Viktoria; Boshko, Valerian

    2016-02-01

    Deuterium thermal desorption spectra were investigated on the samples of austenitic steel 18Cr10NiTi pre-implanted at 295 K with deuterium ions in the dose range from 8 × 1014 to 2.7 × 1018 D/cm2. The kinetics of structural transformation development in the steel layer was traced from deuterium thermodesorption spectra as a function of deuterium concentration. Three characteristic regions with different low rates of deuterium amount desorption as the implantation dose increases were revealed: I—the linear region of low implantation doses (up to 1 × 1017 D/cm2); II—the nonlinear region of medium implantation doses (1 × 1017 to 8 × 1017 D/cm2); III—the linear region of high implantation doses (8 × 1017 to 2.7 × 1018 D/cm2). During the process of deuterium ion irradiation, the coefficient of deuterium retention in steel varies in discrete steps. Each of the discrete regions of deuterium retention coefficient variation corresponds to different implanted-matter states formed during deuterium ion implantation. The low-dose region is characterized by formation of deuterium-vacancy complexes and solid-solution phase state of deuterium in the steel. The total concentration of the accumulated deuterium in this region varies between 2.5 and 3 at.%. The medium-dose region is characterized by the radiation-induced action on the steel in the presence of deuterium with the resulting formation of the energy-stable nanosized crystalline structure of steel, having a developed network of intercrystalline boundaries. The basis for this developed network of intercrystalline boundaries is provided by the amorphous state, which manifests itself in the thermodesorption spectra as a widely temperature-scale extended region of deuterium desorption (structure formation with a varying activation energy). The total concentration of the accumulated deuterium in the region of medium implantation doses makes 7 to 8 at.%. The resulting structure shows stability against the action of

  10. Structural Transformations in Austenitic Stainless Steel Induced by Deuterium Implantation: Irradiation at 295 K.

    PubMed

    Morozov, Oleksandr; Zhurba, Volodymir; Neklyudov, Ivan; Mats, Oleksandr; Progolaieva, Viktoria; Boshko, Valerian

    2016-12-01

    Deuterium thermal desorption spectra were investigated on the samples of austenitic steel 18Cr10NiTi pre-implanted at 295 K with deuterium ions in the dose range from 8 × 10(14) to 2.7 × 10(18) D/cm(2). The kinetics of structural transformation development in the steel layer was traced from deuterium thermodesorption spectra as a function of deuterium concentration. Three characteristic regions with different low rates of deuterium amount desorption as the implantation dose increases were revealed: I-the linear region of low implantation doses (up to 1 × 10(17) D/cm(2)); II-the nonlinear region of medium implantation doses (1 × 10(17) to 8 × 10(17) D/cm(2)); III-the linear region of high implantation doses (8 × 10(17) to 2.7 × 10(18) D/cm(2)). During the process of deuterium ion irradiation, the coefficient of deuterium retention in steel varies in discrete steps. Each of the discrete regions of deuterium retention coefficient variation corresponds to different implanted-matter states formed during deuterium ion implantation. The low-dose region is characterized by formation of deuterium-vacancy complexes and solid-solution phase state of deuterium in the steel. The total concentration of the accumulated deuterium in this region varies between 2.5 and 3 at.%. The medium-dose region is characterized by the radiation-induced action on the steel in the presence of deuterium with the resulting formation of the energy-stable nanosized crystalline structure of steel, having a developed network of intercrystalline boundaries. The basis for this developed network of intercrystalline boundaries is provided by the amorphous state, which manifests itself in the thermodesorption spectra as a widely temperature-scale extended region of deuterium desorption (structure formation with a varying activation energy). The total concentration of the accumulated deuterium in the region of medium implantation doses makes 7 to 8 at.%. The

  11. Meso-Scale Magnetic Signatures for Nuclear Reactor Steel Irradiation Embrittlement Monitoring

    SciTech Connect

    Suter, Jonathan D.; Ramuhalli, Pradeep; McCloy, John S.; Xu, Ke; Hu, Shenyang Y.; Li, Yulan; Jiang, Weilin; Edwards, Danny J.; Schemer-Kohrn, Alan L.; Johnson, Bradley R.

    2015-03-31

    Verifying the structural integrity of passive components in light-water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the ‘state of health’ of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of non-destructive evaluation (NDE) technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results to integrate advanced material characterization techniques with meso-scale computational models to provide an interpretive understanding of the state of degradation in a material. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. In future efforts, microstructural measurements and meso-scale magnetic measurements on thin films will be used to gain insights into the structural state of these materials to study the impact of irradiation on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

  12. Meso-scale magnetic signatures for nuclear reactor steel irradiation embrittlement monitoring

    SciTech Connect

    Suter, J. D. Ramuhalli, P. Hu, S.; Li, Y.; Jiang, W.; Edwards, D. J.; Schemer-Kohrn, A. L.; Johnson, B. R.; McCloy, J. S. Xu, K.

    2015-03-31

    Verifying the structural integrity of passive components in light water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the “state of health” of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of nondestructive evaluation technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results of integrating advanced material characterization techniques with meso-scale computational models. In the future, this will help to provide an interpretive understanding of the state of degradation in structural materials. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. Ongoing research is focused on extending the measurements and models on thin films to gain insights into the structural state of irradiated materials and the resulting impact on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

  13. Meso-scale magnetic signatures for nuclear reactor steel irradiation embrittlement monitoring

    NASA Astrophysics Data System (ADS)

    Suter, J. D.; Ramuhalli, P.; McCloy, J. S.; Xu, K.; Hu, S.; Li, Y.; Jiang, W.; Edwards, D. J.; Schemer-Kohrn, A. L.; Johnson, B. R.

    2015-03-01

    Verifying the structural integrity of passive components in light water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the "state of health" of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of nondestructive evaluation technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results of integrating advanced material characterization techniques with meso-scale computational models. In the future, this will help to provide an interpretive understanding of the state of degradation in structural materials. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. Ongoing research is focused on extending the measurements and models on thin films to gain insights into the structural state of irradiated materials and the resulting impact on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

  14. Warm PreStress effect on highly irradiated reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Hure, J.; Vaille, C.; Wident, P.; Moinereau, D.; Landron, C.; Chapuliot, S.; Benhamou, C.; Tanguy, B.

    2015-09-01

    This study investigates the Warm Prestress (WPS) effect on 16MND5 (A508 Cl3) RPV steel, irradiated up to a fluence of 13 ·1023 n .m-2 (E > 1 MeV) at a temperature of 288 ° C, corresponding to more than 60 years of operations in a French Pressurized Water Reactor (PWR). Mechanical properties, including tensile tests with different strain rates and tension-compression tests on notched specimens, have been characterized at unirradiated and irradiated states and used to calibrate constitutive equations to describe the mechanical behavior as a function of temperature and fluence. Irradiation embrittlement has been determined based on Charpy V-notch impact tests and isothermal quasi-static toughness tests. Assessment of WPS effect has been done through various types of thermomechanical loadings performed on CT(0.5 T) specimens. All tests have confirmed the non-failure during the thermo-mechanical transients. Experimental data obtained in this study have been compared to both engineering-based models and to a local approach (Beremin) model for cleavage fracture. It is shown that both types of modeling give good predictions for the effective toughness after warm prestressing.

  15. Microstructural evolution of ferritic-martensitic steels under heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Topbasi, Cem

    Ferritic-martensitic steels are primary candidate materials for fuel cladding and internal applications in the Sodium Fast Reactor, as well as first-wall and blanket materials in future fusion concepts because of their favorable mechanical properties and resistance to radiation damage. Since microstructure evolution under irradiation is amongst the key issues for these materials in these applications, developing a fundamental understanding of the irradiation-induced microstructure in these alloys is crucial in modeling and designing new alloys with improved properties. The goal of this project was to investigate the evolution of microstructure of two commercial ferritic-martensitic steels, NF616 and HCM12A, under heavy ion irradiation at a broad temperature range. An in situ heavy ion irradiation technique was used to create irradiation damage in the alloy; while it was being examined in a transmission electron microscope. Electron-transparent samples of NF616 and HCM12A were irradiated in situ at the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory with 1 MeV Kr ions to ˜10 dpa at temperatures ranging from 20 to 773 K. The microstructure evolution of NF616 and HCM12A was followed in situ by systematically recording micrographs and diffraction patterns as well as capturing videos during irradiation. In these irradiations, there was a period during which no changes are visible in the microstructure. After a threshold dose (˜0.1 dpa between 20 and 573 K, and ˜2.5 dpa at 673 K) black dots started to become visible under the ion beam. These black dots appeared suddenly (from one frame to the next) and are thought to be small defect clusters (2-5 nm in diameter), possibly small dislocation loops with Burgers vectors of either ½ or . The overall density of these defect clusters increased with dose and saturated around 6 dpa. At saturation, a steady-state is reached in which defects are eliminated and created at the same rates so that the

  16. Microstructural evolution of ferritic-martensitic steels under heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Topbasi, Cem

    Ferritic-martensitic steels are primary candidate materials for fuel cladding and internal applications in the Sodium Fast Reactor, as well as first-wall and blanket materials in future fusion concepts because of their favorable mechanical properties and resistance to radiation damage. Since microstructure evolution under irradiation is amongst the key issues for these materials in these applications, developing a fundamental understanding of the irradiation-induced microstructure in these alloys is crucial in modeling and designing new alloys with improved properties. The goal of this project was to investigate the evolution of microstructure of two commercial ferritic-martensitic steels, NF616 and HCM12A, under heavy ion irradiation at a broad temperature range. An in situ heavy ion irradiation technique was used to create irradiation damage in the alloy; while it was being examined in a transmission electron microscope. Electron-transparent samples of NF616 and HCM12A were irradiated in situ at the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory with 1 MeV Kr ions to ˜10 dpa at temperatures ranging from 20 to 773 K. The microstructure evolution of NF616 and HCM12A was followed in situ by systematically recording micrographs and diffraction patterns as well as capturing videos during irradiation. In these irradiations, there was a period during which no changes are visible in the microstructure. After a threshold dose (˜0.1 dpa between 20 and 573 K, and ˜2.5 dpa at 673 K) black dots started to become visible under the ion beam. These black dots appeared suddenly (from one frame to the next) and are thought to be small defect clusters (2-5 nm in diameter), possibly small dislocation loops with Burgers vectors of either ½ or . The overall density of these defect clusters increased with dose and saturated around 6 dpa. At saturation, a steady-state is reached in which defects are eliminated and created at the same rates so that the

  17. Results of crack-arrest tests on irradiated a 508 class 3 steel

    SciTech Connect

    Iskander, S.K.; Milella, P.P.; Pini, M.A.

    1998-02-01

    Ten crack-arrest toughness values for irradiated specimens of A 508 class 3 forging steel have been obtained. The tests were performed according to the American Society for Testing and Materials (ASTM) Standard Test Method for Determining Plane-Strain Crack-Arrest Fracture Toughness, K{sub la} of Ferritic Steels, E 1221-88. None of these values are strictly valid in all five ASTM E 1221-88 validity criteria. However, they are useful when compared to unirradiated crack-arrest specimen toughness values since they show the small (averaging approximately 10{degrees}C) shifts in the mean and lower-bound crack-arrest toughness curves. This confirms that a low copper content in ASTM A 508 class 3 forging material can be expected to result in small shifts of the transition toughness curve. The shifts due to neutron irradiation of the lower bound and mean toughness curves are approximately the same as the Charpy V-notch (CVN) 41-J temperature shift. The nine crack-arrest specimens were irradiated at temperatures varying from 243 to 280{degrees}C, and to a fluence varying from 1.7 to 2.7 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV). The test results were normalized to reference values that correspond to those of CVN specimens irradiated at 284{degrees}C to a fluence of 3.2 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) in the same capsule as the crack-arrest specimens. This adjustment resulted in a shift to lower temperatures of all the data, and in particular moved two data points that appeared to lie close to or lower than the American Society of Mechanical Engineers K{sub la} curve to positions that seemed more reasonable with respect to the remaining data. A special fixture was designed, fabricated, and successfully used in the testing. For reasons explained in the text, special blocks to receive the Oak Ridge National Laboratory clip gage were designed, and greater-than-standard crack-mouth opening displacements measured were accounted for. 24 refs., 13 figs., 12 tabs.

  18. Irradiation creep of SA 304L and CW 316 stainless steels: Mechanical behaviour and microstructural aspects. Part I: Experimental results

    NASA Astrophysics Data System (ADS)

    Garnier, J.; Bréchet, Y.; Delnondedieu, M.; Pokor, C.; Dubuisson, P.; Renault, A.; Averty, X.; Massoud, J. P.

    2011-06-01

    Solution annealed 304L (SA 304L) and cold work 316 (CW 316) austenitic stainless steel irradiation creep behaviour have been studied thoroughly. Irradiations were carried out in fast breeder reactors BOR-60 (at 330 °C, up to 120 dpa) and EBR-II (at 375 °C, up to 10.5 dpa), and in the OSIRIS mixed spectrum reactor (at 330 °C, up to 9.8 dpa). After an incubation threshold, the irradiation creep of the austenitic stainless steels is linear in stress and in dose. Creep appears to be athermal in this temperature range. A significant difference in the behaviour is measured between the creep of SA 304L and CW 316. In order to study the anisotropy of loop population, which would be the signature of a possible stress induced preferential absorption (SIPA) mechanism for irradiation creep, special attention was given to the measurement of anisotropy of loop distribution between the four families. The anisotropy induced by an applied stress has been shown to be in the range of the statistical scatter in the situation where no stress is applied. TEM microstructural analyses performed on this sample show slight difference between the microstructure of specimens deformed under irradiation and the microstructure of specimens irradiated without stress under the same irradiation conditions.

  19. Microstructural characterization of irradiated PWR steels using the atom probe field-ion microscope

    SciTech Connect

    Miller, M.K.; Burke, M.G.

    1987-08-01

    Atom probe field-ion microscopy has been used to characterize the microstructure of a neutron-irradiated A533B pressure vessel steel weld. The atomic spatial resolution of this technique permits a complete structural and chemical description of the ultra-fine features that control the mechanical properties to be made. A variety of fine scale features including roughly spherical copper precipitates and clusters, spherical and rod-shaped molybdenum carbide and disc-shaped molybdenum nitride precipitates were observed to be inhomogeneously distributed in the ferrite. The copper content of the ferrite was substantially reduced from the nominal level. A thin film of molybdenum carbides and nitrides was observed on grain boundaries in addition to a coarse copper-manganese precipitate. Substantial enrichment of manganese and nickel were detected at the copper-manganese precipitate-ferrite interface and this enrichment extended into the ferrite. Enrichment of nickel, manganese and phosphorus were also measured at grain boundaries.

  20. Effects of low temperature neutron irradiation on deformation behavior of austenitic stainless steels

    SciTech Connect

    Pawel, J.E.; Rowcliffe, A.F.; Alexander, D.J.; Grossbeck, M.L.; Shiba, K.

    1996-04-01

    An austenitic stainless steel, designated 316LN-IG, has been chosen for the first wall/shield (FW/S) structure for the International Thermonuclear Experimental Reactor (ITER). The proposed operational temperature range for the structure (100 to 250{degree}C) is below the temperature regimes for void swelling (400-600{degree}C) and for helium embrittlement (500-700{degree}C). However, the proposed neutron dose is such that large changes in yield strength, deformation mode, and strain hardening capacity could be encountered which could significantly affect fracture properties. Definition of the irradiation regimes in which this phenomenon occurs is essential to the establishment of design rules to protect against various modes of failure.

  1. Composite model of microstructural evolution in austenitic stainless steel under fast neutron irradiation

    SciTech Connect

    Stoller, R.E.; Odette, G.R.

    1986-01-01

    A rate-theory-based model has been developed which includes the simultaneous evolution of the dislocation and cavity components of the microstructure of irradiated austenitic stainless steels. Previous work has generally focused on developing models for void swelling while neglecting the time dependence of the dislocation structure. These models have broadened our understanding of the physical processes that give rise to swelling, e.g., the role of helium and void formation from critically-sized bubbles. That work has also demonstrated some predictive capability by successful calibration to fit the results of fast reactor swelling data. However, considerable uncertainty about the values of key parameters in these models limits their usefulness as predictive tools. Hence the use of such models to extrapolate fission reactor swelling data to fusion reactor conditions is compromised.

  2. Charpy toughness and tensile properties of a neutron irradiated stainless steel submerged-arc weld cladding overlay

    SciTech Connect

    Corwin, W.R.; Berggren, R.G.; Nanstad, R.K.

    1984-01-01

    The possibility of stainless steel cladding increasing the resistance of an operating nuclear reactor pressure vessel to extension of surface flaws is highly dependent upon the irradiated properties of the cladding. Therefore, weld overlay cladding irradiated at temperatures and fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the submerged-arc, single-wire, oscillating electrode method. Three layers of cladding were applied to provide a cladding thickness adequate for fabrication of test specimens. The first layer was type 309, and the upper two layers were type 308 stainless steel. There was considerable dilution of the type 309 in the first layer of cladding as a result of excessive melting of the base plate. Specimens for the irradiation study were taken from near the base plate/cladding interface and also from the upper layers of cladding. Charpy V-notch and tensile specimens were irradiated at 288/sup 0/C to neutron fluences of 2 x 10/sup 23/ n/m/sup 2/ (E > 1 MeV). When irradiated, both types 308 and 309 cladding showed a 5 to 40% increase in yield strength accompanied by a slight increase in ductility in the temperature range from 25 to 288/sup 0/C. All cladding exhibited ductile-to-brittle transition behavior during impact testing.

  3. Parametric study of irradiation effects on the ductile damage and flow stress behavior in ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Chakraborty, Pritam; Biner, S. Bulent

    2015-10-01

    Ferritic-martensitic steels are currently being considered as structural materials in fusion and Gen-IV nuclear reactors. These materials are expected to experience high dose radiation, which can increase their ductile to brittle transition temperature and susceptibility to failure during operation. Hence, to estimate the safe operational life of the reactors, precise evaluation of the ductile to brittle transition temperatures of ferritic-martensitic steels is necessary. Owing to the scarcity of irradiated samples, particularly at high dose levels, micro-mechanistic models are being employed to predict the shifts in the ductile to brittle transition temperatures. These models consider the ductile damage evolution, in the form of nucleation, growth and coalescence of voids; and the brittle fracture, in the form of probabilistic cleavage initiation, to estimate the influence of irradiation on the ductile to brittle transition temperature. However, the assessment of irradiation dependent material parameters is challenging and influences the accuracy of these models. In the present study, the effects of irradiation on the overall flow stress and ductile damage behavior of two ferritic-martensitic steels is parametrically investigated. The results indicate that the ductile damage model parameters are mostly insensitive to irradiation levels at higher dose levels though the resulting flow stress behavior varies significantly.

  4. Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

    NASA Astrophysics Data System (ADS)

    Byun, Thak Sang; Toloczko, Mychailo B.; Saleh, Tarik A.; Maloy, Stuart A.

    2013-01-01

    To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3-148 dpa at 378-504 °C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 °C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa √m occurred in room temperature tests when irradiation temperature was below 400 °C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa √m was measured when the irradiation temperature was above 430 °C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3-148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 °C) irradiation cases, which indicates that the ductile-brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.

  5. Effect of ITER components manufacturing cycle on the irradiation behaviour of 316L(N)-IG steel

    NASA Astrophysics Data System (ADS)

    Rodchenkov, B. S.; Prokhorov, V. I.; Makarov, O. Yu; Shamardin, V. K.; Kalinin, G. M.; Strebkov, Yu. S.; Golosov, O. A.

    2000-12-01

    The main options for the manufacturing of high heat flux (HHF) components is hot isostatic pressing (HIP) using either solid pieces or powder. There was no database on the radiation behaviour of these materials, and in particular stainless steel (SS) 316L(N)-IG with ITER components manufacturing thermal cycle. Irradiation of wrought steel, powder-HIP, solid-HIP and HIPed joints has been performed within the framework of an ITER task. Specimens cut from 316L(N)-IG plate, HIP products, and solid-HIP joints were irradiated in the SM-3 reactor in Dimitrovgrad up to 4 and 10 dpa at 175°C and 265°C. The paper describes the results of post-irradiation tensile and fracture toughness tests.

  6. Tensile, low cycle fatigue and fracture toughness behaviour of type 316L steel irradiated to 0.3 dpa

    NASA Astrophysics Data System (ADS)

    Josefsson, Bertil; Bergenlid, Ulf

    1994-09-01

    The effect of a low dose neutron irradiation on the tensile, low cycle fatigue and fracture toughness properties of type 316L steel plate and weld material was investigated. The specimens were irradiated at a temperature of about 35°C to a neutron fluence of approximately 2.5 × 10 20 n/cm 2 ( E > 1 MeV). The testing was performed at 75, 250 and 450°C. Irradiated tensile specimens showed a substantial radiation hardening combined with some reduction of elongations. There was no significant effect of the irradiation on the low cycle fatigue endurances. The fracture toughness of the TIG weld specimens was roughly half of that of the 316L plate and electron beam weld. Some reductions of toughness owing to the irradiation were observed.

  7. Microstructure and Nano-Hardness of 10 MeV Cl-Ion Irradiated T91 Steel

    NASA Astrophysics Data System (ADS)

    Hu, Jing; Wang, Xianping; Gao, Yunxia; Zhuang, Zhong; Zhang, Tao; Fang, Qianfeng; Liu, Changsong

    2015-12-01

    Hardening and elemental segregation of T91 martenstic steel irradiated by 10 MeV Cl ions to doses from 0.06 dpa to 0.83 dpa were investigated with the nanoindentation technique and transmission electron microscopy (TEM). The results demonstrated that the irradiation hardening was closely related with irradiation dose. By increasing the dose, the hardness increased rapidly at first from the initial value of 3.15 GPa before irradiation, and then tended to saturate at a value of 3.58 GPa at the highest dose of 0.83 dpa. Combined with TEM observation, the mechanism of hardening was preliminary attributed to the formation of M(Fe,Cr)23C6 carbides induced by the high energy Cl-ion irradiation. supported by National Natural Science Foundation of China (Nos. 11374299, 11375230, 11274309)

  8. Porous microstructures induced by picosecond laser scanning irradiation on stainless steel surface

    NASA Astrophysics Data System (ADS)

    Liu, Bin; Jiang, Gedong; Wang, Wenjun; Mei, Xuesong; Wang, Kedian; Cui, Jianlei; Wang, Jiuhong

    2016-03-01

    A study of porous surfaces having micropores significantly smaller than laser spot on the stainless steel 304L sample surface induced by a picosecond regenerative amplified laser, operating at 1064 nm, is presented. Variations in the interaction regime of picosecond laser pulses with stainless steel surfaces at peak irradiation fluences(Fpk=0.378-4.496 J/cm2) with scanning speeds(v=125-1000 μm/s) and scan line spacings(s=0-50 μm) have been observed and thoroughly investigated. It is observed that interactions within these parameters allows for the generation of well-defined structured surfaces. To investigate the formation mechanism of sub-focus micropores, the influence of key processing parameters has been analyzed using a pre-designed laser pulse scanning layout. Appearances of sub-focus ripples and micropores with the variation of laser peak fluence, scanning speed and scan line spacing have been observed. The dependencies of surface structures on these interaction parameters have been preliminarily verified. With the help of the experimental results obtained, interaction parameters for fabrication of large area homogeneous porous structures with the feature sizes in the range of 3-15 μm are determined.

  9. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    SciTech Connect

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; Ghose, S.; Wells, P.; Stan, T.; Almirall, N.; Odette, G. R.; Ecker, L. E.

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitates that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.

  10. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    DOE PAGESBeta

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; Ghose, S.; Wells, P.; Stan, T.; Almirall, N.; Odette, G. R.; Ecker, L. E.

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitatesmore » that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.« less

  11. Irradiation creep and microstructural changes of ODS steels of different Cr-contents during helium implantation under stress

    NASA Astrophysics Data System (ADS)

    Chen, J.; Jung, P.; Henry, J.; de Carlan, Y.; Sauvage, T.; Duval, F.; Barthe, M. F.; Hoffelner, W.

    2013-06-01

    Irradiation creep and microstructural changes of two ferritic ODS steels with 12% and 14% Cr have been studied by homogeneously implantation with helium under uniaxial tensile stresses from 40 to 300 MPa. The maximum dose was about 1.2 dpa (5000 appm-He) with displacement damage rates of 1 × 10-5 dpa/s at a temperature of 300 °C. Irradiation creep compliances were measured to be 4.0 × 10-6 dpa-1 MPa-1 and 10 × 10-6 dpa-1 MPa-1 for 12 and 14Cr ODS, respectively. Subsequently, microstructural evolution was studied in detail by TEM observations, showing dislocation loops and bubbles distributed homogenously in the matrix. Some bubbles were attached to ODS particles. Finally, the effects of Cr content on irradiation creep and microstructural changes are discussed, including earlier results of a 19Cr ODS and a PM2000 ferritic steel. Irradiation creep rates of both 12Cr and 14Cr-ODS ferritic steels a temperature of 300 °C show linear stress dependence up to 300 MPa at. Irradiation creep rate per dose rate and stress at a temperature of 300 °C amounts to 4.0 × 10-6 dpa-1 MPa-1 and 10 × 10-6 dpa-1 MPa-1 for 12Cr- and 14Cr-ODS, respectively. Irradiation creep properties are remarkably insensitive to Cr content, grain size and dispersoid size. Dislocation loops and helium bubbles are distributed homogenously in the matrix. In the case of high density fine dispersoids, most bubbles are attached to ODS particles. This may suppress loop formation as well as growth of bubbles, thereby increasing the resistance of ODS ferritic steels against helium embrittlement.

  12. On the formation of stacking fault tetrahedra in irradiated austenitic stainless steels - A literature review

    NASA Astrophysics Data System (ADS)

    Schibli, Raluca; Schäublin, Robin

    2013-11-01

    Irradiated austenitic stainless steels, because of their low stacking fault energy and high shear modulus, should exhibit a high ratio of stacking fault tetrahedra relative to the overall population of radiation induced nanometric defects. Experimental observations of stacking fault tetrahedra by transmission electron microscopy in commercial-purity stainless steels are however scarce, while they abundantly occur in high-purity or model austenitic alloys irradiated at both low and high temperatures, but not at around 673 K. In commercial alloys, the little evidence of stacking fault tetrahedra does not follow such a trend. These contradictions are reviewed and discussed. Reviewing the three possible formation mechanisms identified in the literature, namely the Silcox and Hirsch Frank loop dissociation, the void collapse and the stacking fault tetrahedra growth, it seems that the later dominates under irradiation. Black dots, are very small defect clusters, smaller than 1 nm in diameter, which cannot be resolved in TEM being below its spatial resolution in diffraction contrast. They can be created directly from the collapse of the cascade as undefined 3D clusters of point defects, namely vacancies, interstitials or impurities, or could be already well-defined nanometric voids, vacancy or interstitial dislocation loops [7]. Dislocation loops, either Frank or perfect dislocation loops, are generated by vacancies or interstitials coalescing as platelets between two adjacent {1 1 1} close-packed planes. Perfect loops are scarcer than Frank loops. For irradiation temperatures below 573 K some authors identified that Frank loops are of interstitial nature, while black dots are predominantly of vacancy nature [8-11]. More recent studies [12] contradict this statement and conclude that Frank loops with sizes in the range of 1-30 nm can be either vacancy or interstitial type. Stacking fault tetrahedra (SFT) are three-dimensional stacking fault configurations in the shape of

  13. Re-weldability of neutron-irradiated stainless steels studied by multi-pass TIG welding

    NASA Astrophysics Data System (ADS)

    Nakata, K.; Oishi, M.; Koshiishi, M.; Hashimoto, T.; Anzai, H.; Saito, Y.; Kono, W.

    2002-12-01

    Weldability of neutron-irradiated stainless steel (SS) has been studied by multi-pass bead-on-plate and build-up tungsten inert gas (TIG) welding, simulating the repair-welding of reactor components. Specimens were submerged arc welding (SAW) joint of Type 304 SS containing 0.5 appm helium (1.8 appm in the SAW weld metal). Sound welding could be obtained by one- to three-pass welding on the plates at weld heat inputs less than 1 MJ/m in the irradiated 304 SS base metal. In the case of the build-up welding of a groove, no visible defects appeared in the specimen at a heat input as low as 0.4 MJ/m. However, build-up welding at a high heat input of 1 MJ/m was prone to weld cracking, owing to the formation of helium bubbles on grain boundaries of the base metal or dendrite boundaries of pre-existing SAW weld metal, in the area within 0.6 mm from the fusion line.

  14. Microstructural evolution of type 304 and 316 stainless steels under neutron irradiation at LWR relevant conditions

    DOE PAGESBeta

    Tan, Lizhen; Stoller, Roger E.; Field, Kevin G.; Yang, Ying; Morgan, Dane; Wirth, Brian D.; Gussev, Maxim N.; Busby, Jeremy T.; Nam, H.

    2015-12-11

    Extension of light water reactors' useful life will expose austenitic internal core components to irradiation damage levels beyond 100 displacements per atom (dpa), which will lead to profound microstructural evolution and consequent degradation of macroscopic properties. Microstructural evolution, including Frank loops, cavities, precipitates, and segregation at boundaries and the resultant radiation hardening in type 304 and 316 stainless steel (SS) variants, were studied in this work via experimental characterization and multiple simulation methods. Experimental data for up to 40 heats of type 304SS and 316SS variants irradiated in different reactors to 0.6–120 dpa at 275–375°C were either generated from thismore » work or collected from literature reports. These experimental data were then combined with models of Frank loop and cavity evolution, computational thermodynamics and precipitation, and ab initio and rate theory integrated radiation-induced segregation models to provide insights into microstructural evolution and degradation at higher radiation doses.« less

  15. Tritium permeation in EUROFER97 steel in EXOTIC-9/1 irradiation experiment

    NASA Astrophysics Data System (ADS)

    Fedorov, A. V.; Magielsen, A. J.; Stijkel, M. P.

    2014-05-01

    This paper presents the results of the tritium permeation study in EUROFER97 carried out within the EXOTIC (EXtraction Of Tritium In Ceramics) irradiation experiment. In the EXOTIC 9/1 experiment, a pebble bed assembly containing Lithium Titanate (Li2TiO3) pebbles is irradiated for 300 days in the High Flux Reactor (HFR), in the temperature range between 340 and 580 °C, reaching a lithium burn up of 3.5% and 1.2 dpa of damage in steel. The primary objective of this experiment was to measure the in-pile tritium release characteristics of Li2TiO3 pebbles. Additionally tritium permeation through the EUROFER97 pebble bed wall was measured on line. The permeation of tritium was studied at steady state conditions, during temperature transients, and at different hydrogen concentrations in the helium purge gas flow. The model used in the analysis of the experimental data which account for co-permeation of tritium and hydrogen is presented. It has been demonstrated that the permeation of tritium under experiment conditions proceeds in the diffusion limited regime. From the analysis of the experimental data the permeability and diffusivity of tritium in EUROFER97 is determined.

  16. Characterization of irradiated AISI 316L stainless steel disks removed from the Spallation Neutron Source

    SciTech Connect

    Vevera, Bradley J; Hyres, James W; McClintock, David A; Riemer, Bernie

    2014-01-01

    Irradiated AISI 316L stainless steel disks were removed from the Spallation Neutron Source (SNS) for post-irradiation examination (PIE) to assess mechanical property changes due to radiation damage and erosion of the target vessel. Topics reviewed include high-resolution photography of the disk specimens, cleaning to remove mercury (Hg) residue and surface oxides, profile mapping of cavitation pits using high frequency ultrasonic testing (UT), high-resolution surface replication, and machining of test specimens using wire electrical discharge machining (EDM), tensile testing, Rockwell Superficial hardness testing, Vickers microhardness testing, scanning electron microscopy (SEM), and energy dispersive spectroscopy (EDS). The effectiveness of the cleaning procedure was evident in the pre- and post-cleaning photography and permitted accurate placement of the test specimens on the disks. Due to the limited amount of material available and the unique geometry of the disks, machine fixturing and test specimen design were critical aspects of this work. Multiple designs were considered and refined during mock-up test runs on unirradiated disks. The techniques used to successfully machine and test the various specimens will be presented along with a summary of important findings from the laboratory examinations.

  17. Evolution of magnetic properties of cladding austenitic steel under irradiation in a reactor

    NASA Astrophysics Data System (ADS)

    Chukalkin, Yu. G.; Kozlov, A. V.; Evseev, M. V.

    2014-03-01

    Magnetic properties of samples of austenitic steel ChS-68 cut from the cladding of a fuel element, which was irradiated in a BN-600 fast-neutron reactor to a maximal damage dose of ˜80 displacements per atom (dpa) at temperatures of 370-587°C, have been investigated. It has been established that irradiation with fast neutrons leads to the formation of ferromagnetic microregions, the effective sizes and concentration of which depend on the damage dose. It has been shown that, at damage doses higher than ˜55 dpa, small spontaneous magnetization and magnetization hysteresis, which are characteristic of the ferromagnetic state, appear in the samples. It is assumed that the ferromagnetic microregions are the nuclei of the α' phase and the radiation-induced segregation microregions, in which the spacing between the nearest iron atoms exceeds the critical distance that determines the change in the sign of exchange interaction. Arguments in favor of this assumption are presented.

  18. Environmental resistance of oxide tags fabricated on 304L stainless steel via nanosecond pulsed laser irradiation

    DOE PAGESBeta

    Lawrence, Samantha Kay; Adams, David P.; Bahr, David F.; Moody, Neville R.

    2015-11-14

    Nanosecond pulsed laser irradiation was used to fabricate colored, mechanically robust oxide “tags” on 304L stainless steel. Immersion in simulated seawater solution, salt fog exposure, and anodic polarization in a 3.5% NaCl solution were employed to evaluate the environmental resistance of these oxide tags. Single layer oxides outside a narrow thickness range (~ 100–150 nm) are susceptible to dissolution in chloride containing environments. The 304L substrates immediately beneath the oxides corrode severely—attributed to Cr-depletion in the melt zone during laser processing. For the first time, multilayered oxides were fabricated with pulsed laser irradiation in an effort to expand the protectivemore » thickness range while also increasing the variety of film colors attainable in this range. Layered films grown using a laser scan rate of 475 mm/s are more resistant to both localized and general corrosion than oxides fabricated at 550 mm/s. Furthermore, in the absence of pre-processing to mitigate Cr-depletion, layered films can enhance environmental stability of the system.« less

  19. Microstructural evolution of P92 ferritic/martensitic steel under Ar{sup +} ion irradiation at elevated temperature

    SciTech Connect

    Jin Shuoxue; Guo Liping; Li Tiecheng; Chen Jihong; Yang Zheng; Luo Fengfeng; Tang Rui; Qiao Yanxin; Liu Feihua

    2012-06-15

    Irradiation damage in P92 ferritic/martensitic steel irradiated by Ar{sup +} ion beams to 7 and 12 dpa at elevated temperatures of 290 Degree-Sign C, 390 Degree-Sign C and 550 Degree-Sign C has been investigated by transmission electron microscopy, scanning electron microscopy and atomic force microscopy. The precipitate periphery (the matrix/carbide interface) was amorphized only at 290 Degree-Sign C, while higher irradiation temperature could prevent the amorphization. The formation of the small re-precipitates occurred at 290 Degree-Sign C after irradiation to 12 dpa. With the increase of irradiation temperature and dose, the phenomenon of re-precipitation became more severe. The voids induced by irradiation were observed after irradiation to 7 dpa at 550 Degree-Sign C, showing that high irradiation temperature ({>=} 550 Degree-Sign C) was a crucial factor which promoted the irradiation swelling. Energy dispersive X-ray analysis revealed that segregation of Cr and W in the voids occurred under irradiation, which may influence mechanical properties of P92 F/M steel. - Graphical Abstract: High density of small voids, about 2.5 nm in diameter, was observed after irradiation to 12 dpa at 550 Degree-Sign C, which was shown in panel a (TEM micrograph). As shown in panel b (SEM image), a large number of nanometer-sized hillocks were formed in the surface irradiated at 550 Degree-Sign C, and the mean size was {approx} 30 nm. The formation of the nanometer-sized hillocks might be due to the voids that appeared as shown in TEM images (panel a). High irradiation temperature ({>=} 550 Degree-Sign C) was a crucial factor for the formation of void swelling. Highlights: Black-Right-Pointing-Pointer Small carbides re-precipitated in P92 matrix irradiated to 12 dpa at 290 Degree-Sign C. Black-Right-Pointing-Pointer High density of voids was observed at 550 Degree-Sign C. Black-Right-Pointing-Pointer Segregation of Cr and W in voids occurred under irradiation.

  20. Influence of Ar-ions irradiation on the oxidation behavior of ferritic-martensitic steel P92 in supercritical water

    NASA Astrophysics Data System (ADS)

    Huang, Xi; Shen, Yinzhong; Zhu, Jun

    2015-02-01

    The corrosion behavior of ferritic-marensitic steel P92 with and without Ar-ions irradiation in supercritical water at 823 K(550 °C)/25 MPa for different exposure times was investigated by a variety of characterization techniques. A distinct difference in oxidation morphology between irradiated and unirradiated samples was observed. The oxide morphology of samples with a relatively moderate radiation intensity was similar with that of samples without irradiation. Many small oxide particles were observed in the region with a relatively high radiation intensity but their size was increased gradually with increasing exposure times. Exfoliation of oxide layer occurred for irradiated samples exposed for 100 h. Chromium-rich oxide layer with a chromium content of more than 20 wt pct along with a small-scale three-layer oxide structures were observed in Ar-ions irradiated samples, arising from the microstructural change in steel samples after the irradiation. Mechanism for the exfoliation of oxide layer is also discussed.

  1. Irradiation-assisted stress corrosion cracking of model austenitic stainless steel.

    SciTech Connect

    Chung, H. M.; Ruther, W. E.; Strain, R. V.; Shack, W. J.; Karlsen, T. M.

    1999-10-26

    Slow-strain-rate tensile (SSRT) tests were conducted on model austenitic stainless steel (SS) alloys that were irradiated at 289 C in He. After irradiation to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup 2} and {approx} 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV), significant heat-to-heat variations in the degree of intergranular and transgranular stress corrosion cracking (IGSCC and TGSCC) were observed. At {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2}, a high-purity heat of Type 316L SS that contains a very low concentration of Si exhibited the highest susceptibility to IGSCC. In unirradiated state, Types 304 and 304L SS did not exhibit a systematic effect of Si content on alloy strength. However, at {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2}, yield and maximum strengths decreased significantly as Si content was increased to >0.9 wt.%. Among alloys that contain low concentrations of C and N, ductility and resistance to TGSCC and IGSCC were significantly greater for alloys with >0.9 wt.% Si than for alloys with <0.47 wt.% Si. Initial data at {approx}0.9 x 10{sup 21} n {center_dot} cm{sup -2} were also consistent with the beneficial effect of high Si content. This indicates that to delay onset of and reduce susceptibility to irradiation-assisted stress corrosion cracking (IASCC), at least at low fluence levels, it is helpful to ensure a certain minimum concentration of Si. High concentrations of Cr were also beneficial; alloys that contain <15.5 wt.% Cr exhibited greater susceptibility to IASCC than alloys with {approx}18 wt.% Cr, whereas an alloy that contains >21 wt.% Cr exhibited less susceptibility than the lower-Cr alloys under similar conditions.

  2. Stress corrosion cracking behavior of irradiated model austenitic stainless steel alloys.

    SciTech Connect

    Chung, H. M.; Karlsen, T. M.; Ruther, W. E.; Shack, W. J.; Strain, R. V.

    1999-07-16

    Slow-strain-rate tensile tests (SSRTs) and posttest fractographic analyses by scanning electron microscopy were conducted on 16 austenitic stainless steel (SS) alloys that were irradiated at 289 C in He. After irradiation to {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2} and {approx}0.9 x 10{sup 21} n{center_dot}cm{sup {minus}2} (E >1 MeV), significant heat-to-heat variations in the degree of intergranular and transgranular stress corrosion cracking (IGSCC and TGSCC) were observed. Following irradiation to a fluence of {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2}, a high-purity laboratory heat of Type 316L SS (Si {approx} 0.024 wt%) exhibited the highest susceptibility to IGSCC. The other 15 alloys exhibited negligible susceptibility to IGSCC at this low fluence. The percentage of TGSCC on the fracture surfaces of SSRT specimens of the 16 alloys at {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2} (E > 1 MeV) could be correlated well with N and Si concentrations; all alloys that contained <0.01 wt.% N and <1.0 wt. % Si were susceptible, whereas all alloys that contained >0.01 wt.% N or >1.0 wt.% Si were relatively resistant. High concentrations of Cr were beneficial. Alloys that contain <15.5 wt.% Cr exhibited greater percentages of TGSCC and IGSCC than those alloys with {approx}18 wt.% Cr, whereas an alloy that contains >21 wt.% Cr exhibited less susceptibility than the lower-Cr alloys under similar conditions.

  3. Heavy-section steel irradiation program. Volume 4, No. 2. Semiannual progress report, April 1993--September 1993

    SciTech Connect

    Corwin, W.R.

    1995-03-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents which have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV`s fracture resistance which occurs during service, since without that radiation damage, it is virtually impossible to postulate a realistic scenario that would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established to provide a quantitative assessment of the effects of neutron irradiation on the material behavior and, in particular, the fracture toughness properties of typical pressure-vessel steels. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into 14 tasks: (1) program management, (2) fracture toughness (K{sub lc}) curve shift in high-copper welds, (3) crack-arrest toughness (K{sub la}) curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K{sub lc} and K{sub la} curve shifts in low upper-shelf (LUS) welds, (6) annealing effects in LUS welds, (7) irradiation effects in a commercial LUS weld, (8) microstructural analysis of irradiation effects, (9) in-service aged material evaluations, (10) correlation monitor materials, (11) special technical assistance, (12) Japan Power Development Reactor steel examination, (13) technical assistance for Joint Coordinating Committee on Civilian Nuclear Reactor Safety (JCCCNRS) Working Groups 3 and 12, and (14) additional requirements for materials.

  4. Depth distribution of Frank loop defects formed in ion-irradiated stainless steel and its dependence on Si addition

    NASA Astrophysics Data System (ADS)

    Chen, Dongyue; Murakami, Kenta; Dohi, Kenji; Nishida, Kenji; Soneda, Naoki; Li, Zhengcao; Liu, Li; Sekimura, Naoto

    2015-12-01

    Although heavy ion irradiation is a good tool to simulate neutron irradiation-induced damages in light water reactor, it produces inhomogeneous defect distribution. Such difference in defect distribution brings difficulty in comparing the microstructure evolution and mechanical degradation between neutron and heavy ion irradiation, and thus needs to be understood. Stainless steel is the typical structural material used in reactor core, and could be taken as an example to study the inhomogeneous defect depth distribution in heavy ion irradiation and its influence on the tested irradiation hardening by nano-indentation. In this work, solution annealed stainless steel model alloys are irradiated by 3 MeV Fe2+ ions at 400 °C to 3 dpa to produce Frank loops that are mainly interstitial in nature. The silicon content of the model alloys is also tuned to change point defect diffusion, so that the loop depth distribution influenced by diffusion along the irradiation beam direction could be discussed. Results show that in low Si (0% Si) and base Si (0.42% Si) samples the depth distribution of Frank loop density quite well matches the dpa profile calculated by the SRIM code, but in high Si sample (0.95% Si), the loop number density in the near-surface region is very low. One possible explanation could be Si's role in enhancing the effective vacancy diffusivity, promoting recombination and thus suppressing interstitial Frank loops, especially in the near-surface region, where vacancies concentrate. By considering the loop depth distribution, the tested irradiation hardening is successfully explained by the Orowan model. A hardening coefficient of around 0.30 is obtained for all the three samples. This attempt in interpreting hardening results may make it easier to compare the mechanical degradation between different irradiation experiments.

  5. Features of structure-phase transformations and segregation processes under irradiation of austenitic and ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Neklyudov, I. M.; Voyevodin, V. N.

    1994-09-01

    The difference between crystal lattices of austenitic and ferritic steels leads to distinctive features in mechanisms of physical-mechanical change. This paper presents the results of investigations of dislocation structure and phase evolution, and segregation phenomena in austenitic and ferritic-martensitic steels and alloys during irradiation with heavy ions in the ESUVI and UTI accelerators and by neutrons in fast reactors BOR-60 and BN-600. The influence of different factors (including different alloying elements) on processes of structure-phase transformation was studied.

  6. Fractographic examination of reduced activation ferritic/martensitic steel charpy specimens irradiated to 30 dpa at 370{degrees}C

    SciTech Connect

    Gelles, D.S.; Hamilton, M.L.; Schubert, L.E.

    1996-10-01

    Fractographic examinations are reported for a series of reduced activation ferritic/Martensitic steel Charpy impact specimens tested following irradiation to 30 dpa at 370{degrees}C in FFTF. One-third size specimens of six low activation steels developed for potential application as structural materials in fusion reactors were examined. A shift in brittle fracture appearance from cleavage to grain boundary failure was noted with increasing manganese content. The results are interpreted in light of transmutation induced composition changes in a fusion environment.

  7. Effects of neutron irradiation on microstructures and hardness of stainless steel weld-overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K.

    2014-06-01

    The microstructures and the hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation at a dose of 7.2 × 1019 n cm-2 (E > 1 MeV) and a flux of 1.1 × 1013 n cm-2 s-1 at 290 °C were investigated by atom probe tomography and by a nanoindentation technique. To isolate the effects of the neutron irradiation, we compared the results of the measurements of the neutron-irradiated samples with those from a sample aged at 300 °C for a duration equivalent to that of the irradiation. The Cr concentration fluctuation was enhanced in the δ-ferrite phase of the irradiated sample. In addition, enhancement of the concentration fluctuation of Si, which was not observed in the aged sample, was observed. The hardening in the δ-ferrite phase occurred due to both irradiation and aging; however, the hardening of the irradiated sample was more than that expected from the Cr concentration fluctuation, which suggested that the Si concentration fluctuation and irradiation-induced defects were possible origins of the additional hardening.

  8. Irradiation embrittlement of 2 {1}/{4}Cr-1Mo steel at 400°C and its electrochemical evaluation

    NASA Astrophysics Data System (ADS)

    Nishiyama, Y.; Fukaya, K.; Suzuki, M.; Eto, M.

    1998-10-01

    The effect of neutron irradiation on mechanical properties of normalized and tempered 2 {1}/{4}Cr-1Mo steel was evaluated by conducting postirradiation tensile and Charpy impact tests. The specimens were irradiated at 400°C to a fluence as high as 3 × 10 24 n/m 2 ( E > 1 MeV). Only slight hardening was observed because of the high temperature of irradiation. However, irradiation at 400°C to a fluence larger than 1 × 10 24 n/m 2 ( E > 1 MeV) caused high Charpy shifts accompanied by intergranular fracture. Results of electrochemical tests indicated that a possible element responsible for intergranular fracture was phosphorus.

  9. Comparison of the mechanical properties of T91 steel from the MEGAPIE, and TWIN-ASTIR irradiation programs

    NASA Astrophysics Data System (ADS)

    Konstantinović, M. J.; Stergar, E.; Lambrecht, M.; Gavrilov, S.

    2016-01-01

    The mechanical properties of spallation target components exposed to combined effects of proton and neutron irradiations and in contact with liquid metal provide important information for the assessment of structural component integrity, which is crucial for the design of accelerator driven reactor concepts such as the MYRRHA reactor. In this study we perform tensile tests on T91 steel samples extracted from the MEGAPIE, and from the TWIN-ASTIR experiment. The tests are performed at different temperatures as well as with and without the contact with liquid metal. In both groups of samples we observed significant influence of liquid metal on the tensile properties, in particular reduction of total elongation. The influence of different conditions in two irradiation programs on the mechanical properties, such as irradiation temperature fluctuations, the presence of neutron/proton irradiation, with and without the contact with lead-bismuth eutectic, different flux and fluence, are also discussed.

  10. Magnetic evaluation of irradiation hardening in A533B reactor pressure vessel steels: Magnetic hysteresis measurements and the model analysis

    NASA Astrophysics Data System (ADS)

    Kobayashi, S.; Yamamoto, T.; Klingensmith, D.; Odette, G. R.; Kikuchi, H.; Kamada, Y.

    2012-03-01

    We report results of measurements of magnetic minor hysteresis loops for neutron-irradiated A533B nuclear reactor pressure vessel steels varying alloy composition and irradiation condition. A minor-loop coefficient, which is obtained from a scaling power law between minor-loop parameters exhibits a steep decrease just after irradiation, followed by a maximum in the intermediate fluence regime for most alloys. A model analysis assuming Avrami-type growth for Cu-rich precipitates and an empirical logarithmic law for relaxation of residual stress demonstrates that an increment of the coefficient due to Cu-rich precipitates increases with Cu and Ni contents and is in proportion to a yield stress change, which is related to irradiation hardening.