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Sample records for fe-15cr-20ni steels irradiated

  1. 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.

  2. Alumina-Forming Austenitic Stainless Steels Strengthened by Laves Phase and MC Carbide Precipitates

    NASA Astrophysics Data System (ADS)

    Yamamoto, Y.; Brady, M. P.; Lu, Z. P.; Liu, C. T.; Takeyama, M.; Maziasz, P. J.; Pint, B. A.

    2007-11-01

    Creep strengthening of Al-modified austenitic stainless steels by MC carbides or Fe2Nb Laves phase was explored. Fe-20Cr-15Ni-(0-8)Al and Fe-15Cr-20Ni-5Al base alloys (at. pct) with small additions of Nb, Mo, W, Ti, V, C, and B were cast, thermally-processed, and aged. On exposure from 650 °C to 800 °C in air and in air with 10 pct water vapor, the alloys exhibited continuous protective Al2O3 scale formation at an Al level of only 5 at. pct (2.4 wt pct). Matrices of the Fe-20Cr-15Ni-5Al base alloys consisted of γ (fcc) + α (bcc) dual phase due to the strong α-Fe stabilizing effect of the Al addition and exhibited poor creep resistance. However, adjustment of composition to the Fe-15Cr-20Ni-5Al base resulted in alloys that were single-phase γ-Fe and still capable of alumina scale formation. Alloys that relied solely on Fe2Nb Laves phase precipitates for strengthening exhibited relatively low creep resistance, while alloys that also contained MC carbide precipitates exhibited creep resistance comparable to that of commercially available heat-resistant austenitic stainless steels. Phase equilibria studies indicated that NbC precipitates in combination with Fe2Nb were of limited benefit to creep resistance due to the solution limit of NbC within the γ-Fe matrix of the alloys studied. However, when combined with other MC-type strengtheners, such as V4C3 or TiC, higher levels of creep resistance were obtained.

  3. 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.

  4. 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)

  5. 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.

  6. 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%.

  7. 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.

  8. 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.

  9. 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.

  10. Microstructural change on electron irradiated oxide dispersion strengthened ferritic steels

    NASA Astrophysics Data System (ADS)

    Kinoshita, H.; Akasaka, N.; Takahashi, H.; Shibahara, I.; Onose, S.

    1992-09-01

    Oxide dispersion strengthened (ODS) ferritic steels were irradiated in a high voltage electron microscope (HVEM) to study their response to irradiation. Fe-13Cr with 0.25 wt% Y2O3 as dispersed particles and containing additions of either 0.45% Nb, 0.45% V and 0.67% Zr were irradiated at 673 and 723 K up to 15 dpa. The Y2O3 particles in all specimens were stable under these irradiation conditions. During irradiation, two types of dislocations were formed but observable voids were not formed. Furthermore, plate-like and granular-like precipitates formed in both the irradiated and nonirradiated regions.

  11. 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.

  12. 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).

  13. Microstructure evolution during helium irradiation and post-irradiation annealing in a nanostructured reduced activation steel

    NASA Astrophysics Data System (ADS)

    Liu, W. B.; Ji, Y. Z.; Tan, P. K.; Zhang, C.; He, C. H.; Yang, Z. G.

    2016-10-01

    Severe plastic deformation, intense single-beam He-ion irradiation and post-irradiation annealing were performed on a nanostructured reduced activation ferritic/martensitic (RAFM) steel to investigate the effect of grain boundaries (GBs) on its microstructure evolution during these processes. A surface layer with a depth-dependent nanocrystalline (NC) microstructure was prepared in the RAFM steel using surface mechanical attrition treatment (SMAT). Microstructure evolution after helium (He) irradiation (24.8 dpa) at room temperature and after post-irradiation annealing was investigated using Transmission Electron Microscopy (TEM). Experimental observation shows that GBs play an important role during both the irradiation and the post-irradiation annealing process. He bubbles are preferentially trapped at GBs/interfaces during irradiation and cavities with large sizes are also preferentially trapped at GBs/interfaces during post-irradiation annealing, but void denuded zones (VDZs) near GBs could not be unambiguously observed. Compared with cavities at GBs and within larger grains, cavities with smaller size and higher density are found in smaller grains. The average size of cavities increases rapidly with the increase of time during post-irradiation annealing at 823 K. Cavities with a large size are observed just after annealing for 5 min, although many of the cavities with small sizes also exist after annealing for 240 min. The potential mechanism of cavity growth behavior during post-irradiation annealing is also discussed.

  14. 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.

  15. Sensitivity of ultrasonic nonlinearity to irradiated, annealed, and re-irradiated microstructure changes in RPV steels

    SciTech Connect

    Matlack, Katie; Kim, J-Y.; Wall, J.J.; Jacobs, L.J.; Sokolov, Mikhail A

    2014-05-01

    The planned life extension of nuclear reactors throughout the US and abroad will cause reactor vessel and internals materials to be exposed to more neutron irradiation than was originally intended. A nondestructive evaluation (NDE) method to monitor radiation damage would enable safe and cost-effective continued operation of nuclear reactors. Radiation damage in reactor pressure vessel (RPV) steels causes microstructural changes that leave the material in an embrittled state. Nonlinear ultrasound is an NDE technique quantified by the measurable acoustic nonlinearity parameter, which is sensitive to microstructural changes in metallic materials such as dislocations, precipitates and their combinations. Recent research has demonstrated the sensitivity of the acoustic nonlinearity parameter to increasing neutron fluence in representative RPV steels. The current work considers nonlinear ultrasonic experiments conducted on similar RPV steel samples that had a combination of irradiation, annealing, re-irradiation, and/or re-annealing to a total neutron fluence of 0.5 5 1019 n/cm2 (E > 1 MeV) at an irradiation temperature of 290 C. The acoustic nonlinearity parameter generally increased with increasing neutron fluence, and consistently decreased from the irradiated to the annealed state over different levels of neutron fluence. Results of the measured acoustic nonlinearity parameter are compared with those from previous measurements on other RPV steel samples. This comprehensive set of results illustrates the dependence of the measured acoustic nonlinearity parameter on neutron fluence, material composition, irradiation temperature and annealing.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. APT characterization of irradiated high nickel RPV steels

    NASA Astrophysics Data System (ADS)

    Miller, M. K.; Russell, K. F.; Sokolov, M. A.; Nanstad, R. K.

    2007-04-01

    Samples of the welds from the Midland and Palisades reactors in the unirradiated condition and after neutron irradiation to a high fluence of up to 3.4 × 1023 m-2 (E > 1 MeV) have been characterized with the Oak Ridge National Laboratory's local electrode atom probe. High number densities, ∼5 and ∼7 × 1023 m-3, respectively, of ∼2-nm-diameter copper-, nickel-, manganese- and silicon-enriched precipitates were observed after neutron irradiation. These copper-enriched precipitates were observed both in the matrix of the steel and also preferentially located along the dislocations. No appreciable differences were observed in the sizes or the compositions of the precipitates in the matrix and on the dislocations. The average interparticle distance along the dislocations was 11 ± 3 nm. Phosphorus segregation was also evident along the dislocations in both welds. No other nanoscale intragranular phases were observed in these neutron irradiated welds.

  1. IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

    SciTech Connect

    Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

    2002-09-01

    Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. 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. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

  2. Oxide dispersion strengthened steel irradiation with helium ions

    NASA Astrophysics Data System (ADS)

    Pouchon, M. A.; Chen, J.; Döbeli, M.; Hoffelner, W.

    2006-06-01

    Oxide dispersion strengthened (ODS) ferritic steels are investigated as possible structural material for the future generation of high temperature gas cooled nuclear reactors. ODS-steels are considered to replace other high temperature materials for tubing or structural parts. The oxide particles serve for interfacial pinning of moving dislocations. Therefore, the creep resistance is improved. In case of the usage of these materials in reactors, the behavior under irradiation must be further clarified. In this paper the effects induced by 4He2+ implantation into a ferritic ODS steel are investigated. The fluence ranges from 1016 to 1017 cm-2 and the energy from 1 to 2 MeV. The induced swelling is investigated for implantations at room temperature and 470 K. It is derived from the irradiation induced surface displacement, which is measured with an atomic force microscope (AFM). With a displacement damage of 0.6 dpa, a volume increase of 0.65% is observed at room temperature and 0.33% at 470 K. A cross-sectional cut is performed by focused ion beam and investigated by transmission electron microcopy (TEM). The defect density observed on the TEM micrographs agrees well with the computational simulation (TRIM) of the damage profile.

  3. Precipitation and cavity formation in austenitic stainless steels during irradiation

    SciTech Connect

    Lee, E.H.; Rowcliffe, A.F.; Mansur, L.K.

    1981-01-01

    Microstructural evolution in austenitic stainless steels subjected to displacement damage at high temperature is strongly influenced by the interactions between helium atoms and second phase particles. Cavity nucleation occurs by the trapping of helium at partially coherent particle-matrix interfaces. The recent precipitate point defect collector theory describes the more rapid growth of precipitate-attached cavities compared to matrix cavities where the precipitate-matrix interface collects point defects to augment the normal point deflect flux to the cavitry. Data are presented which support these ideas. It is shown that during nickel ion irradiation of a titanium-modified stainless steel at 675/sup 0/C the rate of injection of helium has a strong effect on the total swelling and also on the nature and distribution of precipitate phases.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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).

  10. Post-irradiation annealing effects of austenitic stainless steels in IASCC

    SciTech Connect

    Katsura, Ryoei; Ishiyama, Yoshihide; Yokota, Norikatu; Kato, Takahiko; Nakata, Kiyotomo; Fukuya, Kouji; Sakamoto, Hiroshi; Asano, Kyoichi

    1998-12-31

    Post-irradiation annealing effects on the thermal sensitization and IASCC recovery for highly irradiated types 304 and 316L stainless steels were investigated using EPR and SSR tests. Irradiated type 316L stainless steel (neutron fluence: 8 x 10{sup 25} n/m{sup 2}, E > 1 MeV) was not sensitized and IGSCC susceptibility significantly was reduced to 7--0% at 400--700 C (x1h) from 23% at as-irradiated condition. Irradiated type 304 stainless steel (neutron fluence: 1.2 x 10{sup 26} n/m{sup 2}, E > 1MeV) was more easily sensitized than unirradiated material and IGSCC susceptibility was reduced to 62--45% at 400--500 C from 95% at the as-irradiated condition. These results on types 304 and 316L stainless steels indicated that the thermal healing technic enhanced IASCC recovery.

  11. The effect of oxygen on void stability in ion-irradiated steel

    NASA Astrophysics Data System (ADS)

    Seitzman, Larry E.; Dodd, R. Arthur; Kulcinski, Gerald L.

    1990-07-01

    The effect of oxygen on void stability in an Fe-17Ni-13Cr austenitic ternary alloy has been investigated using 15 MeV nickel-ion irradiation at elevated temperatures and preimplantation of 6 MeV oxygen at room temperature. The nickel irradiation was performed over a temperature range of 550 °C to 650 °C. Utilizing transverse specimen preparation techniques, the irradiated steel was examined by transmission electron microscopy (TEM). As little as 10 appm preimplanted oxygen caused a significant increase in the void number density when the steel was irradiated at 550 °C. A near-surface void-denuded zone occurs in the irradiated steel, while a region depleted of visible voids also occurs in the steel injected with 300 appm oxygen or greater and irradiated at 550 °C.

  12. 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.

  13. Hardening and microstructural evolution in A533B steels under neutron irradiation and a direct comparison with electron irradiation

    NASA Astrophysics Data System (ADS)

    Fujii, K.; Nakata, H.; Fukuya, K.; Ohkubo, T.; Hono, K.; Nagai, Y.; Hasegawa, M.; Yoshiie, T.

    2010-05-01

    A533B steels irradiated at 290 °C up to 10 mdpa in the Kyoto University Reactor were examined by hardness, positron annihilation and atom probe measurements. Dose dependent irradiation hardening and formation of Cu-rich clusters were confirmed in medium Cu (0.12% and 0.16%Cu) steels whereas neither hardening nor cluster formation was detected in low Cu (0.03%Cu) steel. No microvoids were formed in any of the steels. Post-irradiation annealing in medium Cu steels revealed that the hardening recovery at temperatures above 350-400 °C could be attributed to compositional changes and dissociation of the Cu-rich clusters. Compared to electron irradiation at almost the same dose and dose rate, KUR irradiation caused almost the same hardening and produced Cu-rich clusters, more solute-enriched with larger size and lower density. Considering lower production of freely-migrating vacancies in neutron irradiation, the results suggested that cascades enhance the formation of Cu-rich clusters.

  14. 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.

  15. 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.

  16. 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 reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially in the first wall and blanket. There has been limited work on fatigue in irradiated alloys but none on irradiated materials containing significant amounts of irradiation-induced helium. To provide scoping data and to study the effects of irradiation on fatigue behavior, 20%-cold-worked type 316 stainless steel from the MFE reference heat was studied.

  17. 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.

  18. Irradiation Induced Defect Characterization in Reactor Pressure Vessel Steel by Small Angle Neutron Scattering

    NASA Astrophysics Data System (ADS)

    Han, Yougn-Soo; Shin, Eun-Joo; Lee, Chang-Hee; Park, Duck-Gun

    The degradation of the mechanical properties of the RPV (Reactor Pressure Vessel) steel during an irradiation in a nuclear power plant is closely related to the irradiation induced defects. The size of these defects is known to be a few nanometer, and the small angle neutron scattering technique is regarded as the best non destructive technique to characterize the nano sized inhomogeneities in bulk samples. The investigated the RPV steel has been used in YeongKwang nuclear power plant at Korea and the Cu content of the RPV steel is 0.06 wt%. The RPV steel was irradiated in the HANARO reactor at KAERI. The small angle neutron scattering experiments were performed by the SANS instrument in the HANARO reactor. The nano sized irradiation induced defects were quantitatively analyzed by SANS and the type of the irradiation induced defects was discussed in detail. The relation between irradiation induced defects and the yield strength was investigated. The characteristics of irradiation induced defects in low Cu containing RPV steel were discussed.

  19. Irradiation damage behavior of low alloy steel wrought and weld materials

    SciTech Connect

    Stofanak, R.J.; Poskie, T.J.; Li, Y.Y.; Wire, G.L.

    1993-10-01

    A study was undertaken to evaluate the irradiation damage response of several different types of low alloy steel: vintage type ASTM A302 Grade B (A302B) plates and welds containing different Ni and Cu concentrations, 3.5% Ni steels similar to ASTM A508 Class 4, welds containing about 1% Ni (similar to type 105S), and 3.5% Ni steels with ``superclean`` composition. All materials were irradiated at several different irradiation damage levels ranging from 0.0003 to 0.06 dpa at 232C (450F). Complete Charpy V-notch impact energy transition temperature curves were generated for all materials before and after irradiation to determine transition temperature at 4IJ (30 ft-lb) or 47J (35 ft-lb) and the upper shelf energy. Irradiation damage behavior was measured by shift in Charpy 41J or 47J transition temperature ({Delta}TT4{sub 41J} or {Delta}TT{sub 47J}) and lowering of upper shelf Charpy energy at a given irradiation damage level. It was found that chemical composition greatly influenced irradiation damage behavior; highest irradiation damage (greatest {Delta}TT) was found in an A302B type weld containing 1.28% Ni and 0.20% Cu while the least damage was found in 3.5% Ni, 0.05% Cu, superclean wrought materials. Combination of Ni and Cu was found to affect irradiation damage behavior at higher irradiation damage levels in the A302B welds where the 1.28% Ni, 0.20% Cu weld showed more damage than a 0.60% Ni, 0.31% Cu weld. For the 3.5% Ni steels, fabrication influenced irradiation behavior in that a silicon (Si) killed material showed greater irradiation damage than a low silicon material. In general, the 3.5% Ni materials with low copper showed less irradiation damage than the A302B materials.

  20. 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.

  1. 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.

  2. 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.

  3. JRQ and JPA irradiated and annealed reactor pressure vessel steels studied by positron annihilation

    NASA Astrophysics Data System (ADS)

    Slugeň, Vladimír; Gokhman, Oleksandr; Pecko, Stanislav; Sojak, Stanislav; Bergner, Frank

    2016-03-01

    The paper is focused on a comprehensive study of JRQ and JPA reactor pressure vessel steels from the positron annihilation lifetime spectroscopy (PALS) point of view. Based on our more than 20 years' experience with characterization of irradiated reactor steels, we confirmed that defects after irradiation start to grow and/or merge into bigger clusters. Experimental results shown that JPA steel is more sensitive to the creation of irradiation-induced defects than JRQ steel. It is most probably due to high copper content (0.29 wt.% in JPA) and copper precipitation has a major impact on neutron-induced defect creation at the beginning of the irradiation. Based on current PALS results, no large vacancy clusters were formed during irradiation, which could cause dangerous embrittlement concerning operation safety of nuclear power plant. The combined PALS, small angle neutron scattering and atomic probe tomography studies support the model for JRQ and JPA steels describing the structure of irradiation-induced clusters as agglomerations of vacancy clusters (consisting of 2-6 vacancies each) and are separated from each other by a distribution of atoms.

  4. Response of nanoclusters in a 9Cr ODS steel to 1 dpa, 525 ?C proton irradiation

    SciTech Connect

    Certain, Alicia; Field, Kevin G; Allen, Todd R.; Sridharan, K.; Miller, Michael K; Bentley, James; Busby, Jeremy T

    2010-01-01

    Ferritic-martensitic (F/M) alloys are expected to play an important role as cladding or structural components in Generation IV and other advanced nuclear systems operating in the temperature range 350-700 C and to doses up to 200 displacements per atom (dpa). Oxide dispersion strengthened (ODS) F/M steels have been developed to operate at higher temperatures than traditional F/M steels. These steels contain nanometer-sized Y-Ti-O nanoclusters for additional strengthening. A proton irradiation to 1 dpa at 525 C has been performed on a 9Cr ODS steel to determine the nanocluster stability at low dose. The evolution of the nanocluster population and the composition at the nanocluster-matrix interface were studied using electron microscopy and atom probe tomography. The data from this study are contrasted to those from a previous study on heavy-ion irradiated 9Cr ODS steel.

  5. 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.

  6. 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

  7. 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).

  8. 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).

  9. 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).

  10. 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.

  11. The radiation swelling effect on fracture properties and fracture mechanisms of irradiated austenitic steels. Part I. Ductility and fracture toughness

    NASA Astrophysics Data System (ADS)

    Margolin, B.; Sorokin, A.; Shvetsova, V.; Minkin, A.; Potapova, V.; Smirnov, V.

    2016-11-01

    The radiation swelling effect on the fracture properties of irradiated austenitic steels under static loading has been studied and analyzed from the mechanical and physical viewpoints. Experimental data on the stress-strain curves, fracture strain, fracture toughness and fracture mechanisms have been represented for austenitic steel of 18Cr-10Ni-Ti grade (Russian analog of AISI 321 steel) irradiated up to neutron dose of 150 dpa with various swelling. Some phenomena in mechanical behaviour of irradiated austenitic steels have been revealed and explained as follows: a sharp decrease of fracture toughness with swelling growth; untypical large increase of fracture toughness with decrease of the test temperature; some increase of fracture toughness after preliminary cyclic loading. Role of channel deformation and channel fracture has been clarified in the properties of irradiated austenitic steel and different tendencies to channel deformation have been shown and explained for the same austenitic steel irradiated at different temperatures and neutron doses.

  12. 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.

  13. 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.

  14. 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).

  15. 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.

  16. Void formation and microstructural development in oxide dispersion strengthened ferritic steels during electron-irradiation

    NASA Astrophysics Data System (ADS)

    Saito, J.; Suda, T.; Yamashita, S.; Ohnuki, S.; Takahashi, H.; Akasaka, N.; Nishida, M.; Ukai, S.

    1998-10-01

    ODS ferritic steels (13Cr-0.5Ti-0.2Y 2O 3) were prepared by the mechanical alloying method followed by the hot extrusion and several heat treatments including recrystallization. ODS steels with different heat treatment and a ferritic/martensitic (F/M) steel for the reference were irradiated to 12 dpa at 670-770 K in HVEM. After recrystallization, the dislocation density decreased with increasing grain size, however, the oxide particles did not show any obvious change in the size and the number density. During the electron-irradiation the microstructure was relatively stable, i.e. oxide particles showed good stability and the dislocation density remained almost constant. A limited void formation was observed in the specimens, and the suppressive effect due to dislocations with high number density was confirmed. From these results, the behavior of microstructure and the limited void formation in ODS steels have been discussed.

  17. Irradiation embrittlement of reactor pressure vessel steel outside the astm specification A508 CL2

    NASA Astrophysics Data System (ADS)

    Pachur, D.; Krawczynski, S. J.; Derz, H.; Pott, G.

    1990-04-01

    Radiation embrittlement of reactor pressure vessel steels is of considerable significance for safety engineering. Steel manufacturers must therefore comply with specifications defined by national design codes. The extent to which a steel deviating from the specification is influenced by irradiation is being examined under the German Research Programme on the Integrity of Reactor Components. Charpy-V specimens were taken from a forged steel block longitudinally and vertically to the direction of main deformation and irradiated in the FRJ-1 research reactor at a temperature of 288 °C corresponding to the operating temperature of power reactors. The neutron fluences obtained ranged between 0.8 × 10 19 and 8 × 10 19n/ cm2. Instrumented pendulum impact tests have been evaluated and the load signals measured were analysed, fitting and calculating transition temperature curves and trend curves.

  18. Helium effects on neutron-irradiated Cr-Mo ferritic steels: A review of recent results

    SciTech Connect

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

    1988-01-01

    Large amounts of transmutation helium will be produced in the first wall of a fusion reactor by the high-energy neutrons from the fusion reaction. Since no fusion reactor is available, the effect of simultaneous helium production and displacement damage from neutron irradiation must be simulated. One method that has been used in ferritic steels is to add nickel to the steels and irradiate them in a mixed-spectrum reactor. In such reactors, the fast neutrons produce displacement damage, while helium is produced by a two-step reaction of /sup 58/Ni with thermal neutrons. This technique has been used to investigate the effect of helium on swelling, tensile properties, impact properties, and elevated-temperature embrittlement. Results indicate that helium accelerates swelling and affects tensile and impact properties of Cr-Mo ferritic steels below /approximately/450/degree/C. However, these steels are highly resistant to elevated-temperature helium embrittlement. 44 refs., 6 figs., 3 tabs.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. MICROSTRUCTURAL EXAMINATION OF LOW ACTIVATION FERRITIC STEELS FOLLOWING IRRADIATION IN ORR

    SciTech Connect

    Gelles, David S.

    2002-09-01

    Microstructural examinations are reported for a series of low activation steels containing Mn following irradiation in the Oak Ridge Reactor at 330 and 400 degrees C to approximately 10 dpa. Alloy compositions included 2Cr, 9Cr and 12Cr steels with V to 1.5 percent and W to 1.0 percent. Results include compositional changes in precipitates and microstructural changes as a function of composition and irradiation temperature. It is concluded that temperatures in ORR are on the order of 50 degrees C higher than anticipated.

  4. 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.

  5. 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.

  6. 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.

  7. Metallography studies and hardness measurements on ferritic/martensitic steels irradiated in STIP

    NASA Astrophysics Data System (ADS)

    Zhang, H.; Long, B.; Dai, Y.

    2008-06-01

    In this work metallography investigations and microhardness measurements have been performed on 15 ferritic/martensitic (FM) steels and 6 weld metals irradiated in the SINQ Target Irradiation Program (STIP). The results demonstrate that all the steels have quite similar martensite lath structures. However, the sizes of the prior austenite grain (PAG) of these steels are quite different and vary from 10 to 86 μm. The microstructure in the fusion zones (FZ) of electron-beam welds (EBWs) of 5 steels (T91, EM10, MANET-II, F82H and Optifer-IX) is similar in respect to the martensite lath structure and PAG size. The FZ of the inert-gas-tungsten weld (TIGW) of the T91 steel shows a duplex structure of large ferrite gains and martensite laths. The microhardness measurements indicate that the normalized and tempered FM steels have rather close hardness values. The unusual high hardness values of the EBW and TIGW of the T91 steel were detected, which suggests that these materials are without proper tempering or post-welding heat treatment.

  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. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Heavy-section steel irradiation program. Progress report, October 1992--March 1993

    SciTech Connect

    Corwin, W.R.

    1998-04-01

    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 one of only two more safety-related components 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 RPV`s fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established at Oak Ridge National Laboratory (ORNL) under sponsorship of the Nuclear Regulatory Commission (NRC). The primary goal of this major safety program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior (in particular, the fracture toughness properties) of typical pressure-vessel steels as they relate to light-water-reactor pressure-vessel integrity. The program centers on experimental assessments of irradiation-induced embrittlement (including the completion of certain irradiation studies previously conducted by the Heavy-Section Steel Technology Program) augmented by detailed examinations and modeling of the accompanying microstructural changes. 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.

  15. 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.

  16. 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.

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

    DOE PAGES

    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

  18. 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.

  19. The role of dislocation channeling in IASCC initiation of neutron irradiated austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale Jennings

    The objective of this study was to understand the role of dislocation channeling in the initiation of irradiation-assisted stress corrosion cracking (IASCC) of neutron irradiated austenitic stainless steel using a novel four-point bend test. Stainless steels used in this study were irradiated in the BOR-60 fast reactor at 320 °C, and included a commercial purity 304L stainless steel irradiated to 5.5, 10.2, and 47.5 dpa, and two high purity stainless steels, Fe-18Cr-12Ni and Fe-18Cr-25Ni, irradiated to ~10 dpa. The four-point bend test produced the same relative IASCC susceptibility as constant extension rate tensile (CERT) experiments performed on the same irradiated alloys in boiling water reactor normal water chemistry. The cracking susceptibility of the CP 304L alloy was high at all irradiation dose levels, enhanced by the presence of MnS inclusions in the alloy microstructure, which dissolve in the NWC environment. Dissolution of the MnS inclusion results in formation of an oxide cap that occludes the inclusion site, creating a crevice condition with a high propensity for crack initiation. Crack initiation at these locations was induced by stress concentration at the intersecting grain boundary, resulting from the intersection of a discontinuous dislocation channels (DC). Stress to initiate an IASCC crack decreased with dose due earlier DC initiation. The HP Fe-18Cr-12Ni alloy had low susceptibility to IASCC, while the high Ni alloy exhibited no cracking susceptibility. The difference in susceptibility among these conditions was attributed to the propensity for DCs to transmit across grain boundaries, which controls stress accumulation at DC -- grain boundary intersections.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Microstructural changes of austenitic steels caused by proton irradiation under various conditions

    NASA Astrophysics Data System (ADS)

    Fukuda, T.; Sagisaka, M.; Isobe, Y.; Hasegawa, A.; Sato, M.; Abe, K.; Nishida, Y.; Kamada, T.; Kaneshima, Y.

    2000-12-01

    In austenitic steels used for light water reactors (LWRs), neutron irradiation induces many kinds of degradation. For example, irradiation assisted stress corrosion cracking (IASCC) and swelling are two forms of this degradation. Although there are a great number of studies on radiation induced segregation (RIS) and void swelling at high temperatures (>400° C) corresponding to fast and fusion reactor conditions, up to now there have been few irradiation studies at low temperatures. This paper presents microchemical and microstructural changes in type 347 and 310+Nb stainless steels due to light ion irradiation. These samples were implanted with He+ and irradiated with 2 MeV H2+ at 300°C, 350°C and 400°C. This simulated generation of transmutant He in a fusion environment. In addition, at 300°C test pieces stressed close to the yield stress were also irradiated with the same ions. The irradiation tests were carried out using the Dynamitron accelerator at Tohoku University.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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).

  12. Characterization of Structural Conditions of AISI 316 Analog Stainless Steel Irradiated in the BN-350 Reactor

    SciTech Connect

    Maksimkin, O. P.; Tsai, K V.; Turubarova, L. G.; Doronina, T. A.; Garner, Francis A.

    2004-08-24

    In several recently published studies conducted on a Soviet analog of AISI 321 stainless steel irradiated in either fast reactors or light water reactors, it was shown that the void swelling phenomenon extended to temperatures as low as {approx} 300 C, when produced by neutron irradiation at dpa rates in the range 10(-7 power) to 10(-8 power) dpa/sec. Other studies yielded similar results for AISI 316. In the current study a blanket duct assembly from BN-350, constructed from the Soviet analog of AISI 316, also exhibits swelling at dpa rates on the order of 10(-8 power) dpa/sec, with voids seen as low as 281 C and only 1.3 dpa. It appears that low-temperature swelling at low dpa rates occurs in 300 series stainless steels in general, and during irradiations conducted in either fast or mixed spectrum reactors.

  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

    NASA Astrophysics Data System (ADS)

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

    2015-03-01

    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. 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.

  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. 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.

  18. Charpy impact toughness of martensitic steels irradiated in FFTF: Effect of heat treatment

    NASA Astrophysics Data System (ADS)

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

    Charpy tests were made on plates of 9Cr-1MoVNb and 12Cr-1MoVW steels given four different normalizing-and-tempering treatments. One-third-size Charpy specimens from each steel were irradiated to 7.4 - 8 (times) 10(sup 26) n/m(sup 2) (about 34 - 37 dpa) at 420 C in the Materials Open Test Assembly of the Fast Flux Test Facility. Specimens were also thermally aged to 20000 h at 400 C to determine the effect of aging during irradiation. Previous work on these steels irradiated to 4 - 5 dpa at 365 C in MOTA were reexamined in light of the new results. The tests indicated that prior austenite grain size, which was varied by different normalizing treatments, had an effect on impact behavior of the 9Cr-1MoVNb but not on the 12Cr-1MoVW. Tempering treatment had relatively little effect on the shift in DBTT for both steels. Conclusions are presented on how heat treatment can be used to optimize properties.

  19. 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.

  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. 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.

  2. 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.

  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. 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.

  5. Response of reduced activation ferritic steels to high-fluence ion-irradiation

    NASA Astrophysics Data System (ADS)

    Tanigawa, H.; Ando, M.; Katoh, Y.; Hirose, T.; Sakasegawa, H.; Jitsukawa, S.; Kohyama, A.; Iwai, T.

    2001-09-01

    Effects of high-fluence irradiation in fusion-relevant helium production condition on defect cluster formation and swelling of reduced activation ferritic/martensitic steels (RAFs), JLF-1 (Fe-9Cr-2W-V-Ta) and F82H (Fe-8Cr-2W-V-Ta), have been investigated. Dual-ion (nickel plus helium ions) irradiation using electrostatic accelerators was adopted to simulate fusion neutron environment. The irradiation has been carried out up to a damage level of 100 displacement per atom (dpa) at around 723 K, at the HIT facility in the University of Tokyo. Thin foils for transmission electron microscopy (TEM) were prepared with a focused ion beam (FIB) microsampling system. The system enabled not only the broad cross-sectional TEM observation, but also the detailed study of irradiated microstructure, since unfavorable effects of ferromagnetism of a ferritic steel specimen were completely suppressed with this system by sampling a small volume in interests from the irradiated material.

  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. Irradiation behavior of weldments of austenitic stainless steel made by various welding techniques

    SciTech Connect

    Shiba, Kiyoyuki; Sawai, Tomotsugu; Jitsukawa, Shiro; Hishinuma, Akimichi; Pawel, J.E.

    1996-12-31

    Austenitic stainless steel is one of the candidate materials for nuclear fusion reactor applications. Here, an austenitic stainless steel, 316F, irradiated in the High Flux Isotope Reactor to doses of about 8 to 33 dpa at 400 and 500 C was investigated. Electron beam (EB) welding and metal inert gas (MIG) welding techniques were used to make weldment specimens. Weldment specimens were made from their weld metal or weld joint (including heat affected zone) regions of the weldments. Base metal was also studied for comparison. Microstructures of these specimens were observed by TEM. Tensile tests were carried out at the nominal irradiation temperature in vacuum. Solution annealed 316F showed the large irradiation hardening at 400 C, while the change in yield stress observed at 500 C was not so large. Weldments specimens had the same temperature and dose dependence as the base metal. The differences between EB and MIG after irradiation were small, compared to the differences before irradiation, except for the slight less ductility of MIG weldments. The defect microstructures of weldments were the same as base metal.

  8. 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.

  9. Grafting of HEMA onto dopamine coated stainless steel by 60Co-γ irradiation method

    NASA Astrophysics Data System (ADS)

    Jin, Wanqin; Yang, Liming; Yang, Wei; Chen, Bin; Chen, Jie

    2014-12-01

    A novel method for grafting of 2-hydroxyethyl methacrylate (HEMA) onto the surface of stainless steel (SS) was explored by using 60Co-γ irradiation. The surface of SS was modified by coating of dopamine before radiation grafting. The grafting reaction was performed in a simultaneous irradiation condition. The chemical structures change of the surface before and after grafting was demonstrated by Fourier transform infrared (FTIR) spectrometer. The hydrophilicity of the samples was determined by water contact angle measurement in the comparison of the stainless steel in the conditions of pristine, dopamine coated and HEMA grafted. Surface morphology of the samples was characterized by atomic force microscope (AFM) and scanning electron microscope (SEM). The corrosion resistance properties of the samples were evaluated by Tafel polarization curve. The hemocompatibility of the samples were tested by platelet adhesion assay.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  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. Strain hardening and plastic instability properties of austenitic stainless steels after proton and neutron irradiation

    NASA Astrophysics Data System (ADS)

    Byun, T. S.; Farrell, K.; Lee, E. H.; Hunn, J. D.; Mansur, L. K.

    2001-10-01

    Strain hardening and plastic instability properties were analyzed for EC316LN, HTUPS316, and AL6XN austenitic stainless steels after combined 800 MeV proton and spallation neutron irradiation to doses up to 10.7 dpa. The steels retained good strain-hardening rates after irradiation, which resulted in significant uniform strains. It was found that the instability stress, the stress at the onset of necking, had little dependence on the irradiation dose. Tensile fracture stress and strain were calculated from the stress-strain curve data and were used to estimate fracture toughness using an existing model. The doses to plastic instability and fracture, the accumulated doses at which the yield stress reaches instability stress or fracture stress, were predicted by extrapolation of the yield stress, instability stress, and fracture stress to higher dose. The EC316LN alloy required the highest doses for plastic instability and fracture. Plastic deformation mechanisms are discussed in relation to the strain-hardening properties of the austenitic stainless steels.

  1. 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.

  2. 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.

  3. Estimation of mechanical properties of irradiated nuclear pressure vessel steel by use of subsized CT specimen and small punch specimen

    SciTech Connect

    Mao, X. . Dept. of Mechanical Engineering); Takahashi, H. ); Kodaira, T. )

    1991-11-01

    This paper reports on the 2-1/4 Cr-1M{sub 0} steel that has been selected as the material for the reactor pressure vessel (RPV) of a multipurpose experimental high temperature gas cooled reactor designed by JAERI. The 2-1/4 Cr-1M{sub 0} steel has successful records for high temperature pressure vessels in the petrochemical industries and the ASME Code Case authorizes the use of the steel in these pressure vessels. However, the steel has not been used to nuclear reactor pressure vessels so far. Since the material in the so-called belt line region of the nuclear pressure vessels undergo changes in toughness and strength due to neutron irradiation, it is quite urgent to collect the fracture toughness and strength data of the irradiated steel for the evaluation of the structural intergravity of the reactor pressure vessel of high radiation resistance. In order to study irradiation damage of 2-1/4 Cr-1M{sub 0} steel, small specimens are required because of the severe limitations on specimen size in irradiated-material testing facilities (e.g. the limited space available for testing in nuclear reactors and the narrow damage zone produced by charged particle accelerators). In order to obtain more information about fracture properties of the 2-1/4 Cr- 1M{sub 0} steel from specimens, a subsized compact tensile (CT) specimen, a small punch (SP) specimen and tensile specimen of the irradiated 2-1/4 Cr-1M{sub 0} steel were used to provide radiation effects on fracture toughness, yield strength and ultimate strength. The small punch test, which has been developed recently provides information of the yield and ultimate strength as well as fracture toughness. This report describes the behavior of the neutron irradiation embrittlement of the nuclear reactor pressure vessel steel 2-1/4 Cr-1M{sub 0} by use of new testing approach - subsized specimen techniques.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Tensile properties of ferritic/martensitic steels irradiated in STIP-I

    NASA Astrophysics Data System (ADS)

    Dai, Y.; Long, B.; Tong, Z. F.

    2008-06-01

    Specimens of ferritic/martensitic (FM) steels T91, F82H, Optimax-A and the electron beam weld (EBW) of F82H were irradiated in the Swiss spallation neutron source (SINQ) Target-3 in a temperature range of 90-370 °C to displacement doses between 3 and 12 dpa. Tensile tests were performed at room temperature and the irradiation temperatures. The tensile test results demonstrated that the irradiation hardening increased with dose up to about 10 dpa. Meanwhile, the uniform elongation decreased to less than 1%, while the total elongation remained greater than 5%, except for an F82H specimen of 9.8 dpa tested at room temperature, which failed in elastic deformation regime. At higher doses of 11-12 dpa, the ductility of some specimens recovered, which could be due to the annealing effect of a short period of high temperature excursion. The results do not show significant differences in tensile properties for the different FM steels in the present irradiation conditions.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

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

    DOE PAGES

    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

  16. 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.

  17. The effects of oxide evolution on mechanical properties in proton- and neutron-irradiated Fe-9%Cr ODS steel

    NASA Astrophysics Data System (ADS)

    Swenson, M. J.; Dolph, C. K.; Wharry, J. P.

    2016-10-01

    The objective of this study is to evaluate the effect of irradiation on the strengthening mechanisms of a model Fe-9%Cr oxide dispersion strengthened steel. The alloy was irradiated with protons or neutrons to a dose of 3 displacements per atoms at 500 °C. Nanoindentation was used to measure strengthening due to irradiation, with neutron irradiation causing a greater increase in yield strength than proton irradiation. The irradiated microstructures were characterized using transmission electron microscopy and atom probe tomography (APT). Cluster analysis reveals solute migration from the Y-Ti-O-rich nanoclusters to the surrounding matrix after both irradiations, though the effect is more pronounced in the neutron-irradiated specimen. Because the dissolved oxygen atoms occupy interstitial sites in the iron matrix, they contribute significantly to solid solution strengthening. The dispersed barrier hardening model relates microstructure evolution to the change in yield strength, but is only accurate if solid solution contributions to strengthening are considered simultaneously.

  18. Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

    SciTech Connect

    Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E.

    1995-07-01

    The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab.

  19. UV irradiation enhances the bonding strength between citric acid-crosslinked gelatin and stainless steel.

    PubMed

    Inoue, Motoki; Sasaki, Makoto; Katada, Yasuyuki; Taguchi, Tetsushi

    2011-11-01

    The effect of ultraviolet ray (UV) irradiation on the bonding strength between low carbon stainless steel 316 (SUS316L) and trisuccinimidyl citrate (TSC)-crosslinked alkali-treated gelatin (AlGelatin-TSC) was investigated. The UV irradiation effectively generated hydroxyl groups on the surface of SUS316L. The bonding strength between AlGelatin-TSC and SUS316L before UV irradiation was 0.345±0.007 MPa, and upon UV irradiation it increased to 0.750±0.069 MPa. In order to explain this enhanced bonding strength, the surface of SUS316L was examined using its water contact angle and X-ray photoelectron spectroscopy. Furthermore, the N 1s peaks derived from the TSC succinimidyl group were assigned to the surface of SUS316L after the immobilization of the TSC. This indicates that ester bond formation between the TSC active esters and the SUS316L hydroxyl groups contributed to the enhanced bonding strength. Therefore, UV irradiation and subsequent TSC immobilization is a simple way to functionalize biometal surfaces with various structures. This has practical applications for medical devices such as drug-eluting stents, dental implants, and metallic artificial bone.

  20. 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.

  1. 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.

  2. Void swelling in high dose ion-irradiated reduced activation ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Wang, Xu; Monterrosa, Anthony M.; Zhang, Feifei; Huang, Hao; Yan, Qingzhi; Jiao, Zhijie; Was, Gary S.; Wang, Lumin

    2015-07-01

    To determine the void swelling resistance of reduced-activation ferritic-martensitic steels CNS I and CNS II at high doses, ion irradiation was performed up to 188 dpa (4.6 × 1017 ion/cm2) at 460 °C using 5 MeV Fe++ ions. Helium was pre-implanted at levels of 10 and 100 appm at room temperature to investigate the role of helium on void swelling. Commercial FM steel T91 was also irradiated in this condition and the swelling results are of included in this paper as a reference. Voids were observed in all conditions. The 9Cr CNS I samples implanted with 10 appm helium exhibited lower swelling than 9Cr T91 irradiated at the same condition. The 12Cr CNS II with 10 and 100 appm helium showed significantly lower swelling than CNS I and T91. The swelling rate for CNS I and CNS II were determined to be 0.02%/dpa and 0.003%/dpa respectively. Increasing the helium content from 10 to 100 appm shortened the incubation region and increased the void density but had no effect on the swelling rates.

  3. 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.

  4. 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

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Helium effects on mechanical properties and microstructure of high fluence ion-irradiated RAFM steel

    NASA Astrophysics Data System (ADS)

    Ogiwara, H.; Kohyama, A.; Tanigawa, H.; Sakasegawa, H.

    2007-08-01

    Reduced-activation ferritic/martensitic steels, RAFS, are leading candidates for the blanket and first wall of fusion reactors, and effects of displacement damage and helium production on mechanical properties and microstructures are important to these applications. Because it is the most effective way to obtain systematic and accurate information about microstructural response under fusion environment, single-(Fe 3+) and dual-(Fe 3+ + He +) irradiations were performed followed by TEM observation and nano-indentation hardness measurement. Dual-ion irradiation at 420 °C induced finer defect clusters compared to single-ion irradiation. These fine defect clusters caused large differences in the hardness increase between these irradiations. TEM analysis clarified that radiation induced precipitates were MX precipitates (M: Ta, W). Small defects invisible to TEM possibly caused the large increase in hardness, in addition to the hardness increment produced by radiation induced MX. In this work, radiation hardening and microstructural evolution accompanied by the synergistic effects to high fluences are discussed.

  19. 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.

  20. 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.

  1. Irradiation creep and microstructural changes in an advanced ODS ferritic steel during helium implantation under stress

    NASA Astrophysics Data System (ADS)

    Chen, J.; Pouchon, M. A.; Kimura, A.; Jung, P.; Hoffelner, W.

    2009-04-01

    An advanced oxide dispersion strengthened (ODS) ferritic steel with very fine oxide particles has been homogeneously implanted with helium under uniaxial tensile stresses from 20 to 250 MPa to a maximum dose of about 0.38 dpa (1650 appm-He) with displacement damage rates of 4.4 × 10 -6 dpa/s at temperatures of 573 and 773 K. The samples were in the form of miniaturized dog-bones, where during the helium implantation the straining and the electrical resistance were monitored simultaneously. Creep compliances were measured to be 4.0 × 10 -6 and 11 × 10 -6 dpa -1 MPa -1 at 573 and 773 K, respectively. The resistivity of ODS steel samples decreased with dose, indicating segregation and/or precipitation. Evolution of microstructure during helium implantation was studied in detail by TEM. The effects of ODS particle size on irradiation creep and microstructural changes was investigated by comparing the results from the present advanced ODS (K1) to a commercial ODS ferritic steels (PM2000) with much bigger oxide particles.

  2. 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.

  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. 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.

  5. 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.

  6. Concomitant formation of different nature clusters and hardening in reactor pressure vessel steels irradiated by heavy ions

    NASA Astrophysics Data System (ADS)

    Fujii, K.; Fukuya, K.; Hojo, T.

    2013-11-01

    Specimens of A533B steels containing 0.04, 0.09 and 0.21 wt%Cu were irradiated at 290 °C to 3 dpa with 3 MeV Fe ions and subjected to atom probe analyses, transmission electron microscopy observations and hardness measurements. The atom probe analysis results showed that two types of solute clusters were formed: Cu-enriched clusters containing Mn, Ni and Si atoms as irradiation-enhanced solute atom clusters and Mn/Ni/Si-enriched clusters as irradiation-induced solute atom clusters. Both cluster types occurred in the highest Cu-content steel and the ratio of Mn/Ni/Si-enriched clusters to Cu-enriched clusters increased with irradiation doses. It was confirmed that the cluster formation was a key factor in the microstructure evolution until the high dose irradiation was reached even in the low Cu content steels though the dislocation loops with much lower density than that of the clusters were observed as matrix damage. The difference in the hardening efficiency due to the difference in the nature of the clusters was small. The irradiation-induced clustering of undersized Si atoms suggested that a clustering driving force other than vacancy-driven diffusion, probably an interstitial mechanism, may become important at higher dose rates.

  7. 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

  8. 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

  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. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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

  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. Evolution of the mechanical properties and microstructure of ferritic-martensitic steels irradiated in the BOR-60 reactor

    NASA Astrophysics Data System (ADS)

    Shamardin, V. K.; Golovanov, V. N.; Bulanova, T. M.; Povstyanko, A. V.; Fedoseev, A. E.; Ostrovsky, Z. E.; Goncharenko, Yu. D.

    2002-12-01

    The effect of neutron irradiation on mechanical properties of low-activation ferritic-martensitic (FM) steels 0.1C-9Cr-1W, V, Ta, B and 0.1C-12Cr-2W, V, Ti, B is studied under tension at temperatures of 330-540 °C and doses of 50 dpa. Steel 0.1C-13Cr-Mo, V, Nb, B was chosen for comparison. At irradiation temperatures of 330-340 °C, the radiation hardening of steel with 9%Cr achieves saturation at a dose of 10 dpa. In this case as compared to steels with 12%Cr, the fracture surface is characterized as ductile without cleavage traces. At irradiation temperatures higher than 420 °C, there is no difference in the behavior of the materials under investigation. The data on radiation creep obtained by direct measurement and from the profilometry data satisfy a model ɛ¯/ σ¯=B 0+D Ṡ, when B0 and D have the values typical for steels of FM type.

  17. On the (in)adequacy of the Charpy impact test to monitor irradiation effects of ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Chaouadi, R.

    2007-02-01

    Irradiation embrittlement studies rely very often on Charpy impact data, in particular the ductile-to-brittle transition temperature (DBTT). However, while the DBTT-shift is equivalent to the increase of the fracture toughness transition temperature of ferritic steels, it is not the case for ferritic/martensitic steels. The aim of this study is to critically assess experimental data obtained on a 9%Cr-ferritic/martensitic steel, Eurofer-97, to better understand the underlying mechanisms involved during the fracture process. More specifically, a dedicated analysis using the load diagram approach allows to unambiguously reveal the actual effects of irradiation on physically rather than empirically based parameters. A comparison is made between a ferritic and ferritic/martensitic steel to better identify the possible similarities and differences. Tensile, Charpy impact and fracture toughness tests data are examined in a global approach to assess the actual rather than apparent irradiation effects. The adequacy or inadequacy of the Charpy impact test to monitor irradiation effects is extensively discussed.

  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. 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.

  20. Gas bubbles evolution peculiarities in ferritic-martensitic and austenitic steels and alloys under helium-ion irradiation

    NASA Astrophysics Data System (ADS)

    Chernov, I. I.; Kalashnikov, A. N.; Kalin, B. A.; Binyukova, S. Yu

    2003-12-01

    Transmission electron microscopy has been used to investigate the gas bubble evolution in model alloys of the Fe-C system, ferritic-martensitic steels of 13Cr type, nickel and austenitic steels under 40-keV helium-ion irradiation up to a fluence of 5 × 10 20 m -2 at the temperature of 920 K. It was shown that helium-ion irradiation at high temperature resulted in formation of bubbles with a greater size and a smaller density in Fe and ferritic-martensitic steels than those in nickel and austenitic steels. Large gaseous bubbles in ferritic component are uniformly distributed in grains body in Fe-C alloys as well as in ferritic-martensitic steels. The bubbles with a higher density and a smaller size than those in ferritic component are formed in martensitic grains of steels and Fe-C alloys with a high carbon content ( NC>0.01 wt%), which leads to a small level of swelling of martensite in comparison with that of ferrite. In addition, the bubbles in martensitic grains have a tendency to ordered distribution.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Irradiation-assisted stress corrosion cracking of austenitic stainless steels: Recent progress and new approaches

    SciTech Connect

    Chung, H.M.; Ruther, W.E.; Sanecki, J.E.; Hins, A.; Zaluzec, N.J.; Kassner, T.F.

    1996-09-01

    Irradiation-assisted stress corrosion cracking (IASCC) of several types of BWR field components fabricated from solution-annealed austenitic stainless steels (SSs), including a core internal weld, were investigated by means of slow-strain-rate test (SSRT), scanning electron microscopy (SEM), Auger electron spectroscopy (AES), and field-emission-gun advanced analytical electron microscopy (FEG-AAEM). Based on the results of the tests and analyses, separate effects of neutron fluence, tensile properties, alloying elements and major impurities identified in the American Society for Testing and Materials (ASTM) specifications, minor impurities, water chemistry, and fabrication-related variables were determined. The results indicate strongly that minor impurities not specified by the ASTM-specifications play important roles, probably through a complex synergism with grain-boundary Cr depletion. These impurities, typically associated with steelmaking and component fabrication processes, are very low or negligible in solubility in steels and are the same impurities that have been known to promote intergranular SCC significantly when they are present in water as ions or soluble compounds. It seems obvious that IASCC is a complex integral problem which involves many variables that are influenced strongly by not only irradiation conditions, water chemistry, and stress but also iron and steelmaking processes, fabrication of the component, and joining and welding. Therefore, for high-stress components in particular, it would be difficult to mitigate IASCC problems at high fluence based on the consideration of water chemistry alone, and other considerations based on material composition and fabrication procedure would be necessary as well.

  7. Characterization of 08Cr16Ni11Mo3 stainless steel irradiated in the BN-350 reactor

    NASA Astrophysics Data System (ADS)

    Maksimkin, O. P.; Tsai, K. V.; Turubarova, L. G.; Doronina, T.; Garner, F. A.

    2004-08-01

    In several recently published studies conducted on a Soviet analog of AISI 321 stainless steel irradiated in either fast reactors or light water reactors, it was shown that the void swelling phenomenon extended to temperatures as low as ˜300° C, when produced by neutron irradiation at dpa rates in the range 10 -7 to 10 -8 dpa/s. Other studies yielded similar results for AISI 316. In the current study a blanket duct assembly from BN-350, constructed from the Soviet analog of AISI 316, also exhibits swelling at dpa rates on the order of 10 -8 dpa/s, with voids seen as low as 281 °C and only 1.3 dpa. It appears that low-temperature swelling at low dpa rates occurs in 300 series stainless steels in general, and during irradiations conducted in either fast or mixed spectrum reactors.

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

    DOE PAGES

    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

  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. 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.

  11. 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.

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

    DOE PAGES

    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

  13. Re-weldability tests of irradiated austenitic stainless steel by a TIG welding method

    NASA Astrophysics Data System (ADS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Kalinin, George

    2000-12-01

    Austenitic stainless steel (SS) is widely used for the in-vessel and ex-vessel components of fusion reactors. In particular, SS316L(N)-IG (IG-ITER Grade) is used for the vacuum vessel (VV), pipe lines, blanket modules, branch pipe lines connecting the module coolant system with the manifold and for the other structures of ITER. One of the most important requirements for the VV and the water cooling branch pipelines is the possibility to repair different defects by welding. Those components which may require re-welding should be studied carefully. The SS re-weldability issue has a large impact on the design of in-vessel components, in particular, the design and efficiency of radiation shielding by the modules. Moreover, re-welded components should operate for the lifetime of the reactor. This paper deals with the study of re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. Tungsten inert gas (TIG) welding was used for re-welding of the SS.

  14. 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.

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

    DOE PAGES

    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

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Stressed capsules of austenitic and martensitic steels irradiated in SINQ Target-4 in contact with liquid lead bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Dai, Y.; Gavillet, D.; Restani, R.

    2008-06-01

    In the MEGAPIE target, the steels used for the proton beam entrance window and other components in the spallation reaction zone suffer not only from the irradiation damage produced by protons and neutrons but also from the corrosion and embrittlement induced by liquid lead-bismuth eutectic (LBE). Although these effects have been separately studied by a number of authors, the synergistic effects of irradiation, LBE corrosion and embrittlement are little understood. This work presents detailed analyses of two stressed capsules made of the austenitic steel EC316LN and the martensitic steel 9Cr2WVTa, which were irradiated in SINQ Target-4 in contact with LBE at calculated temperatures of 315 and 225 °C, respectively. The Electron Probe Microanalysis (EPMA) on the cross-sections of the capsules showed that the stagnant LBE induced only slight corrosion on both capsules and no cracks existed in the wall of the EC316LN capsule. Some cracks were observed in the electron beam weld (EBW) and its vicinity of the 9Cr2WVTa capsule, which can be attributed to the high stress inside the wall, the hardening of the material induced by either welding (without re-tempering) or irradiation, and the effects of LBE embrittlement.

  1. 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.

  2. The effects of 800 MeV proton irradiation on the corrosion of tungsten, tantalum, stainless steel, and gold

    SciTech Connect

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

    1997-12-01

    Real time electrochemical data were acquired for tungsten, tantalum, stainless steel 304L, and gold targets during proton irradiation at the LANSCE Weapons Neutron Research Facility. The goal of this research was to establish a better understanding of the corrosion properties of materials as a function of proton irradiation and gain insight into the mechanism of the observed phenomena. The following electrochemical observations were made during proton irradiation of W, Ta, SS304, and Au: (1) the open circuit potential of all materials increased with increasing proton fluence; (2) the corrosion rate (at the OCP) of W and SS304 increased with increasing proton fluence; (3) the passive dissolution rate for SS304 and Ta decreased with increasing proton fluence; (4) the anodic dissolution rate for W increased with increasing proton fluence; (5) the pitting potential for SS304 increased with proton fluence, which is an indication that the material is less susceptible to pitting attack during irradiation.

  3. In-pile and post-irradiation creep of type 304 stainless steel under different neutron spectra

    NASA Astrophysics Data System (ADS)

    Kurata, Y.; Itabashi, Y.; Mimura, H.; Kikuchi, T.; Amezawa, H.; Shimakawa, S.; Tsuji, H.; Shindo, M.

    2000-12-01

    In addition to post-irradiation creep tests, in-pile creep tests were performed using newly developed technology with in situ measurement under different neutron spectra. The in-pile creep properties of type 304 stainless steel at 550°C appear to depend on neutron spectrum, but a spectral effect on post-irradiation creep properties is not clearly seen. The rupture time of in-pile creep under a high thermal neutron flux condition is the shortest. The order of the rupture time following the high thermal flux condition is post-irradiation creep, in-pile creep with a thermal neutron shield condition and finally creep of unirradiated material, all in increasing order. It is suggested that the acceleration of creep deformation and fracture observed in irradiation creep tests may be related to enhancement of thermal creep in terms of FMD increased under a high thermal neutron flux in addition to increased helium embrittlement.

  4. 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.

  5. Effects of laser irradiation on iron loss reduction for Fe-3%Si grain-oriented silicon steel

    SciTech Connect

    Imafuku, Muneyuki . E-mail: crystal@re.nsc.co.jp; Suzuki, Hiroshi; Akita, Koichi; Iwata, Keiji; Fujikura, Masahiro

    2005-02-01

    The effects of laser irradiation on iron loss reduction for Fe-3%Si grain-oriented silicon steel sheet were investigated. The local tensile residual stress states near the laser irradiated cavity lines were observed by using the new X-ray stress measurement method for a single crystal. Although the higher laser power induced the larger tensile residual stresses, the minimum iron loss was obtained at the medium tensile residual stress conditions of about 100-200 MPa. The increase of Vickers hardness was observed with increasing laser power, which was the mark of the plastic deformations induced by the laser irradiation. The tensile residual stress reduces eddy current loss and the plastic deformation increases hysteresis loss of the material. The total iron loss is determined by the balance of these two effects of laser irradiation.

  6. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Jumel, Stéphanie; Van-Duysen, Jean Claude

    2005-04-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called ';Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, …) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program.

  7. Heavy-Section Steel Irradiation Program. Volume 2, No. 1: Semiannual progress report, October 1990--March 1991

    SciTech Connect

    Corwin, W.R.

    1994-07-01

    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 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. 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 reactor pressure-vessel integrity. The HSSI Program is arranged into nine 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 (LUS) weld, (6) irradiation effects in a commercial LUS weld, (7) microstructural analysis of irradiation, (8) in-service aged material evaluations, and (9) correlation monitor materials. During this period, additional analyses on the effects of precleavage stable ductile tearing on the toughness of high-copper welds 72W and 73W demonstrated that the size effects observed in the transition region are not due to substantial differences in ductile tearing behavior. Possible modifications to irradiated duplex crack-arrest specimens were examined to increase the likelihood of their successful testing. Characterization of a second batch of 72W and 73W welds was begun and results of the Charpy V-notch testing is provided. A review of literature on the annealing response of reactor pressure vessel steels was initiated.

  8. Irradiation performance of 9--12 Cr ferritic/martensitic stainless steels and their potential for in-core application in LWRs

    SciTech Connect

    Jones, R.H.; Gelles, D.S.

    1993-08-01

    Ferritic-martensitic stainless steels exhibit radiation stability and stress corrosion resistance that make them attractive replacement materials for austenitic stainless steels for in-core applications. Recent radiation studies have demonstrated that 9% Cr ferritic/martensitic stainless steel had less than a 30C shift in ductile-to-brittle transition temperature (DBTT) following irradiation at 365C to a dose of 14 dpa. These steels also exhibit very low swelling rates, a result of the microstructural stability of these alloys during radiation. The 9 to 12% Cr alloys to also exhibit excellent corrosion and stress corrosion resistance in out-of-core applications. Demonstration of the applicability of ferritic/martensitic stainless steels for in-core LWR application will require verification of the irradiation assisted stress corrosion cracking behavior, measurement of DBTT following irradiation at 288C, and corrosion rates measurements for in-core water chemistry.

  9. Atom Probe Tomography Characterization of the Solute Distributions in a Neutron-Irradiated and Annealed Pressure Vessel Steel Weld

    SciTech Connect

    Miller, M.K.

    2001-01-30

    A combined atom probe tomography and atom probe field ion microscopy study has been performed on a submerged arc weld irradiated to high fluence in the Heavy-Section Steel irradiation (HSSI) fifth irradiation series (Weld 73W). The composition of this weld is Fe - 0.27 at. % Cu, 1.58% Mn, 0.57% Ni, 0.34% MO, 0.27% Cr, 0.58% Si, 0.003% V, 0.45% C, 0.009% P, and 0.009% S. The material was examined after five conditions: after a typical stress relief treatment of 40 h at 607 C, after neutron irradiation to a fluence of 2 x 10{sup 23} n m{sup {minus}2} (E > 1 MeV), and after irradiation and isothermal anneals of 0.5, 1, and 168 h at 454 C. This report describes the matrix composition and the size, composition, and number density of the ultrafine copper-enriched precipitates that formed under neutron irradiation and the change in these parameters with post-irradiation annealing treatments.

  10. Development of radiation damage during in-situ Kr++ irradiation of Fesbnd Nisbnd Cr model austenitic steels

    NASA Astrophysics Data System (ADS)

    Desormeaux, M.; Rouxel, B.; Motta, A. T.; Kirk, M.; Bisor, C.; de Carlan, Y.; Legris, A.

    2016-07-01

    In situ irradiations of 15Cr/15Nisbnd Ti and 15Cr/25Nisbnd Ti model austenitic steels were performed at the Intermediate Voltage Electron Microscope (IVEM)-Tandem user Facility (Argonne National Laboratory) at 600 °C using 1 MeV Kr++. The experiment was designed in the framework of cladding development for the GEN IV Sodium Fast Reactors (SFR). It is an extension of previous high dose irradiations on those model alloys at JANNuS-Saclay facility in France, aimed at investigating swelling mechanisms and microstructure evolution of these alloys under irradiation [1]. These studies showed a strong influence of Ni in decreasing swelling. In situ irradiations were used to continuously follow the microstructure evolution during irradiation using both diffraction contrast imaging and recording of diffraction patterns. Defect analysis, including defect size, density and nature, was performed to characterize the evolving microstructure and the swelling. Comparison of 15Cr/15Nisbnd Ti and 15Cr/25Nisbnd Ti irradiated microstructure has lent insight into the effect of nickel content in the development of radiation damage caused by heavy ion irradiation. The results are quantified and discussed in this paper.

  11. Influence of silicon on swelling and microstructure in Russian austenitic stainless steel EI-847 irradiated to high neutron doses

    NASA Astrophysics Data System (ADS)

    Porollo, S. I.; Shulepin, S. V.; Konobeev, Yu. V.; Garner, F. A.

    2008-08-01

    Void swelling and microstructural development of niobium-stabilized EI-847 austenitic stainless steel with a range of silicon levels were investigated by destructive examination of fuel pin cladding irradiated in three fast reactors located in either Russia or Kazakhstan. The tendency of void swelling to be progressively reduced by increasing silicon concentration appears to be a very general phenomenon in this steel, whether observed in simple, single-variable experiments on well-defined materials or when observed in multivariable, time-dependent irradiations conducted on commercially produced steels over a wide range of irradiation temperatures, neutron spectra and dpa rates. The role of silicon on microstructural development is expressed both in the solid solution via its influence on dislocation and void microstructure and via its influence on formation of radiation-induced phases that in turn alter the matrix composition. Surprisingly, increases in silicon level in this study do not accelerate the formation of silicon-rich G-phase, but act to increase the formation of Nb (C,N) precipitates. Such precipitates are known to be associated with delayed void swelling.

  12. Raman spectroscopic analysis of iron chromium oxide microspheres generated by nanosecond pulsed laser irradiation on stainless steel.

    PubMed

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

    2015-06-15

    Iron chromium oxide microspheres were generated by pulsed laser irradiation on the surface of two commercial samples of stainless steel at room temperature. An Ytterbium pulsed fiber laser was used for this purpose. Raman spectroscopy was used for the characterization of the microspheres, whose size was found to be about 0.2-1.7 μm, as revealed by SEM analysis. The laser irradiation on the surface of the stainless steel modified the composition of the microspheres generated, affecting the concentration of the main elemental components when laser power was increased. Furthermore, the peak ratio of the main bands in the Raman spectra has been associated to the concentration percentage of the main components of the samples, as revealed by Energy-Dispersive X-ray Spectroscopy (EDS) analysis. These experiments showed that it is possible to generate iron chromium oxide microspheres on stainless steel by laser irradiation and that the concentration percentage of their main components is associated with the laser power applied.

  13. Effects of thermal aging and neutron irradiation on the mechanical properties of stainless steel weld overlay cladding

    SciTech Connect

    Haggag, F.M.; Nanstad, R.K.

    1991-01-01

    Stainless steel weld overlay cladding was fabricated using the three-wire, series-arc method. Three layers of cladding were applied to a pressure vessel plate to provide adequate thickness for fabrication of test specimens. Since irradiation of the stainless steel cladding to 5 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV) was conducted at 288{degrees}C for 1605 h, tensile, Charpy V-notch (CVN), precracked Charpy V-notch (PCVN), and compact fracture toughness specimens were thermally aged at 288{degrees}C for 1605 h. Additional specimens are being aged to 20,000 and 50,000 h. Thermal aging of three-wire, series-arc stainless steel weld overlay cladding at 288{degrees}C for 1604 h resulted in appreciable decrease (16%) in the CVN upper-shelf energy, 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 {times} 10{sup 19} neutrons/cm{sup 2} (>MeV), was a 22% reduction in the CVN upper-shelf energy and a 29{degrees}C shift at the 41-J level. The effect of thermal aging on tensile properties was very small or negligible. However, the combined effect after neutron irradiation was an increase in the yield strength (6 to 34% at test temperatures from 288 to {minus}125{degrees}C) and no apparent change in ultimate strength and total elongation. Also, neutron irradiation reduced the initiation fracture toughness (J{sub Ic}) much more than did thermal aging. 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 specimen become available.

  14. Effects of thermal aging and neutron irradiation on the mechanical properties of stainless steel weld overlay cladding

    SciTech Connect

    Haggag, F.M.; Nanstad, R.K.

    1991-12-31

    Stainless steel weld overlay cladding was fabricated using the three-wire, series-arc method. Three layers of cladding were applied to a pressure vessel plate to provide adequate thickness for fabrication of test specimens. Since irradiation of the stainless steel cladding to 5 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV) was conducted at 288{degrees}C for 1605 h, tensile, Charpy V-notch (CVN), precracked Charpy V-notch (PCVN), and compact fracture toughness specimens were thermally aged at 288{degrees}C for 1605 h. Additional specimens are being aged to 20,000 and 50,000 h. Thermal aging of three-wire, series-arc stainless steel weld overlay cladding at 288{degrees}C for 1604 h resulted in appreciable decrease (16%) in the CVN upper-shelf energy, 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 {times} 10{sup 19} neutrons/cm{sup 2} (>MeV), was a 22% reduction in the CVN upper-shelf energy and a 29{degrees}C shift at the 41-J level. The effect of thermal aging on tensile properties was very small or negligible. However, the combined effect after neutron irradiation was an increase in the yield strength (6 to 34% at test temperatures from 288 to {minus}125{degrees}C) and no apparent change in ultimate strength and total elongation. Also, neutron irradiation reduced the initiation fracture toughness (J{sub Ic}) much more than did thermal aging. 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 specimen become available.

  15. Irradiation effects on 17-7 PH stainless steel, A-201 carbon steel, and titanium-6-percent-aluminum-4-percent-vanadium alloy

    NASA Technical Reports Server (NTRS)

    Hasse, R. A.; Hartley, C. B.

    1972-01-01

    Irradiation effects on three materials from the NASA Plum Brook Reactor Surveillance Program were determined. An increase of 105 K in the nil-ductility temperature for A-201 steel was observed at a fluence of approximately 3.1 x 10 to the 18th power neutrons/sq cm (neutron energy E sub n greater than 1.0 MeV). Only minor changes in the mechanical properties of 17-7 PH stainless steel were observed up to a fluence of 2 x 10 to the 21st power neutrons/sq cm (E sub n greater than 1.0 MeV). The titanium-6-percent-aluminum-4-percent-vanadium alloy maintained its notch toughness up to a fluence of 1 x 10 to the 21st power neutrons/sq cm (E sub n greater than 1.0 MeV).

  16. Integrated analysis of millisecond laser irradiation of steel by comprehensive optical diagnostics and numerical simulation

    NASA Astrophysics Data System (ADS)

    Doubenskaia, M.; Smurov, I.; Nagulin, K. Yu.

    2016-04-01

    Complimentary optical diagnostic tools are applied to provide comprehensive analysis of thermal phenomena in millisecond Nd:YAG laser irradiation of steel substrates. The following optical devices are employed: (a) infrared camera FLIR Phoenix RDASTM equipped by InSb sensor with 3 to 5 µm band pass arranged on 320 × 256 pixels array, (b) ultra-rapid camera Phantom V7.1 with SR-CMOS monochrome sensor in the visible spectral range, up to 105 frames per second for 64 × 88 pixels array, (c) original multi-wavelength pyrometer in the near-infrared range (1.370-1.531 µm). The following laser radiation parameters are applied: variation of energy per pulse in the range 15-30 J at a constant pulse duration of 10 ms with and without application of protective gas (Ar). The evolution of true temperature is restored based on the method of multi-colour pyrometry; by this way, melting/solidification dynamics is analysed. Emissivity variation with temperature is studied, and hysteresis type functional dependence is found. Variation of intensity of surface evaporation visualised by the camera Phantom V7.1 is registered and linked with the surface temperature evolution, different surface roughness and influence of protective gas atmosphere. Determination of the vapour plume temperature based on relatively intensities of spectral lines is done. The numerical simulation is carried out applying the thermal model with phase transitions taken into account.

  17. Phase diffusionless γ↔α transformations and their effect on physical, mechanical and corrosion properties of austenitic stainless steels irradiated with neutrons and charged particles

    NASA Astrophysics Data System (ADS)

    Maksimkin, O. P.

    2016-04-01

    The work presents relationships of γ→α' and α'→γ-transformations in reactor 12Cr18Ni10Ti and 08Cr16Ni11Mo3 austenitic stainless steels induced by cold work, irradiation and/or temperature. Energy and mechanical parameters of nucleation and development of deformation-induced martensitic α'-phase in the non-irradiated and irradiated steels are given. The mechanisms of localized static deformation were investigated and its effect on martensitic γ→α' transformation is determined. It has been shown that irradiation of 12Cr18Ni10Ti steel with heavy Kr ions (1.56MeV/nucleon, fluence of 1·1015 cm-2) results in formation of α'-martensite in near-surface layer of the sample. Results of systematic research on reversed α'→γ-transformation in austenitic metastable stainless steels irradiated with slow (VVR-K) and fast (BN-350) neutrons are presented. The effect of annealing on strength and magnetic characteristics was determined. It was found that at the temperature of 400 °C in the irradiated with neutrons samples (59 dpa) an increase of ferromagnetic α'-phase and microhardness was observed. The obtained results could be used during assessment of operational characteristics of highly irradiated austenitic steels during transportation and storage of Fuel Assemblies for fast nuclear reactors.

  18. Stability of nanoclusters in 14YWT oxide dispersion strengthened steel under heavy ion-irradiation by atom probe tomography

    SciTech Connect

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

    2014-12-01

    14YWT oxide dispersion strengthened (ODS) ferritic steel was irradiated with of 5 MeV Ni2+ ions, at 300 C, 450 C, and 600 C to a damage level of 100 dpa. The stability of Ti–Y–O nanoclusters was investigated by applying atom probe tomography (APT) in voltage mode, of the samples before and after irradiations. The average size and number density of the nanoclusters was determined using the maximum separation method. These techniques allowed for the imaging of nanoclusters to sizes well below the resolution limit of conventional transmission electron microscopy techniques. The most significant changes were observed for samples irradiated at 300 C where the size (average Guinier radius) and number density of nanoclusters were observed to decrease from 1.1 nm to 0.8 nm and 12 1023 to 3.6 1023, respectively. In this study, the nanoclusters are more stable at higher temperature.

  19. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: The case of Fe-Cu model alloys

    NASA Astrophysics Data System (ADS)

    Subbotin, A. V.; Panyukov, S. V.

    2016-08-01

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. We show that the obtained results are in good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  20. Study of irradiated Hadfield steel using transmission Mössbauer spectroscopy with high velocity resolution and conversion electron Mössbauer spectroscopy

    NASA Astrophysics Data System (ADS)

    Semionkin, V. A.; Neshev, F. G.; Tsurin, V. A.; Milder, O. B.; Oshtrakh, M. I.

    2010-03-01

    Proton irradiated Hadfield steel foil was studied using transmission Mössbauer spectroscopy with high velocity resolution and conversion electron Mössbauer spectroscopy. It was shown that proton irradiation leads to structural changes in the foil as well as to surface oxidation with ferric hydrous oxide formation (ferrihydrite). Moreover, oxidation on the foil underside was higher than on the foil right side.

  1. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Chiapetto, M.; Becquart, C. S.; Domain, C.; Malerba, L.

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a "grey alloy" approach that extends the already existing OKMC model for neutron irradiated Fe-C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe-C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  2. Irradiation creep and stress-enhanced swelling of Fe-16Cr-15Ni-Nb austenitic stainless steel in BN-350

    SciTech Connect

    Vorobjev, A.N.; Porollo, S.I.; Konobeev, Yu.V.

    1997-04-01

    Irradiation creep and void swelling will be important damage processes for stainless steels when subjected to fusion neutron irradiation at elevated temperatures. The absence of an irradiation device with fusion-relevant neutron spectra requires that data on these processes be collected in surrogate devices such as fast reactors. This paper presents the response of an annealed austenitic steel when exposed to 60 dpa at 480{degrees}C and to 20 dpa at 520{degrees}C. This material was irradiated as thin-walled argon-pressurized tubes in the BN-350 reactor located in Kazakhstan. These tubes were irradiated at hoop stresses ranging from 0 to 200 MPa. After irradiation both destructive and non-destructive examination was conducted.

  3. Relationship Between Grain Boundary Structure and Radiation Induced Segregation in a Neutron Irradiated 9 wt. % Cr Model Ferritic/Martensitic Steel

    SciTech Connect

    Field, Kevin G; Miller, Brandon; Chichester, Heather J.M.; Sridharan, K.; Allen, Todd R.

    2014-01-01

    Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs but has only been demonstrated in ion irradiated specimens. A 9 wt. % Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of neutron radiation environment on the RIS-GB structure dependence. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.

  4. Relationship between lath boundary structure and radiation induced segregation in a neutron irradiated 9 wt.% Cr model ferritic/martensitic steel

    SciTech Connect

    Field, Kevin G.; Miller, Brandon D.; Chichester, Heather J. M.; Sridharan, Kumar; Allen, Todd R.

    2014-02-01

    Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes which migrate to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs where low energy structures have suppressed RIS responses. This relationship between local GB structure and RIS has been demonstrated primarily in ion-irradiated specimens. A 9 wt.% Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of a neutron radiation environment on the RIS response at different GB structures. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.

  5. Accumulation and annealing of radiation defects under low-temperature electron and neutron irradiation of ODS steel and Fe-Cr alloys

    NASA Astrophysics Data System (ADS)

    Arbuzov, V. L.; Goshchitskii, B. N.; Sagaradze, V. V.; Danilov, S. E.; Kar'kin, A. E.

    2010-10-01

    The processes of accumulation and annealing of radiation defects at low-temperature (77 K) electron and neutron irradiation and their effect on the physicomechanical properties of Fe-Cr alloys and oxide dispersion strengthened (ODS) steel have been studied. It has been shown that the behavior of radiation defects in ODS steel and Fe-Cr alloys is qualitatively similar. Above 250 K, radiation-induced processes of the solid solution decomposition become conspicuous. These processes are much less pronounced in ODS steel because of specific features of its microstructure. Processes related to the overlapping of displacement cascades under neutron irradiation have been considered. It has been shown that, in this case, it is the increase in the size of vacancy clusters, rather than the growth of their concentration, that is prevailing. Possible mechanisms of the radiation hardening of the ODS steel and the Fe-13Cr alloy upon irradiation and subsequent annealing have been discussed.

  6. Modeling of microstructure evolution in austenitic stainless steels irradiated under light water reactor condition

    NASA Astrophysics Data System (ADS)

    Gan, J.; Was, G. S.; Stoller, R. E.

    2001-10-01

    A model for microstructure development in austenitic alloys under light water reactor irradiation conditions is described. The model is derived from the model developed by Stoller and Odette to describe microstructural evolution under fast neutron or fusion reactor irradiation conditions. The model is benchmarked against microstructure measurements in 304 and 316 SS irradiated in a boiling water reactor core using one material-dependent and three irradiation-based parameters. The model is also adapted for proton irradiation at higher dose rate and higher temperature and is calibrated against microstructure measurements for proton irradiation. The model calculations show that for both neutron and proton irradiations, in-cascade interstitial clustering is the driving mechanism for loop nucleation. The loss of interstitial clusters to sinks by interstitial cluster diffusion was found to be an important factor in determining the loop density. The model also explains how proton irradiation can produce an irradiated dislocation microstructure similar to that in neutron irradiation.

  7. Production and segregation of transmutation elements Ca, Ti, Sc in the F82H steel under mixed spectrum irradiation of high energy protons and spallation neutrons

    NASA Astrophysics Data System (ADS)

    Kuksenko, Viacheslav; Pareige, Cristelle; Pareige, Philippe; Dai, Yong

    2014-04-01

    Ferritic/martensitic steel F82H was irradiated at 345 °C in a mixed proton-neutron spectrum in the Swiss spallation neutron source up to 20.3 dpa. Nanoscale investigations using the atom probe tomography (APT) technique were performed in order to study the atomic scale evolution of the microstructure of the F82H steel under irradiation. Spallation products Ca, Ti and Sc have been detected. The irradiation led to the production of about 370 appm of Ca, 90 appm of Sc and 800 appm of Ti. APT experiments revealed that regardless their low bulk concentrations, the spallation products extensively participate in the evolution of the microstructure: formation of radiation-induced clusters, segregation at the dislocation loops and alteration of the microchemistry of carbides. In this paper, a quantitative description of the observed features is presented and results are compared with TEM data of the literature obtained on the same steel and under similar irradiation conditions.

  8. Temper and neutron irradiation embrittlement in 2 1/4 Cr-1 Mo steels for pressure vessels of high-temperature gas-cooled reactors

    SciTech Connect

    Suzuki, M.; Fukaya, K.; Kodaira, T.; Oku, T.

    1984-09-01

    A 2 1/4 Cr-1 Mo steel is a promising candidate material for structural components of the pressure vessel of the experimental very high temperature gascooled reactor (VHTR) in Japan. Since the service temperature of such components is expected to be about 400/sup 0/C, the behavior of the temper and neutron irradiation embrittlements in these chromium-molybdenum steels should be confirmed from the viewpoint of structural integrity. The experimental verification on the degree of the embrittlement due to thermal aging, including the effect of applied stress and neutron irradiation, is described. Steel containing substantial amounts (about 100 ppm) of phosphorus atoms, which are believed to cause the temper embrittlement, showed that applied stress enhanced the embrittlement due to thermal aging. Embrittlement caused by neutron irradiation appears to be minimal in the case of the material containing <1000 ppm of copper as impurity with neutrons irradiated at about 400/sup 0/C.

  9. Void Swelling Of Aisi 321 Analog Stainless Steel Irradiated At Low Dpa Rates In The Bn-350 Reactor

    SciTech Connect

    Maksimkin, O. P.; Tsai, K. V.; Turubarova, L. G.; Doronina, T. A.; Garner, Francis A.

    2006-03-01

    In several recently published studies conducted on a Soviet analog of AISI 321 stainless steel irradiated in either fast reactors or light water reactors, it was shown that the void swelling phenomenon extended to temperatures as low as ~300ºC or less, when produced by neutron irradiation at dpa rates in the range 10-7 to 10-8 dpa/sec. Other studies yielded similar results for AISI 316 and the Russian analog of AISI 316. In the current study a blanket duct assembly from BN-350, constructed from the Soviet analog of AISI 321, also exhibits swelling at dpa rates on the order of 10-8 dpa/sec, with voids seen as low as 281oC and only 0.65 dpa. It appears that low-temperature swelling occurs at low dpa rates in 300 series stainless steels in general, and also occurs during irradiations conducted in either fast or mixed spectrum reactors. Therefore it is expected that a similar behavior will be observed in fusion devices as well.

  10. Void Swelling of AISI 321 Analog Stainless Steel Irradiated at Low DPA Rates in the BN-350 Reactor

    SciTech Connect

    Maksimkin, O. P.; Tsai, K. V.; Turubarova, L. G.; Doronina, T. A.; Garner, Francis A.

    2007-08-01

    In several recently published studies conducted on a Soviet analog of AISI 321 stainless steel irradiated in either fast reactors or light water reactors, it was shown that the void swelling phenomenon extended to temperatures as low as ~300ºC or less, when produced by neutron irradiation at dpa rates in the range 10-7 to 10-8 dpa/sec. Other studies yielded similar results for AISI 316 and the Russian analog of AISI 316. In the current study a blanket duct assembly from BN-350, constructed from the Soviet analog of AISI 321, also exhibits swelling at dpa rates on the order of 10-8 dpa/sec, with voids seen as low as 281C and only 0.65 dpa. It appears that low-temperature swelling occurs at low dpa rates in 300 series stainless steels in general, and also occurs during irradiations conducted in either fast or in mixed spectrum reactors as shown in other studies.

  11. Comparison of different experimental and analytical measures of the thermal annealing response of neutron-irradiated RPV steels

    SciTech Connect

    Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K.

    1997-05-01

    The thermal annealing response of several materials as indicated by Charpy transition temperature (TT) and upper-shelf energy (USE), crack initiation toughness, K{sub Jc}, predictive models, and automated-ball indentation (ABI) testing are compared. The materials investigated are representative reactor pressure vessel (RPV) steels (several welds and a plate) that were irradiated for other tasks of the Heavy-Section Steel Irradiation (HSSI) Program and are relatively well characterized in the unirradiated and irradiated conditions. They have been annealed at two temperatures, 343 and 454 C (650 and 850 F) for varying lengths of time. The correlation of the Charpy response and the fracture toughness, ABI, and the response predicted by the annealing model of Eason et al. for these conditions and materials appears to be reasonable. The USE after annealing at the temperature of 454 C appears to recover at a faster rate than the TT, and even over-recovers (i.e., the recovered USE exceeds that of the unirradiated material).

  12. Void swelling of AISI 321 analog stainless steel irradiated at low dpa rates in the BN-350 reactor

    NASA Astrophysics Data System (ADS)

    Maksimkin, O. P.; Tsai, K. V.; Turubarova, L. G.; Doronina, T.; Garner, F. A.

    2007-08-01

    In several recently published studies conducted on a Soviet analog of AISI 321 stainless steel irradiated in either fast reactors or light water reactors, it was shown that the void swelling phenomenon extended to temperatures as low as ˜300 °C or less, when produced by neutron irradiation at dpa rates in the range 10 -7-10 -8 dpa/s. Other studies yielded similar results for AISI 316 and the Russian analog of AISI 316. In the current study a blanket assembly duct from BN-350, constructed from the Soviet analog of AISI 321, also exhibits swelling at dpa rates on the order of 10 -8 dpa/s, with voids seen as low as 281 °C and only 0.65 dpa. It appears that low-temperature swelling occurs at low dpa rates in 300 series stainless steels in general, and also occurs during irradiations conducted in either fast or in mixed spectrum reactors as shown in other studies.

  13. Investigation on femto-second laser irradiation assisted shock peening of medium carbon (0.4% C) steel

    NASA Astrophysics Data System (ADS)

    Majumdar, Jyotsna Dutta; Gurevich, Evgeny L.; Kumari, Renu; Ostendorf, Andreas

    2016-02-01

    In the present study, the effect of femtosecond laser irradiation on the peening behavior of 0.4% C steel has been evaluated. Laser irradiation has been conducted with a 100 μJ and 300 fs laser with multiple pulses under varied energy. Followed by laser irradiation, a detailed characterization of the processed zone was undertaken by scanning electron microscopy, and X-ray diffraction technique. Finally, the residual stress distribution, microhardness and wear resistance properties of the processed zone were also evaluated. Laser processing leads to shock peening associated with plasma formation and its expansion, formation of martensite and ferrito-pearlitic phase in the microstructure. Due to laser processing, there is introduction of residual stress on the surface which varies from high tensile (140 MPa) to compressive (-335 MPa) as compared to 152 MPa of the substrate. There is a significant increase in microhardness to 350-500 VHN as compared to 250 VHN of substrate. The fretting wear behavior against hardened steel ball shows a significant reduction in wear depth due to laser processing. Finally, a conclusion of the mechanism of wear has been established.

  14. Influence of cold work level on the irradiation creep and creep rupture of titanium-modified austenitic stainless steels

    SciTech Connect

    Garner, F.A.; Hamilton, M.L. ); Eiholzer, C.R. ); Toloczko, M.B. ); Kumar, A.S. )

    1992-06-01

    A titanium-modified austenitic type stainless steel was tested at three cold work levels to determine its creep and creep rupture properties under both thermal aging and neutron irradiation conditions. Both the thermal and irradiation creep behavior exhibit a complex non-monotonic relationship with cold work level that reflects the competition between a number of stress-sensitive and temperature-dependent microstructural processes. Increasing the degree of cold work to 30% form the conventional 20% level was detrimental to its performance, especially for applications above 550{degrees}C. The 20% cold work level is preferable to the 10% level, in terms of both in- reactor creep rapture response and initial strength.

  15. Irradiation creep and creep rupture of titanium-modified austenitic stainless steels and their dependence on cold work level

    SciTech Connect

    Garner, F.A.; Hamilton, M.L.; Eiholzer, C.R.; Toloczko, M.B.; Kumar, A.S.

    1991-11-01

    A titanium-modified austenitic type stainless steel was tested at three cold work levels to determine its creep and creep rupture properties under both thermal aging and neutron irradiation conditions. Both the thermal and irradiation creep behavior exhibit a complex non-monotonic relationship with cold work level that reflects the competition between a number of stress-sensitive and temperature-dependent microstructural processes. Increasing the degree of cold work to 30% from the conventional 20% level was detrimental to its performance, especially for applications above 550{degrees}c. The 20% cold work level is preferable to the 10% level, in terms of both in-reactor creep rupture response and initial strength.

  16. Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

    NASA Astrophysics Data System (ADS)

    Uwaba, Tomoyuki; Ito, Masahiro; Maeda, Koji

    2011-09-01

    The C3M irradiation test, which was conducted in the experimental fast reactor, "Joyo", demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, "Monju". The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and 137Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  17. The measurement of material degradation in neutron irradiated Mn-Mo-Ni low alloy steels using magnetic techniques

    NASA Astrophysics Data System (ADS)

    Chang, Kee-Ok; Chi, Se-Hwan; Kim, Taek-Soo; Kim, Byoung-Chul; Lee, Sam-Lai

    1999-12-01

    To examine the application of magnetic method to the evaluation of radiation damage and thermal recovery in Mn-Mo-Ni reactor pressure vessel(RPV) steels, changes in the magnetic parameters and Vickers microhardness(Hv) due to neutron irradiation and heat treatment were measured and compared for RPV surveillance specimens which were irradiated to the neutron fluence of 2.4×1019n/cm2 (E⩾1.0 MeV) in a typical pressurized water reactor environment at 288 °C. Results show that the coercivity(Hc) increased due to irradiation, whereas susceptibility and Barkhausen noise energy(BNE) decreased. Saturation magnetization(Ms) remained unchanged while Vickers microhardness increased. For isothermally heat treated specimens, both the magnetic parameters and Vickers microhardness returned to the unirradiated condition: Thus, the BNE and susceptibility increased while the coercivity and Vickers microhardness decreased. From the comparison of irradiated base metal and weld metal, it is found that the susceptibility and BNE of the weld metal were smaller than those of base metal and the microhardness and coercivity of weld metal were larger than those of base metal. All these consistent changes in magnetic parameters show a possibility that magnetic techniques may be used for the evaluation of material degradation.

  18. Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

    This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. One of the temperature profiles was that the specimens experienced neutron irradiation in both reactors, under which the irradiation temperature transiently increased to 290 °C from room temperature with increasing reactor power during reactor startup periods. Another was that the specimens were pre-heated to about 150 °C prior to the irradiation to suppress the transient temperature increase. Tensile tests at 290 °C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Difference of the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. Although influence of neutron irradiation involving transient temperature increase to 290 °C from room temperature on the yield strength and the Vickers hardness is buried in the trend curves of existing data, the influence was also found certainly in increment of in yield strength, existence of modest yield drop, and loss of strain hardening capacity and ductility. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, seemed to have important implications regarding the interpretation of not irradiation hardening, but deformation of the austenitic stainless steel.

  19. New Dosimetric Interpretation of the DV50 Vessel-Steel Experiment Irradiated in the OSIRIS MTR Reactor Using the Monte-Carlo Code TRIPOLI-4®

    NASA Astrophysics Data System (ADS)

    Malouch, Fadhel

    2016-02-01

    An irradiation program DV50 was carried out from 2002 to 2006 in the OSIRIS material testing reactor (CEA-Saclay center) to assess the pressure vessel steel toughness curve for a fast neutron fluence (E > 1 MeV) equivalent to a French 900-MWe PWR lifetime of 50 years. This program allowed the irradiation of 120 specimens out of vessel steel, subdivided in two successive irradiations DV50 n∘1 and DV50 n∘2. To measure the fast neutron fluence (E > 1 MeV) received by specimens after each irradiation, sample holders were equipped with activation foils that were withdrawn at the end of irradiation for activity counting and processing. The fast effective cross-sections used in the dosimeter processing were determined with a specific calculation scheme based on the Monte-Carlo code TRIPOLI-3 (and the nuclear data ENDF/B-VI and IRDF-90). In order to put vessel-steel experiments at the same standard, a new dosimetric interpretation of the DV50 experiment has been performed by using the Monte-Carlo code TRIPOLI-4 and more recent nuclear data (JEFF3.1.1 and IRDF-2002). This paper presents a comparison of previous and recent calculations performed for the DV50 vessel-steel experiment to assess the impact on the dosimetric interpretation.

  20. Extreme embrittlement of austenitic stainless steel irradiated to 75--81 dpa at 335--360 C

    SciTech Connect

    Porollo, S.I.; Vorobjev, A.N.; Konobeev, Yu.V.; Garner, F.A.

    1998-03-01

    This paper presents the results of an experiment conducted in the BN-350 fast reactor in Kazakhstan that involved the irradiation of argon-pressurized thin-walled tubes (0--2000 MPa hoop stress) constructed from Fe-16Cr-15Ni-3Mo-Nb stabilized steel in contact with the sodium coolant, which enters the reactor at {approximately}270 C. Tubes in the annealed condition reached 75 dpa at 335 C, and another set in the 20% cold-worked condition reached 81 dpa at 360 C. Upon disassembly all tubes, except those in the stress-free condition, were found to have failed in an extremely brittle fashion. The stress-free tubes exhibited diameter changes that imply swelling levels ranging from 9 to 16%. It is expected that stress-enhancement of swelling induced even larger swelling levels in the stressed tubes. The embrittlement is explained in terms of the sensitivity of the swelling regime to displacement rate and the large, unprecedented levels of swelling reached at 335--360 C at these high neutron fluences. The failure mechanism appears to be identical to that observed at similar swelling levels in other austenitic steels irradiated in US fast reactors at 400--425 C, whereby stress-concentration between voids and nickel segregation at void surfaces predisposes the steel to an epsilon martensite transformation followed by formation of alpha martensite at crack tips. The very slow strain rate inherent in such creep tests and the relatively high helium levels may also contribute to the failure.

  1. Neutron Exposure Parameters for the Dosimetry Capsule in the Heavy-Section Steel Irradiation Program Tenth Irradiation Series

    SciTech Connect

    C.A. Baldwin; F.B.K. Kam; I. Remec

    1998-10-01

    This report describes the computational methodology for the least-squares adjustment of the dosimetry data from the HSSI 10.OD dosimetry capsule with neutronics calculations. It presents exposure rates at each dosimetry location for the neutron fluence greater than 1.0 MeV, fluence greater than 0.1 MeV, and displacements per atom. Exposure parameter distributions are also described in terms of three- dimensional fitting functions. When fitting functions are used it is suggested that an uncertainty of 6% (1 o) should be associated with the exposure rate values. The specific activity of each dosimeter at the end of irradiation is listed in the Appendix.

  2. Void swelling and precipitation in a titanium-modified austenitic stainless steel under proton irradiation

    NASA Astrophysics Data System (ADS)

    Kimoto, T.; Shiraishi, H.

    1985-07-01

    The correlation between void swelling and precipitation behavior in a 10% cold worked Fe-16.2Ni-14.6Cr-2.37Mo-1.79Mn-0.53Si-0.24Ti-0.06C alloy was examined with 200 keV proton irradiation. Swelling peak temperature after the proton irradiation to 10 dpa was about 823 K, and void swelling decreased steeply with increase in irradiation temperature from 823 to 923 K. Void swelling increased rapidly from 1.9 to 12.1% with increase in irradiation dose from 20 to 45 dpa at 873 K. Fine intragranular TiC precipitates, which were formed during initial stage of irradiation, dissolved gradually with increase in irradiation dose from 10 to 45 dpa at 873 K, while the amount of precipitation of needle-shaped Fe 2P phase containing titanium increased with increasing dose. The reduction of sink strength of the TiC precipitates due to the dissolution during irradiation was thought to cause the increase of swelling rate with increase in irradiation dose from 20 to 45 dpa at 873 K.

  3. The comparison of microstructure and nanocluster evolution in proton and neutron irradiated Fe-9%Cr ODS steel to 3 dpa at 500 °C

    NASA Astrophysics Data System (ADS)

    Swenson, M. J.; Wharry, J. P.

    2015-12-01

    A model Fe-9%Cr oxide dispersion strengthened (ODS) steel was irradiated with protons or neutrons to a dose of 3 displacements per atom (dpa) at a temperature of 500 °C, enabling a direct comparison of ion to neutron irradiation effects at otherwise fixed irradiation conditions. The irradiated microstructures were characterized using transmission electron microscopy and atom probe tomography including cluster analysis. Both proton and neutron irradiations produced a comparable void and dislocation loop microstructure. However, the irradiation response of the Ti-Y-O oxide nanoclusters varied. Oxides remained stable under proton irradiation, but exhibited dissolution and an increase in Y:Ti composition ratio under neutron irradiation. Both proton and neutron irradiation also induced varying extents of Si, Ni, and Mn clustering at existing oxide nanoclusters. Protons are able to reproduce the void and loop microstructure of neutron irradiation carried out to the same dose and temperature. However, since nanocluster evolution is controlled by both diffusion and ballistic impacts, protons are rendered unable to reproduce the nanocluster evolution of neutron irradiation at the same dose and temperature.

  4. Cluster dynamics modeling of the effect of high dose irradiation and helium on the microstructure of austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Brimbal, Daniel; Fournier, Lionel; Barbu, Alain

    2016-01-01

    A mean field cluster dynamics model has been developed in order to study the effect of high dose irradiation and helium on the microstructural evolution of metals. In this model, self-interstitial clusters, stacking-fault tetrahedra and helium-vacancy clusters are taken into account, in a configuration well adapted to austenitic stainless steels. For small helium-vacancy cluster sizes, the densities of each small cluster are calculated. However, for large sizes, only the mean number of helium atoms per cluster size is calculated. This aspect allows us to calculate the evolution of the microstructural features up to high irradiation doses in a few minutes. It is shown that the presence of stacking-fault tetrahedra notably reduces cavity sizes below 400 °C, but they have little influence on the microstructure above this temperature. The binding energies of vacancies to cavities are calculated using a new method essentially based on ab initio data. It is shown that helium has little effect on the cavity microstructure at 300 °C. However, at higher temperatures, even small helium production rates such as those typical of sodium-fast-reactors induce a notable increase in cavity density compared to an irradiation without helium.

  5. Stability Of Nanoclusters In 14YWT Oxide Dispersion Strengthened Steel Under Heavy Ion-irradiation By Atom Probe Tomography

    SciTech Connect

    He, Jianchao; Wan, F.; Sridharan, Kumar; Allen, Todd R.; Certain, Alicia G.; Shutthanandan, V.; Wu, Yaqiao

    2014-12-01

    14YWT oxide dispersion strengthened (ODS) ferritic steel was irradiated with of 5 MeV Ni2+ ions, at 300 °C, 450 °C, and 600 °C to a damage level of 100 dpa. The stability of Ti–Y–O nanoclusters was investigated by applying atom probe tomography (APT) in voltage mode, of the samples before and after irradiations. The average size and number density of the nanoclusters was determined using the maximum separation method. These techniques allowed for the imaging of nanoclusters to sizes well below the resolution limit of conventional transmission electron microscopy techniques. The most significant changes were observed for samples irradiated at 300 °C where the size (average Guinier radius) and number density of nanoclusters were observed to decrease from 1.1 nm to 0.8 nm and 12 × 1023 to 3.6 × 1023, respectively. In this study, the nanoclusters are more stable at higher temperature.

  6. Microstructural characterization and density change of 304 stainless steel reflector blocks after long-term irradiation in EBR-II

    NASA Astrophysics Data System (ADS)

    Huang, Y.; Wiezorek, J. M. K.; Garner, F. A.; Freyer, P. D.; Okita, T.; Sagisaka, M.; Isobe, Y.; Allen, T. R.

    2015-10-01

    While thin reactor structural components such as cladding and ducts do not experience significant gradients in dpa rate, gamma heating rate, temperature or stress, thick components can develop strong local variations in void swelling and irradiation creep in response to gradients in these variables. In this study we conducted microstructural investigations by transmission electron microscopy of two 52 mm thick 304-type stainless steel hex-blocks irradiated for 12 years in the EBR-II reactor with accumulated doses ranging from ∼0.4 to 33 dpa. Spatial variations in the populations of voids, precipitates, Frank loops and dislocation lines have been determined for 304 stainless steel sections exposed to different temperatures, different dpa levels and at different dpa rates, demonstrating the existence of spatial gradients in the resulting void swelling. The microstructural measurements compare very well with complementary density change measurements regarding void swelling gradients in the 304 stainless steel hex-block components. The TEM studies revealed that the original cold-worked-state microstructure of the unirradiated blocks was completely erased by irradiation, replaced by high densities of interstitial Frank loops, voids and carbide precipitates at both the lowest and highest doses. At large dose levels the amount of volumetric void swelling correlated directly with the gamma heating gradient-related temperature increase (e.g. for 28 dpa, ∼2% swelling at 418 °C and ∼2.9% swelling at 448 °C). Under approximately iso-thermal local conditions, volumetric void swelling was found to increase with dose level (e.g. ∼0.2% swelling at 0.4 dpa, ∼0.5% swelling at 4 dpa and ∼2% swelling at 28 dpa). Carbide precipitate formation levels were found to be relatively independent of both dpa level and temperature and induced a measurable densification. Void swelling was dominant at the higher dose levels and caused measurable decreases in density. Void swelling

  7. Evolution of microstructure and mechanical properties of VVER-1000 RPV steels under re-irradiation

    NASA Astrophysics Data System (ADS)

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Erak, D.; Zhurko, D.

    2015-01-01

    This is a comprehensive study of microstructure and mechanical properties evolution at re-irradiation after recovery annealing of VVER-1000 RPV weld and base metals as well as the effect of annealing on the microstructure and properties of base metal in the zone of the temperature gradient that is implemented during annealing using special heating device. It is shown that the level of radiation-induced microstructural changes under accelerated re-irradiation of weld and base metal is not higher than for the primary irradiation. Thus, we can predict that re-embrittlement of VVER-1000 RPV materials considering the flux effect will not exceed the typical embrittlement rate for the primary irradiation.

  8. Ultra high vacuum fracture and transfer device for AES analysis of irradiated austenitic stainless steel

    SciTech Connect

    Urie, M.W.; Panayotou, N.F.; Robinson, J.E.

    1980-01-01

    An ultrahigh vacuum fracture and transfer device for analysis of irradiated and non-irradiated SS 316 fuel cladding is described. Mechanical property tests used to study the behavior of cladding during reactor transient over-power conditions are reported. The stress vs temperature curves show minimal differences between unirradiated cladding and unfueled cladding. The fueled cladding fails at a lower temperature. All fueled specimens failed in an intergranular mode. (FS)

  9. Microstructural evolution of austenitic stainless steels irradiated to 17 dpa in spectrally tailored experiment of the ORR and HFIR at 400{degrees}C

    SciTech Connect

    Wakai, E.; Hashimoto, N.; Gibson, L.T.

    1997-08-01

    The microstructural evolution of austenitic JPCA aged and solution annealed JPCA, 316R, C, K, and HP steels irradiated at 400{degrees}C in spectrally tailored experiments of the ORR and HFIR has been investigated. The helium generation rates were about 12-16 appm He/dpa on the average up to 17.3 dpa. The number densities and average diameters of dislocation loops in the steels have ranges of 3.3 x 10{sup 21} m{sup -3} and 15.2-26.3 nm, respectively, except for HP steel for which they are 1.1 x 10{sup 23} m{sup -3} and 8.0 nm. Precipitates are formed in all steels except for HP steel, and the number densities and average diameters have ranges of 5.2 x 10{sup 20} - 7.7 x 10{sup 21} m{sup -3} and 3.4- 19.3 nm, respectively. In the 216R, C, and K steels, the precipitates are also formed at grain boundaries, and the mean sizes of these are about 110, 50, and 50 nm, respectively. The number densities of cavities are about 1 x 10{sup 22} m{sup -3} in all the steels. The swelling is low in the steels which form the precipitates.

  10. Synergistic effects on dislocation loops in reduced-activation martensitic steel investigated by single and sequential hydrogen/helium ion irradiation

    NASA Astrophysics Data System (ADS)

    Zhang, Weiping; Luo, Fengfeng; Yu, Yanxia; Zheng, Zhongcheng; Shen, Zhenyu; Guo, Liping; Ren, Yaoyao; Suo, Jinping

    2016-10-01

    Single-beam and sequential-beam irradiations were performed to investigate the H/He synergistic effect on dislocation loops in reduced-activation ferritic/martensitic (RAFM) steels. The irradiations were carried out with 10 keV H+, 18 keV He+ and 160 keV Ar+, alone and in combination at 723 K. He+ single-beam irradiation induced much larger dislocation loops than that induced by both H+ and Ar+ single-beam irradiation. H+ post-irradiation after He+ irradiation further increased the size of dislocation loops, whilst He+ post-irradiation or Ar+ post-irradiation following H+ irradiation only slightly increased the size of dislocation loops. The experiment results indicate that pre-implanted H+ can drastically inhibit the growth while post-implanted H+ can significantly enhance the growth of dislocation loops induced by He+ irradiation. The mechanisms behind the complex synergistic phenomena between H and He and the different roles that H and He played in the growth of dislocation loops are discussed.

  11. Effects of thermal aging and neutron irradiation on the mechanical properties of three-wire stainless steel weld overlay cladding

    SciTech Connect

    Haggag, F.M.; Nanstad, R.K.

    1997-05-01

    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 of aging and 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 {minus}125{degrees}C) but no apparent change in ultimate tensile strength or total elongation. Neutron irradiation reduced the initiation fracture toughness (J{sub Ic}) much more than did thermal aging alone. Irradiation slightly decreased the tearing modulus, but no reduction was caused by thermal aging alone. Other results from tensile, CVN, and fracture toughness specimens showed that the effects of thermal aging at 288 or 343{degrees}C for 20,000 h each were very small and similar to those at 288{degrees}C for 1605 h. The effects of long-term thermal exposure time (50,000 h and greater) at 288{degrees}C will be investigated as the specimens become available in 1996 and beyond.

  12. Effect of neutron irradiation on the properties of the repair welds of the 15Kh2MFA steel

    SciTech Connect

    Morozov, A.M.; Khachaturyants, L.V.

    1986-07-01

    The authors studied the effect of neutron irradiation on the tendency of the metal belonging to the heat affected zone of the weld toward brittle fracture (an increase in the critical temperature of brittleness). For comparison, the authors studied the radiation embrittlement of the original base metal (steel 15Kh2MFA) subjected to the conventional heat treatment of the reactor frames consisting of hardening and high-temperature tempering. Along with these materials, the radiational embrittlement of the base metal in the rehardened condition without tempering was studied. It was concluded that the presence of the regions repaired according to this technology and located in the frame at the level of the reactor core does not pose the problem of decreased resistance to brittle fracture.

  13. Pre-irradiation spatial distribution and stability of boride particles in rapidly solidified boron-doped stainless steels

    SciTech Connect

    Kanani, N.; Arnberg, L.; Harling, O.K.

    1981-01-01

    The time temperature behavior of boride particles has been studied in rapidly solidified ultra low carbon and nitrogen modified 316 stainless steel with 0.088% boron and 0.45% zirconium. The results show that the as-splat-cooled specimens exhibit precipitates at the grain boundaries and triple junctions. For temperatures up to about 750/sup 0/C no significant microstructural changes occur for short heat treatment times. In the temperature range of 750 to 950/sup 0/C, however, particles typically 100 to 500 A in diameter containing Zr and B are formed within the grains. Higher temperatures enhance the formation of such particles and give rise to particle networks. The results show that a fine and uniform distribution of the boride particles almost exclusively within the grains can be achieved if proper annealing conditions are chosen. This type of distribution is an important requirement for the homogeneous production of helium during neutron irradiation in fast reactors.

  14. Corrosion processes of austenitic stainless steels and copper-based materials in gamma-irradiated aqueous environments

    SciTech Connect

    Glass, R.S.

    1985-09-01

    The US Department of Energy is evaluating a site located at Yucca Mountain in Nye County, Nevada, as a potential high-level nuclear waste repository. The rock at the proposed repository horizon (above the water table) is densely welded, devitrified tuff, and the fluid environment in the repository is expected to be primarily air-steam. A more severe environment would be present in the unlikely case of intrusion of vadose groundwater into the repository site. For this repository location, austenitic stainless steels and copper-based materials are under consideration for waste container fabrication. This study focuses on the effects of gamma irradiation on the electrochemical mechanisms of corrosion for the prospective waste container materials. The radiolytic production of such species as hydrogen peroxide and nitric acid are shown to exert an influence on corrosion mechanisms and kinetics.

  15. Extreme embrittlement of austenitic stainless steel irradiated to 75-81 dpa at 335-360{degrees}C

    SciTech Connect

    Porollo, S.I.; Vorobjev, A.N.; Konobeev, Yu.V.

    1997-04-01

    It is generally accepted that void swelling of austenitic steels ceases below some temperature in the range 340-360{degrees}C, and exhibits relatively low swelling rates up to {approximately}400{degrees}C. This perception may not be correct at all irradiation conditions, however, since it was largely developed from data obtained at relatively high displacement rates in fast reactors whose inlet temperatures were in the range 360-370{degrees}C. There is an expectation, however, that the swelling regime can shift to lower temperatures at low displacement rates via the well-known {open_quotes}temperature shift{close_quotes} phenomenon. It is also known that the swelling rates at the lower end of the swelling regime increase continuously at a sluggish rate, never approaching the terminal 1%/dpa level within the duration of previous experiments. This paper presents the results of an experiment conducted in the BN-350 fast reactor in Kazakhstan that involved the irradiation of argon-pressurized thin-walled tubes (0-200 MPa hoop stress) constructed from Fe-16Cr-15Ni-3Mo-Nb stabilized steel in contact with the sodium coolant, which enters the reactor at {approx}270{degrees}C. Tubes in the annealed condition reached 75 dpa at 335{degrees}C, and another set in the 20% cold-worked condition reached 81 dpa at 360{degrees}C. Upon disassembly all tubes, except those in the stress-free condition, were found to have failed in an extremely brittle fashion. The stress-free tubes exhibited diameter changes that imply swelling levels ranging from 9 to 16%. It is expected that stress-enhancement of swelling induced even larger swelling levels in the stressed tubes.

  16. The Effect of Local Heating by Laser Irradiation for Aluminum, Deep Drawing Steel and Copper Sheets in Incremental Sheet Forming

    NASA Astrophysics Data System (ADS)

    Lehtinen, Pekka; Väisänen, Tapio; Salmi, Mika

    Incremental sheet forming is a technique where a metal sheet is formed into a product usually by a CNC-controlled (Computer Numerical Control) round tipped tool. The part is formed as the tool indents into the sheet and follows a contour of the desired product. In single point incremental forming (SPIF) there is no need for tailored tools and dies, since the process requires only a CNC machine, a clamping rig and a simple tool. The effect of applying local heating by laser irradiation from the bottom side of the metal sheet is investigated with a SPIF approach. Using a laser light source for local heating should increase the material ductility and decrease material strength, and thus, increase the formability. The research was performed using 0.50-0.75 mm thick, deep drawing steel, aluminum and copper sheets. The forming was done with a round tipped tool, whose tip diameter was 4 mm. In order to achieve selective heating, a 1 kW fiber laser was attached to a 3-axis stepper motor driven CNC milling machine. The results show that the applied heating increased the maximum achievable wall angle of aluminum and copper products. However, for the steel sheets the local heating reduced the maximum achievable wall angle and increased the surface roughness.

  17. Re-weldability tests of irradiated 316L(N) stainless steel using laser welding technique

    NASA Astrophysics Data System (ADS)

    Yamada, Hirokazu; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Kalinin, George; Kohno, Wataru; Morishima, Yasuo

    2002-12-01

    SS316L(N)-IG is the candidate material for the in-vessel and ex-vessel components of fusion reactors such as ITER (International Thermonuclear Experimental Reactor). This paper describes a study on re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. The laser welding process is used for re-welding of the water cooling branch pipeline repairs. It is clarified that re-welding of SS316L(N)-IG irradiated up to about 0.2 dpa (3.3 appm He) can be carried out without a serious deterioration of tensile properties due to helium accumulation. Therefore, repair of the ITER blanket cooling pipes can be performed by the laser welding process.

  18. Microstructural developments in neutron-irradiated mild steel submerged-arc weld metal

    NASA Astrophysics Data System (ADS)

    Buswell, J. T.; Bischler, P. J. E.; Fenton, S. T.; Ward, A. E.; Phythian, W. J.

    1993-10-01

    The microstructures of Magnox submerged-arc welds have been characterised to investigate the effects of surveillance and accelerated irradiation at temperatures in the range 190-290°C. The radiation hardening and embrittlement is influenced by the precipitation of Cu from solid solution. Mn has been found in the Cu-rich precipitates, together with an indication of P. The precipitates have structure coherent with the ferrite matrix and maintain a constant mean diameter during extended irradiation. Evidence has been obtained indicating that dislocation loops contribute to a matrix damage component in these welds.

  19. Comparison of swelling and cavity microstructural development for type 316 stainless steel irradiated in EBR-II and HFIR

    SciTech Connect

    Maziasz, P.J.

    1983-01-01

    Comparison of swelling and cavity microstructures for one heat of 20% cold-worked (CW) type 316 stainless steel (316) irradiated at 500 to 650/sup 0/C in EBR-II (up to 75 dpa) and HFIR (up to 61 dpa) suggests that void growth and swelling are suppressed by the higher helium generation found in HFIR. Instead of voids, many small bubbles develop in the CW 316 in HFIR and resist conversion to voids. However, similar comparison of solution-annealed (SA) 316 irradiated in EBR-II and HFIR at 500 to 550/sup 0/C leads to an opposite conclusion; void swelling is enhanced by helium in HFIR. Many more bubbles nucleate in SA 316 at low fluence in HFIR compared to EBR-II, but bimodel distributions and rapid coarsening eventually lead to high swelling due to high concentrations of matrix ands precipitate-associated voids in HFIR. A key to the swelling resistance of the CW 316 in HFIR appears to be the development of a sufficiently cavity-dominated sink system in the early stages of evolution.

  20. The effects of silicon and titanium on void swelling and phase transformations in neutron irradiated 12Cr-15Ni steels

    NASA Astrophysics Data System (ADS)

    Boothby, R. M.; Williams, T. M.

    1988-05-01

    12Cr-15Ni-0.25Ti steels with Si additions of 0.5, 0.9 and 1.4 wt% have been irradiated to a maximum dose of 47 dpa at temperatures ranging from 399 to 649°C. Detailed microstructural examinations of void swelling, precipitation behaviour and austenite instability have been made. Assessments of swelling and matrix phase transformations have also been made using density and induced magnetization measurements respectively. Austenite instability was increased by Si additions; the transformation product was usually ferrite although some martensite was also observed, and compositional fluctuations in untransformed austenite were detected. Precipitation, particularly of G phase, became more extensive and swelling in solution-treated alloys was reduced at higher Si contents. Enhanced growth of voids attached to G phase precipitates was observed. Cold-working decreased both swelling and ferrite formation. A fine dispersion of TiC was effective in suppressing swelling at high irradiation temperature as long as the precipitates remained stable. The stability of TiC was increased by cold-working but reduced by Si additions.

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

    SciTech Connect

    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 protective 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.

  2. Precipitate behavior in self-ion irradiated stainless steels at high doses

    NASA Astrophysics Data System (ADS)

    Jiao, Z.; Was, G. S.

    2014-06-01

    To study radiation-induced precipitation at high doses, solution annealed 304L SS and cold worked 316 SS were irradiated to 46 and 260 dpa at 380 °C using 5 MeV Fe++ and the radiation-induced precipitates were examined using atom probe tomography. Ni/Si-rich clusters were observed in all examined conditions. G-phase precipitates were observed in 316 SS at 46 dpa but only appeared in 304L SS at 260 dpa. Using the neutron irradiation to 46 dpa at 320 °C as a reference, the temperature shift for cold worked 316 SS appeared to be smaller than that of solution annealed 304L SS, probably due to the high density of dislocations, which served as defect sinks and mitigated the effect of high dose rate.

  3. A Hierarchical Upscaling Method for Predicting Strength of Materials under Thermal, Radiation and Mechanical loading - Irradiation Strengthening Mechanisms in Stainless Steels

    SciTech Connect

    Li, Dongsheng; Zbib, Hussein M.; Garmestani, Hamid; Sun, Xin; Khaleel, Mohammad A.

    2011-07-01

    Stainless steels based on Fe-Cr-Ni alloys are the most popular structural materials used in reactors. High energy particle irradiation of in this kind of polycrystalline structural materials usually produces irradiation hardening and embrittlement. The development of predictive capability for the influence of irradiation on mechanical behavior is very important in materials design for next-generation reactors. Irradiation hardening is related to structural information crossing different length scale, such as composition, dislocation, crystal orientation distribution and so on. To predict the effective hardening, the influence factors along different length scales should be considered. A multiscale approach was implemented in this work to predict irradiation hardening of iron based structural materials. Three length scales are involved in this multiscale model: nanometer, micrometer and millimeter. In the microscale, molecular dynamics (MD) was utilized to predict on the edge dislocation mobility in body centered cubic (bcc) Fe and its Ni and Cr alloys. On the mesoscale, dislocation dynamics (DD) models were used to predict the critical resolved shear stress from the evolution of local dislocation and defects. In the macroscale, a viscoplastic self-consistent (VPSC) model was applied to predict the irradiation hardening in samples with changes in texture. The effects of defect density and texture were investigated. Simulated evolution of yield strength with irradiation agrees well with the experimental data of irradiation strengthening of stainless steel 304L, 316L and T91. This multiscale model we developed in this project can provide a guidance tool in performance evaluation of structural materials for next-generation nuclear reactors. Combining with other tools developed in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the models developed will have more impact in improving the reliability of current reactors and affordability of new

  4. Microstructure of Au-ion irradiated 316L and FeNiCr austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Jublot-Leclerc, S.; Li, X.; Legras, L.; Lescoat, M.-L.; Fortuna, F.; Gentils, A.

    2016-11-01

    Thin foils of 316L were irradiated in situ in a Transmission Electron Microscope with 4 MeV Au ions at 450 °C and 550 °C. Similar irradiations were performed at 450 °C in FeNiCr. The void and dislocation microstructure of 316L is found to depend strongly on temperature. At 450 °C, a dense network of dislocation lines is observed in situ to grow from black dot defects by absorption of other black dots and interstitial clusters whilst no Frank loops are detected. At 550 °C, no such network is observed but large Frank loops and perfect loops whose sudden appearance is concomitant with a strong increase in void density as a result of a strong coupling between voids and dislocations. Moreover, differences in both alloys microstructure show the major role played by the minor constituents of 316L, increasing the stacking fault formation energy, and possibly leading to significant differences in swelling behaviour.

  5. Experimental study on double-pulse laser ablation of steel upon multiple parallel-polarized ultrashort-pulse irradiations

    NASA Astrophysics Data System (ADS)

    Schille, Joerg; Schneider, Lutz; Kraft, Sebastian; Hartwig, Lars; Loeschner, Udo

    2016-07-01

    In this paper, double-pulse laser processing is experimentally studied with the aim to explore the influence of ultrashort pulses with very short time intervals on ablation efficiency and quality. For this, sequences of 50 double pulses of varied energy and inter-pulse delay, as adjusted between 400 fs and 18 ns by splitting the laser beam into two optical paths of different length, were irradiated to technical-grade stainless steel. The depth and the volume of the craters produced were measured in order to evaluate the efficiency of the ablation process; the crater quality was analyzed by SEM micrographs. The results obtained were compared with craters produced with sequences of 50 single pulses and energies equal to the double pulse. It is demonstrated that double-pulse processing cannot exceed the ablation efficiency of single pulses of optimal fluence, but the ablation crater surface formed smoother if inter-pulse delay was in the range between 10 ns and 18 ns. In addition, the influence of pulse duration and energy distribution between the individual pulses of the double pulse on ablation was studied. For very short inter-pulse delay, no significant effect of energy variation within the double pulse on removal rate was found, indicating that the double pulse acts as a big single pulse of equal energy. Further, the higher removal efficiency was achieved when double-pulse processing using femtosecond pulses instead of picosecond pulses.

  6. Comparative irradiation study of reactor pressure-vessel steel weld metals

    SciTech Connect

    Leitz, C.; Gerscha, A.; Hofmann, G.; Strobel, H.J.

    1982-01-01

    Charpy-V specimens of several test welds made of MnMo, MnCrMo, MnNiMo and MnNiCrMo types were irradiated simultaneously in two series at 315/sup 0/C and 285/sup 0/C, respectively, with neutron fluences between 8 x 10/sup 18/ and 5 x 10/sup 19/ n/cm/sup 2/ (E > 1 MeV). The known influence of copper on the 41 J temperature shift (..delta..T/sub 41/) was clearly detected for variations from 0.08 to about 0.3%. A considerable variation in embrittlement of low copper (below 0.08%) welds was not traceable to copper or phosphorus. Obviously it is related to variations of the nickel content (up to 1.7%). The effects of copper and nickel seem to be independent on each other. Both, however, have an influence on the slope of the ..delta..T/sub 41/ against fluence curves. The effects of other elements as silicon, phosphorus, arsenic, manganese, chromium, molybdenum and of welding procedure and welding flux are discussed in the paper. Due to overlapping, these effects could not be clearly defined. The Charpy-V upper shelf energy drop shows an effect equivalent to overaging. The shelf drop recovers by further irradiation in the region of 1 to 4 x 10/sup 19/ n/cm/sup 2/ for all low copper weld metals except that with 1.7% nickel. The shelf drop sequence of the materials tested is in general similar to ..delta..T/sub 41/ sequence with some differences in detail.

  7. Microstructural Characterization of Deformation Localization at Small Strains in a Neutron Irradiated 304 Stainless Steel

    SciTech Connect

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

    2014-01-01

    Deformation localization and structure evolution were investigated in an AISI 304 austenitic stainless steel deformed to 0.8% strain. Using SEM-EBSD, it was shown local plastic deformation may reach significant levels even when the bulk averaged strain level remains below 1%. Local misorientation values up to 24 were observed in these regions of high local plastic deformation. EBSD analysis of FIB lift-out specimens demonstrated that local misorientation level was highest near the free surface and diminished with increasing depth. (S)TEM observations on the same specimen indicated the local density of dislocation channels may vary up to an order of magnitude depending on local grain configuration, distance to the surface and/or local grain boundary structure. It was found that in the case of RT deformation, dislocation defect-free channels may contain twin or may be twin-free with twinning occurring inside channels. Formation of BCC-phase colonies (martensite) was observed in near-surface layer whereas no transformation in the volume of the specimen was detected at this strain level. Martensite formation was associated with channel-grain boundary intersection points where high local misorientation was observed using EBSD.

  8. A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels

    SciTech Connect

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

    2007-11-01

    The reactor pressure vessels (RPVs) of commercial nuclear power plants are subject to embrittlement due to exposure to high-energy neutrons from the core, which causes changes in material toughness properties that increase with radiation exposure and are affected by many variables. Irradiation embrittlement of RPV beltline materials is currently evaluated using Regulatory Guide 1.99 Revision 2 (RG1.99/2), which presents methods for estimating the shift in Charpy transition temperature at 30 ft-lb (TTS) and the drop in Charpy upper shelf energy (ΔUSE). The purpose of the work reported here is to improve on the TTS correlation model in RG1.99/2 using the broader database now available and current understanding of embrittlement mechanisms. The USE database and models have not been updated since the publication of NUREG/CR-6551 and, therefore, are not discussed in this report. The revised embrittlement shift model is calibrated and validated on a substantially larger, better-balanced database compared to prior models, including over five times the amount of data used to develop RG1.99/2. It also contains about 27% more data than the most recent update to the surveillance shift database, in 2000. The key areas expanded in the current database relative to the database available in 2000 are low-flux, low-copper, and long-time, high-fluence exposures, all areas that were previously relatively sparse. All old and new surveillance data were reviewed for completeness, duplicates, and discrepancies in cooperation with the American Society for Testing and Materials (ASTM) Subcommittee E10.02 on Radiation Effects in Structural Materials. In the present modeling effort, a 10% random sample of data was reserved from the fitting process, and most aspects of the model were validated with that sample as well as other data not used in calibration. The model is a hybrid, incorporating both physically motivated features and empirical calibration to the U.S. power reactor surveillance

  9. Determination of creep compliance and creep-swelling coupling coefficients for neutron-irradiated titanium-modified stainless steel at @400 degree C

    SciTech Connect

    Toloczko, M.B. ); Garner, F.A. ); Eiholzer, C.R. )

    1991-11-01

    Irradiation creep data from FFTF-MOTA at {approximately}400{degrees}C were analyzed for nine 20% cold-worked titanium-modified type 316 stainless steels, each of which exhibits a different duration for the transient regime of swelling. One of these steels was the fusion prime candidate alloy designated PCA. The others were various developmental breeder reactor heats. The analysis was based on the assumption that the B{sub 0} + DS creep model applies to these steels at this temperature. This assumption was found to be valid. A creep-swelling coupling coefficient of D {approx} 0.6 {times} 10{sup {minus}2} MPa{sup {minus}1} was found for all steels that had developed a significant level of swelling. This result is in excellent agreement with the results of earlier studies conducted in EBR-II using annealed AISI 304L and also 10% and 20% cold-worked AISI 316 stainless steels. There appears to be some enhancement of swelling by stress, contradicting an important assumption in the analysis and leading to an apparent but misleading nonlinearity of creep with respect to stress.

  10. Determination of creep compliance and creep-swelling coupling coefficients for neutron-irradiated titanium-modified stainless steel at {approximately}400{degree}C

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

    1991-11-01

    Irradiation creep data from FFTF-MOTA at {approximately}400{degrees}C were analyzed for nine 20% cold-worked titanium-modified type 316 stainless steels, each of which exhibits a different duration for the transient regime of swelling. One of these steels was the fusion prime candidate alloy designated PCA. The others were various developmental breeder reactor heats. The analysis was based on the assumption that the B{sub 0} + DS creep model applies to these steels at this temperature. This assumption was found to be valid. A creep-swelling coupling coefficient of D {approx} 0.6 {times} 10{sup {minus}2} MPa{sup {minus}1} was found for all steels that had developed a significant level of swelling. This result is in excellent agreement with the results of earlier studies conducted in EBR-II using annealed AISI 304L and also 10% and 20% cold-worked AISI 316 stainless steels. There appears to be some enhancement of swelling by stress, contradicting an important assumption in the analysis and leading to an apparent but misleading nonlinearity of creep with respect to stress.

  11. Correlation of microstructure with hardness and wear resistance in (TiC, SiC)/stainless steel surface composites fabricated by high-energy electron-beam irradiation

    NASA Astrophysics Data System (ADS)

    Yun, Eunsub; Kim, Yong Chan; Lee, Sunghak; Kim, Nack J.

    2004-03-01

    Stainless-steel-based surface composites reinforced with TiC and SiC carbides were fabricated by high-energy electron beam irradiation. Four types of powder/flux mixtures, i.e., TiC, (Ti + C), SiC, and (Ti + SiC) powders with 40 wt. pct of CaF2 flux, were deposited evenly on an AISI 304 stainless steel substrate, which was then irradiated with an electron beam. TiC agglomerates and pores were found in the surface composite layer fabricated with TiC powders because of insufficient melting of TiC powders. In the composite layer fabricated with Ti and C powders having lower melting points than TiC powders, a number of primary TiC carbides were precipitated while very few TiC agglomerates or pores were formed. This indicated that more effective TiC precipitation was obtained from the melting of Ti and C powders than of TiC powders. A large amount of precipitates such as TiC and Cr7C3 improved the hardness, high-temperature hardness, and wear resistance of the surface composite layer two to three times greater than that of the stainless steel substrate. In particular, the surface composite fabricated with SiC powders had the highest volume fraction of Cr7C3 distributed along solidification cell boundaries, and thus showed the best hardness, high-temperature hardness, and wear resistance.

  12. Neutron and X-ray diffraction analysis of the effect of irradiation dose and temperature on microstructure of irradiated HT-9 steel

    NASA Astrophysics Data System (ADS)

    Mosbrucker, P. L.; Brown, D. W.; Anderoglu, O.; Balogh, L.; Maloy, S. A.; Sisneros, T. A.; Almer, J.; Tulk, E. F.; Morgenroth, W.; Dippel, A. C.

    2013-11-01

    Material harvested from several positions within a nuclear fuel duct (the ACO-3 duct) used in a 6-year irradiation of a fuel assembly in the Fast Flux Test Reactor Facility (FFTF) was examined using neutron and high-energy X-ray diffraction. Samples with a wide range of irradiation dose and irradiation temperature history, reaching doses of up to 147 dpa and temperatures of up to 777 K, were examined. The response of various microstructural characteristics such as the weight fraction of M23C6 carbides, the dislocation density and character, and the crystallographic texture were determined using whole profile analysis of the diffraction data and related to the macroscopic mechanical behavior. For instance, the dislocation density was observed to be intimately linked with observed flow strength of the irradiated materials, following the Taylor law. In general, at the high doses studied in this work, the irradiation temperature is the predominant controlling factor of the dislocation density and, thus, the flow strength of the irradiated material. The results, representing some of the first diffraction work done on samples exposed to such a high received dose, demonstrate how non-destructive and stand-off diffraction techniques can be used to characterize irradiation induced microstructure and at least estimate mechanical properties in irradiated materials without exposing workers to radiation hazards.

  13. Dependence of the rate of steady-state swelling of fuel-element claddings made of ChS68 steel on the characteristics of neutron irradiation

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

    Rate of steady-state swelling of fuel-element sheaths made of the 06Kh16N15M2G2TFR steel in the course of their operation in a BN-600 reactor has been calculated. In the calculations, the diffusion characteristics of point defects and the results of the determination of the characteristics of the irradiation-induced porosity have been used. The dependence of the dose rate of steady-state swelling on neutron-irradiation characteristics has been analyzed. It has been established that the dose rate of swelling at the steady-state stage is independent of the energy of migration of vacancies and the rate of generation of atomic displacements.

  14. Elastic modules and thermal conductivity of neutron irradiated type 13Cr2MoNbVB ferritic—martensitic steel

    NASA Astrophysics Data System (ADS)

    Zakharova, M. I.; Artemov, N. A.; Petrov, D. V.

    1996-10-01

    The temperature dependencies of elastic modulus, internal friction, Poisson ratio in the range from 20 to 600°C as well as of thermal conductivity and electrical resistance at temperatures from 20 to 1000°C have been determined for type 13Cr2MoNbVB ferritic—martensitic steel irradiated at 280°C in the BN-350 fast reactor to neutron fluence of 4.03 × 10 26 n/m 2 ( E > 0.1 MeV). During isochronal annealings at temperatures up to 0.65 Tm the recovery of the properties for irradiated steel has been investigated in the range of homologous temperatures from 0.28 Tm to 0.53 Tm. A pronounce recovery was observed at four substages in the temperature interval from 380 to 620°C. The activation energies determined for all recovery substages varied from 1.11 eV to 4.09 eV.

  15. Sub-micron magnetic patterns and local variations of adhesion force induced in non-ferromagnetic amorphous steel by femtosecond pulsed laser irradiation

    NASA Astrophysics Data System (ADS)

    Zhang, Huiyan; Feng, Yuping; Nieto, Daniel; García-Lecina, Eva; Mcdaniel, Clare; Díaz-Marcos, Jordi; Flores-Arias, María Teresa; Gerard M., O.'connor; Baró, Maria Dolors; Pellicer, Eva; Sort, Jordi

    2016-05-01

    Periodic ripple and nanoripple patterns are formed at the surface of amorphous steel after femtosecond pulsed laser irradiation (FSPLI). Formation of such ripples is accompanied with the emergence of a surface ferromagnetic behavior which is not initially present in the non-irradiated amorphous steel. The occurrence of ferromagnetic properties is associated with the laser-induced devitrification of the glassy structure to form ferromagnetic (α-Fe and Fe3C) and ferrimagnetic [(Fe,Mn)3O4 and Fe2CrO4] phases located in the ripples. The generation of magnetic structures by FSPLI turns out to be one of the fastest ways to induce magnetic patterning without the need of any shadow mask. Furthermore, local variations of the adhesion force, wettability and nanomechanical properties are also observed and compared to those of the as-cast amorphous alloy. These effects are of interest for applications (e.g., biological, magnetic recording, etc.) where both ferromagnetism and tribological/adhesion properties act synergistically to optimize material performance.

  16. Heat treatment effects on impact toughness of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated to 100 dpa

    SciTech Connect

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

    1997-08-01

    Plates of 9Cr-1MoVNb and 12Cr-1MoVW steels were given four different heat treatments: two normalizing treatments were used and for each normalizing treatment two tempers were used. Miniature Charpy specimens from each heat treatment were irradiated to {approx}19.5 dpa at 365{degrees}C and to {approx}100 dpa at 420{degrees}C in the Fast Flux Test Facility (FFTF). In previous work, the same materials were irradiated to 4-5 dpa at 365{degrees}C and 35-36 dpa at 420{degrees}C in FFTF. The tests indicated that prior austenite grain size, which was varied by the different normalizing treatments, had a significant effect on impact behavior of the 9Cr-1MoVNb but not on the 12Cr-1MoVW. Tempering treatment had relatively little effect on the shift in DBTT for both steels. Conclusions are presented on how heat treatment can be used to optimize impact properties.

  17. Correlation of radiation-induced changes in microstructure/microchemistry, density and thermo-electric power of type 304L and 316 stainless steels irradiated in the Phénix reactor

    NASA Astrophysics Data System (ADS)

    Renault Laborne, Alexandra; Gavoille, Pierre; Malaplate, Joël; Pokor, Cédric; Tanguy, Benoît

    2015-05-01

    Annealed specimens of type 304L and 316 stainless steel and cold-worked 316 specimens were irradiated in the Phénix reactor in the temperature range 381-394 °C and to different damage doses up to 39 dpa. The microstructure and microchemistry of both 304L and 316 have been examined using the combination of the different techniques of TEM to establish the void swelling and precipitation behavior under neutron irradiation. TEM observations are compared with results of measurements of immersion density and thermo-electric power obtained on the same irradiated stainless steels. The similarities and differences in their behavior on different scales are used to understand the factors in terms of the chemical composition and metallurgical state of steels, affecting the precipitation under irradiation and the swelling behavior. Irradiation induces the formation of some precipitate phases (e.g., M6C and M23C6-type carbides, and γ'- and G-phases), Frank loops and cavities. According to the metallurgical state and chemical composition of the steel, the amount of each type of radiation-induced defects is not the same, affecting their density and thermo-electric power.

  18. Radiation-induced instability of MnS precipitates and its possible consequences on irradiation-induced stress corrosion cracking of austenitic stainless steels

    SciTech Connect

    Chung, H.M.; Sanecki, J.E.; Garner, F.A.

    1996-12-01

    Irradiation-assisted stress corrosion cracking (IASCC) is a significant materials issue for the light water reactor (LWR) industry and may also pose a problem for fusion power reactors that will use water as coolant. A new metallurgical process is proposed that involves the radiation-induced release into solution of minor impurity elements not usually thought to participate in IASCC. MnS-type precipitates, which contain most of the sulfur in stainless steels, are thought to be unstable under irradiation. First, Mn transmutes strongly to Fe in thermalized neutron spectra. Second, cascade-induced disordering and the inverse Kirkendall effect operating at the incoherent interfaces of MnS precipitates are thought to act as a pump to export Mn from the precipitate into the alloy matrix. Both of these processes will most likely allow sulfur, which is known to exert a deleterious influence on intergranular cracking, to re-enter the matrix. To test this hypothesis, compositions of MnS-type precipitates contained in several unirradiated and irradiated heats of Type 304, 316, and 348 stainless steels (SSs) were analyzed by Auger electron spectroscopy. Evidence is presented that shows a progressive compositional modification of MnS precipitates as exposure to neutrons increases in boiling water reactors. As the fluence increases, the Mn level in MnS decreases, whereas the Fe level increases. The S level also decreases relative to the combined level of Mn and Fe. MnS precipitates were also found to be a reservoir of other deleterious impurities such as F and O which could be also released due to radiation-induced instability of the precipitates.

  19. Irradiation response in weldment and HIP joint of reduced activation ferritic/martensitic steel, F82H

    SciTech Connect

    Hirose, Takanori; Sokolov, Mikhail A; Ando, M.; Tanigawa, H.; Shiba, K.; Stoller, Roger E; Odette, G.R.

    2013-11-01

    This work investigates irradiation response in the joints of F82H employed for a fusion breeding blanket. The joints, which were prepared using welding and diffusion welding, were irradiated up to 6 dpa in the High Flux Isotope Reactor at the Oak Ridge National Laboratory. Post-irradiation tests revealed hardening in weldment (WM) and base metal (BM) greater than 300 MPa. However, the heat affected zones (HAZ) exhibit about half that of WM and BM. Therefore, neutron irradiation decreased the strength of the HAZ, leaving it in danger of local deformation in this region. Further the hardening in WM made with an electron beam was larger than that in WM made with tungsten inert gas welding. However the mechanical properties of the diffusion-welded joint were very similar to those of BM even after the irradiation.

  20. Evaluation of critical resolved shear strength and deformation mode in proton-irradiated austenitic stainless steel using micro-compression tests

    NASA Astrophysics Data System (ADS)

    Jin, Hyung-Ha; Ko, Eunsol; Kwon, Junhyun; Hwang, Seong Sik; Shin, Chansun

    2016-03-01

    Micro-compression tests were applied to evaluate the changes in the strength and deformation mode of proton-irradiated commercial austenitic stainless steel. Proton irradiation generated small dots at low dose levels and Frank loops at high dose levels. The increase in critical resolved shear stresses (CRSS) was measured from micro-compression of pillars and the Schmid factor calculated from the measured loading direction. The magnitudes of the CRSS increase were in good agreement with the values calculated from the barrier hardening model using the measured size and density of radiation defects. The deformation mode changed upon increasing the irradiation dose level. At a low radiation dose level, work hardening and smooth flow behavior were observed. Increasing the dose level resulted in the flow behavior changing to a distinct heterogeneous flow, yielding a few large strain bursts in the stress-strain curves. The change in the deformation mode was related to the formation and propagation of defect-free slip bands. The effect of the orientation of the pillar or loading direction on the strengths is discussed.

  1. Relationship between swelling and irradiation creep in cold-worked PCA stainless steel irradiated to {approximately}178 dpa at {approximately}400{degrees}C

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.

    1993-09-01

    The eighth and final irradiation segment for pressurized tubes constructed from the fusion Prime Candidate Alloy (PCA) has been completed in FFTF. At 178 dpa and {approximately}400{degrees}C, the irradiation creep of 20% cold-worked PCA has become dominated by the {open_quotes}creep disappearance{close_quotes} phenomenon. The total diametral deformation rate has reached the limiting value of 0.33%/dpa at the three highest stress levels employed in this test. The stress-enhancement of swelling tends to camouflage the onset of creep disappearance, however, requiring the use of several non-traditional techniques to extract the creep coefficients. No failures occurred in these tubes, even though the swelling ranged from {approximately}20 to {approximately}40%.

  2. Heavy-Section Steel Irradiation Program. Volume 2, No. 2: Semiannual progress report, April--September 1991

    SciTech Connect

    Corwin, W.R.

    1994-10-01

    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 stools 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 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-sheer 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 April to September 1991.

  3. Tensile and fatigue data for irradiated and unirradiated AISI 310 stainless steel and titanium - 5 percent aluminum - 2.5 percent tin: Application of the method of universal slopes

    NASA Technical Reports Server (NTRS)

    Debogdan, C. E.

    1973-01-01

    Irradiated and unirradiated tensile and fatigue specimens of AISI 310 stainless steel and Ti-5Al-2.5Sn were tested in the range of 100 to 10,000 cycles to failure to determine the applicability of the method of universal slopes to irradiated materials. Tensile data for both materials showed a decrease in ductility and increase in ultimate tensile strength due to irradiation. Irradiation caused a maximum change in fatigue life of only 15 to 20 percent for both materials. The method of universal slopes predicted all the fatigue data for the 310 SS (irradiated as well as unirradiated) within a life factor of 2. For the titanium alloy, 95 percent of the data was predicted within a life factor of 3.

  4. A facile preparation route for netlike microstructures on a stainless steel using an ethanol-mediated femtosecond laser irradiation.

    PubMed

    Bian, Hao; Yang, Qing; Liu, Hewei; Chen, Feng; Du, Guangqing; Si, Jinhai; Hou, Xun

    2013-03-01

    Netlike or porous microstructures are highly desirable in metal implants and biomedical monitoring applications. However, realization of such microstructures remains technically challenging. Here, we report a facile and environmentally friendly method to prepare netlike microstructures on a stainless steel by taking the full advantage of the liquid-mediated femtosecond laser ablation. An unordered netlike structure and a quasi-ordered array of holes can be fabricated on the surface of stainless steel via an ethanol-mediated femtosecond laser line-scan method. SEM analysis of the surface morphology indicates that the porous netlike structure is in the micrometer scale and the diameter of the quasi-ordered holes ranges from 280 nm to 320 nm. Besides, we find that the obtained structures are tunable by altering the laser processing parameters especially scanning speed.

  5. Double Sided Irradiation for Laser-assisted Shearing of Ultra High Strength Steels with Process Integrated Hardening

    NASA Astrophysics Data System (ADS)

    Brecher, Christian; Emonts, Michael; Eckert, Markus; Weinbach, Matthias

    Most small or medium sized parts produced in mass production are made by shearing and forming of sheet metal. This technology is cost effective, but the achievable quality and geometrical complexity are limited when working high and highest strength steel. Based on the requirements for widening the process limits of conventional sheet metal working the Fraunhofer IPT has developed the laser-assisted sheet metal working technology. With this enhancement it is possible to produce parts made of high and highest strength steel with outstanding quality, high complexity and low tool wear. Additionally laser hardening has been implemented to adjust the mechanical properties of metal parts within the process. Currently the process is limited to lower sheet thicknesses (<2 mm) to maintain short cycle times. To enable this process for larger geometries and higher sheet thicknesses the Fraunhofer IPT developed a system for double sided laser-assisted sheet metal working within progressive dies.

  6. Molecular desorption of stainless steel vacuum chambers irradiated with 4.2 MeV/u lead ions

    NASA Astrophysics Data System (ADS)

    Mahner, E.; Hansen, J.; Laurent, J.-M.; Madsen, N.

    2003-01-01

    In preparation for the heavy ion program of the Large Hadron Collider at CERN, accumulation and cooling tests with lead ion beams have been performed in the Low Energy Antiproton Ring. These tests have revealed that due to the unexpected large outgassing of the vacuum system, the dynamic pressure of the ring could not be maintained low enough to reach the required beam intensities. To determine the actions necessary to lower the dynamic pressure rise, an experimental program has been initiated for measuring the molecular desorption yields of stainless steel vacuum chambers by the impact of 4.2 MeV/u lead ions with the charge states +27 and +53. The test chambers were exposed either at grazing or at perpendicular incidence. Different surface treatments (glow discharges, nonevaporable getter coating) are reported in terms of the molecular desorption yields for H2, CH4, CO, Ar, and CO2. Unexpected large values of molecular yields per incident ion up to 2×104 molecules/ion have been observed. The reduction of the ion-induced desorption yield due to continuous bombardment with lead ions (beam cleaning) has been investigated for five different stainless steel vacuum chambers. The implications of these results for the vacuum system of the future Low Energy Ion Ring and possible remedies to reduce the vacuum degradation are discussed.

  7. Highly antibacterial activity of N-doped TiO2 thin films coated on stainless steel brackets under visible light irradiation

    NASA Astrophysics Data System (ADS)

    Cao, Shuai; Liu, Bo; Fan, Lingying; Yue, Ziqi; Liu, Bin; Cao, Baocheng

    2014-08-01

    In this study, the radio frequency (RF) magnetron sputtering method was used to prepare a TiO2 thin film on the surface of stainless steel brackets. Eighteen groups of samples were made according to the experimental parameters. The crystal structure and surface morphology were characterized by X-ray diffraction, and scanning electron microscopy, respectively. The photocatalytic properties under visible light irradiation were evaluated by measuring the degradation ratio of methylene blue. The sputtering temperature was set at 300 °C, and the time was set as 180 min, the ratio of Ar to N was 30:1, and annealing temperature was set at 450 °C. The thin films made under these parameters had the highest visible light photocatalytic activity of all the combinations of parameters tested. Antibacterial activities of the selected thin films were also tested against Lactobacillus acidophilus and Candida albicans. The results demonstrated the thin film prepared under the parameters above showed the highest antibacterial activity.

  8. Void swelling and microstructure evolution at very high damage level in self-ion irradiated ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Getto, E.; Sun, K.; Monterrosa, A. M.; Jiao, Z.; Hackett, M. J.; Was, G. S.

    2016-11-01

    The void swelling and microstructure evolution of ferritic-martensitic alloys HT9, T91 and T92 were characterized following irradiation with Fe++ ions at 460 °C to damage levels of 75-650 displacements per atom with 10 atom parts per million pre-implanted helium. Steady state swelling rate of 0.033%/ dpa was determined for HT9, the least swelling resistant alloy, and 0.007%/ dpa in T91. In T91, resistance was due to suppression of void nucleation. Swelling resistance was greatest in T92, with a low density (∼1 × 1020 m-3) of small voids that had not grown appreciably, indicating suppression of nucleation and growth. Additional heats of T91 indicated that alloy composition was not the determining factor of swelling resistance. Carbon and chromium-rich M2X precipitates formed at 250 dpa and were correlated with decreased nucleation in T91 and T92, but did not affect void growth in HT9. Dislocation and G-phase microstructure evolution was analyzed up to 650 dpa in HT9.

  9. Results of charpy V-notch impact testing of structural steel specimens irradiated at {approximately}30{degrees}C to 1 x 10{sup 16} neutrons/cm{sup 2} in a commercial reactor cavity

    SciTech Connect

    Iskander, S.K.; Stoller, R.E.

    1997-04-01

    A capsule containing Charpy V-notch (CVN) and mini-tensile specimens was irradiated at {approximately} 30{degrees}C ({approximately} 85{degrees}F) in the cavity of a commercial nuclear power plant to a fluence of 1 x 10{sup 16} neutrons/cm{sup 2} (> 1MeV). The capsule included six CVN impact specimens of archival High Flux Isotope Reactor A212 grade B ferritic steel and five CVN impact specimens of a well-studied A36 structural steel. This irradiation was part of the ongoing study of neutron-induced damage effects at the low temperature and flux experienced by reactor supports. The plant operators shut down the plant before the planned exposure was reached. The exposure of these specimens produced no significant irradiation-induced embrittlement. Of interest were the data on unirradiated specimens in the L-T orientation machined from a single plate of A36 structural steel, which is the same specification for the structural steel used in some reactor supports. The average CVN energy of five unirradiated specimens obtained from one region of the plate and tested at room temperature was {approximately} 99 J, while the energy of 11 unirradiated specimens from other locations of the same plate was 45 J, a difference of {approximately} 220%. The CVN impact energies for all 18 specimens ranged from a low of 32 J to a high of 111 J. Moreover, it appears that the University of Kansas CVN impact energy data of the unirradiated specimens at the 100-J level are shifted toward higher temperatures by about 20 K. The results were an example of the extent of scatter possible in CVN impact testing. Generic values for the CVN impact energy of A36 should be used with caution in critical applications.

  10. Heavy-Section Steel Irradiation (HSSI) Program (W6953) Monthly Letter Status Report - January 2001 - ORNL/HSSI (6953) MLSR-2001/4

    SciTech Connect

    Rosseel, T.M.

    2001-02-20

    This report is issued monthly by the staff of the Heavy-Section Steel Irradiation (HSSI) Program (JCN:W6953) to provide the Nuclear Regulatory Commission (NRC) staff with summaries of technical highlights, important issues, and financial and milestone status within the program. This report gives information on several topics corresponding to events during the reporting month: (1) overall project objective, (2) technical activities, (3) meetings and trips, (4) publications and presentations, (5) property acquired, (6) problem areas, and (7) plans for the next reporting period. Next the report gives a breakdown of overall program costs as well as cost summaries and earned-value-based estimates for performance for the total program and for each of the eight program tasks. The seven tasks correspond to the 189, dated March 23, 1998, and modified by the inclusion of the former ''Embrittlement Data Base and Dosimetry Evaluation'' Program, JCN 6164 in March, 1999. The final part of the report provides financial status for all tasks and status reports for selected milestones within each task. The task milestones address the period from October 2000 to March 2003, while the individual task budgets address the period from October 2000 to February 2001. Beginning in October, 1992, the monthly business calendar of the Oak Ridge National Laboratory was changed and no longer coincides with the Gregorian/Julian calendar. The business month now ends earlier than the last day of the calendar month to allow adequate time for processing required financial reports to the Department of Energy. The precise reporting period for each month is indicated on the financial and milestone charts by including the exact start and finish dates for the current business month.

  11. Nuclear transmutation in steels

    NASA Astrophysics Data System (ADS)

    Belozerova, A. R.; Shimanskii, G. A.; Belozerov, S. V.

    2009-05-01

    The investigations of the effects of nuclear transmutation in steels that are widely used in nuclear power and research reactors and in steels that are planned for the application in thermonuclear fusion plants, which are employed under the conditions of a prolonged action of neutron irradiation with different spectra, made it possible to study the effects of changes in the isotopic and chemical composition on the tendency of changes in the structural stability of these steels. For the computations of nuclear transmutation in steels, we used a program complex we have previously developed on the basis of algorithms for constructing branched block-type diagrams of nuclide transformations and for locally and globally optimizing these diagrams with the purpose of minimizing systematic errors in the calculation of nuclear transmutation. The dependences obtained were applied onto a Schaeffler diagram for steels used for structural elements of reactors. For the irradiation in fission reactors, we observed only a weak influence of the effects of nuclear transmutation in steels on their structural stability. On the contrary, in the case of irradiation with fusion neutrons, a strong influence of the effects of nuclear transmutation in steels on their structural stability has been noted.

  12. Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ≈295 °C to ≈6.5 dpa

    DOE PAGES

    Maloy, Stuart A.; Saleh, Tarik A.; Anderoglu, Osman; Romero, Tobias J.; Odette, G. Robert; Yamamoto, Takuya; Li, S.; Cole, James I.; Fielding, Randall

    2015-08-06

    Tensile test results at 25 and 300 °C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress–strain curves are analyzed to provide true stress–strain constitutive σ(ε) laws for all of these alloys. In the irradiated condition, the σ(ε) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσy) and reductions in uniform strain ductility (eu) are observed, where as the latter can be understood in terms ofmore » the alloy's σ(ε) behavior. Increases in the average σ(ε) in the range of 0–10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are analyzed to establish relations between Δσy and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σyu). The latter shows that higher σyu correlates with lower Δσy. In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher eu than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. As a result, notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.« less

  13. Experimental investigation of stress effect on swelling and microstructure of Fe-16Cr-15Ni-3Mo-Nb austenitic stainless steel under low-temperature irradiation up to high damage dose in the BOR-60 reactor

    NASA Astrophysics Data System (ADS)

    Neustroev, V. S.; Ostrovsky, Z. E.; Shamardin, V. K.

    2004-08-01

    The present paper was devoted to investigation of the stress effect on swelling and microstructure evolution of the Fe-15.8Cr-15.3Ni-2.8Mo-0.6Nb steel irradiated in the BOR-60 reactor at temperatures from 395 to 410 °C and damage doses from 79 to 98 dpa. Was found out that the stress increase leads to an increase of swelling, that can be associated with a decrease in incubation period with a practically constant swelling rate. Voids concentration increases at the first stage of irradiation when the void sizes are practically constant, and then the concentration reaches some saturation and swelling increase is caused by void growth.

  14. 10 CFR 36.33 - Irradiator pools.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Irradiator pools. 36.33 Section 36.33 Energy NUCLEAR... Requirements for Irradiators § 36.33 Irradiator pools. (a) For licenses initially issued after July 1, 1993, irradiator pools must either: (1) Have a water-tight stainless steel liner or a liner...

  15. 10 CFR 36.33 - Irradiator pools.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Irradiator pools. 36.33 Section 36.33 Energy NUCLEAR... Requirements for Irradiators § 36.33 Irradiator pools. (a) For licenses initially issued after July 1, 1993, irradiator pools must either: (1) Have a water-tight stainless steel liner or a liner...

  16. 10 CFR 36.33 - Irradiator pools.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Irradiator pools. 36.33 Section 36.33 Energy NUCLEAR... Requirements for Irradiators § 36.33 Irradiator pools. (a) For licenses initially issued after July 1, 1993, irradiator pools must either: (1) Have a water-tight stainless steel liner or a liner...

  17. Erratum for: Master equation and Fokker-Planck methods for void nucleation and growth in irradiation swelling, Vacancy cluster evolution and swelling in irradiated 316 stainless steel and Radiation swelling behavior and its dependence on temperature, dose

    SciTech Connect

    Surh, M P; Sturgeon, J B; Wolfer, W G

    2005-01-03

    We have recently discovered an error in our void nucleation code used in three prior publications [1-3]. A term was omitted in the model for vacancy re-emission that (especially at high temperature) affects void nucleation and growth during irradiation as well as void annealing and Ostwald ripening of the size distribution after irradiation. The omission was not immediately detected because the calculations predict reasonable void densities and swelling behaviors when compared to experiment at low irradiation temperatures, where void swelling is prominent. (Comparable neutron irradiation experiments are less prevalent at higher temperatures, e.g., > 500 C.)

  18. Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ≈295 °C to ≈6.5 dpa

    SciTech Connect

    Maloy, Stuart A.; Saleh, Tarik A.; Anderoglu, Osman; Romero, Tobias J.; Odette, G. Robert; Yamamoto, Takuya; Li, S.; Cole, James I.; Fielding, Randall

    2015-08-06

    Tensile test results at 25 and 300 °C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress–strain curves are analyzed to provide true stress–strain constitutive σ(ε) laws for all of these alloys. In the irradiated condition, the σ(ε) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσy) and reductions in uniform strain ductility (eu) are observed, where as the latter can be understood in terms of the alloy's σ(ε) behavior. Increases in the average σ(ε) in the range of 0–10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are analyzed to establish relations between Δσy and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σyu). The latter shows that higher σyu correlates with lower Δσy. In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher eu than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. As a result, notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.

  19. Fe-15Ni-13Cr austenitic stainless steels for fission and fusion reactor applications - Part III: Phase stability during heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Lee, E. H.; Mansur, L. K.

    2000-01-01

    The phase stability in Fe-15Ni-13Cr alloys was investigated as a function of minor alloying additions after 4 MeV Ni ion irradiation at 948 K. The results showed that the stability of precipitate phases was dictated mainly by the defects produced by radiation damage and preferential segregation of Si and Ni at defects. In addition, radiation enhanced diffusion and cascade induced dissolution and mixing allowed kinetically sluggish phases to form rapidly under irradiation. These radiation effects caused an enhancement, retardation, or modification of thermal phases, and formation of new phases. The relative stability of precipitate phases varied sensitively with alloy composition. The roles of each alloying element on phase stability and the impact of radiation on the mechanisms of phase evolution were systematically studied and documented. The knowledge obtained from this work provides guidelines for designing alloys that lead to develop desired precipitate microstructures under irradiation.

  20. Accumulation and annealing of radiation defects and the hydrogen effect thereon in an austenitic steel 16Cr15Ni3Mo1Ti upon low-temperature neutron and electron irradiation

    NASA Astrophysics Data System (ADS)

    Arbuzov, V. L.; Gothchitskii, B. N.; Danilov, S. E.; Zaluzhnyi, A. G.; Zuev, Yu. N.; Kar'kin, A. E.; Parkhomenko, V. D.; Sagaradze, V. V.

    2016-01-01

    The effect of hydrogen, accumulation and annealing of radiation defects on the physicomechanical properties of an austenitic Kh16N15M3T1 steel (16Cr15Ni3Mo1Ti) has been investigated upon low-temperature (77 K) neutron and electron irradiations. It has been shown that, when its concentration is about 300 at ppm, hydrogen reduces plasticity by 25%. The presence of helium (2.0-2.5 at ppm) introduced by the tritium-trick method exerts an effect on the yield strength and hardly affects embrittlement. Upon both electron and neutron irradiation, there is a linear relation between the increment of the yield strength and the square root of the increment of the residual electrical resistivity (the concentration of radiation defects). The annealing of vacancies occurs in the neighborhood of 300 K (energy for vacancy migration is 1.0-1.0 eV). Vacancy clusters dissociate near 480 K (energy for dissociation is 1.4-1.5 eV).

  1. Microstructural origins of radiation-induced changes in mechanical properties of 316 L and 304 L austenitic stainless steels irradiated with mixed spectra of high-energy protons and spallation neutrons

    NASA Astrophysics Data System (ADS)

    Sencer, B. H.; Bond, G. M.; Hamilton, M. L.; Garner, F. A.; Maloy, S. A.; Sommer, W. F.

    2001-07-01

    A number of candidate alloys were exposed to a particle flux and spectrum at Los Alamos Neutron Science Center (LANSCE) that closely match the mixed high-energy proton/neutron spectra expected in accelerator production of tritium (APT) window and blanket applications. Austenitic stainless steels 316 L and 304 L are two of these candidate alloys possessing attractive strength and corrosion resistance for APT applications. This paper describes the dose dependence of the irradiation-induced microstructural evolution of SS 316 L and 304 L in the temperature range 30-60°C and consequent changes in mechanical properties. It was observed that the microstructural evolution during irradiation was essentially identical in the two alloys, a behavior mirrored in their changes in mechanical properties. With one expection, it was possible to correlate all changes in mechanical properties with visible microstructural features. A late-term second abrupt decrease in uniform elongation was not associated with visible microstructure, but is postulated to be a consequence of large levels of retained hydrogen measured in the specimens. In spite of large amounts of both helium and hydrogen retained, approaching 1 at.% at the highest exposures, no visible cavities were formed, indicating that the gas atoms were either in solution or in subresolvable clusters.

  2. Modeling precipitation thermodynamics and kinetics in type 316 austenitic stainless steels with varying composition as an initial step toward predicting phase stability during irradiation

    NASA Astrophysics Data System (ADS)

    Shim, Jae-Hyeok; Povoden-Karadeniz, Erwin; Kozeschnik, Ernst; Wirth, Brian D.

    2015-07-01

    The long-term evolution of precipitates in type 316 austenitic stainless steels at 400 °C has been simulated using a numerical model based on classical nucleation theory and the thermodynamic extremum principle. Particular attention has been paid to the precipitation of radiation-induced phases such as γ‧ and G phases. In addition to the original compositions, the compositions for radiation-induced segregation at a dose level of 5, 10 or 20 dpa have been used in the simulation. In a 316 austenitic stainless steel, γ‧ appears as the main precipitate with a small amount of G phase forming at 10 and 20 dpa. On the other hand, G phase becomes relatively dominant over γ‧ at the same dose levels in a Ti-stabilized 316 austenitic stainless steel, which tends to suppress the formation of γ‧. Among the segregated alloying elements, the concentration of Si seems to be the most critical for the formation of radiation-induced phases. An increase in dislocation density as well as increased diffusivity of Mn and Si significantly enhances the precipitation kinetics of the radiation-induced phases within this model.

  3. High energy X-ray diffraction study of the relationship between the macroscopic mechanical properties and microstructure of irradiated HT-9 steel

    NASA Astrophysics Data System (ADS)

    Tomchik, C.; Almer, J.; Anderoglu, O.; Balogh, L.; Brown, D. W.; Clausen, B.; Maloy, S. A.; Sisneros, T. A.; Stubbins, J. F.

    2016-07-01

    Samples harvested from an HT-9 fuel test assembly (ACO-3) irradiated for six years in the Fast Flux Test Facility (FFTF) reaching 2-147 dpa at 382-504 °C were deformed in-situ while collecting high-energy X-ray diffraction data to monitor microstructure evolution. With the initiation of plastic deformation, all samples exhibited a clear load transfer from the ferrite matrix to carbide particulate. This behavior was confirmed by modeling of the control material. The evolution of dislocation density in the material as a result of deformation was characterized through full pattern line profile analysis. The dislocation densities increased substantially after deformation, the level of dislocation evolution observed was highly dependent upon the irradiation temperature of the sample. Differences in both the yield and hardening behavior between samples irradiated at higher and lower temperatures suggest the existence of a transition in tensile behavior at an irradiation temperature near 420 °C dividing regions of distinct damage effects.

  4. Irradiation creep of SA 304L and CW 316 stainless steels: Mechanical behaviour and microstructural aspects. Part II: Numerical simulation and test of SIPA model

    NASA Astrophysics Data System (ADS)

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

    2011-06-01

    A cluster dynamic model has been adapted to test the Stress Induced Preferential Absorption of Defect (SIPA) on Frank loops hypothesis concerning irradiation creep, to reproduce quantitatively both microstructure evolution and its stress induced anisotropy and macroscopic creep rate. It is concluded that SIPA on Frank loops model can account for the observed defects structure, but is unable to reproduce quantitatively the creep rate.

  5. A study of the neutron irradiation effects on the susceptibility to embrittlement of A316L and T91 steels in lead bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Sapundjiev, D.; Al Mazouzi, A.; Van Dyck, S.

    2006-09-01

    The effects of neutron irradiation on the susceptibility to liquid metal embrittlement of two primary selected materials for MYRRHA project an accelerator driven system (ADS), was investigated by means of slow strain rate tests (SSRT). The latter were carried out at 200 °C in nitrogen and in liquid Pb-Bi at a strain rate of 5 × 10 -6 s -1. The small tensile specimens were irradiated at the BR-2 reactor in the MISTRAL irradiation rig at 200 °C for 3 reactor cycles to reach a dose of about 1.50 dpa. The SSR tests were carried out under poor and under dissolved oxygen conditions (˜1.5 × 10 -12 wt% dissolved oxygen) which at this temperature will favour formation of iron and chromium oxides. Although both materials differ in structure (fcc for A316L against bcc for T91), their flow behaviour in contact with liquid lead bismuth eutectic before and after irradiation is very similar. Under these testing conditions none of them was found susceptible to liquid metal embrittlement (LME).

  6. Use of double and triple-ion irradiation to study the influence of high levels of helium and hydrogen on void swelling of 8-12% Cr ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Kupriiyanova, Y. E.; Bryk, V. V.; Borodin, O. V.; Kalchenko, A. S.; Voyevodin, V. N.; Tolstolutskaya, G. D.; Garner, F. A.

    2016-01-01

    In accelerator-driven spallation (ADS) devices, some of the structural materials will be exposed to intense fluxes of very high energy protons and neutrons, producing not only displacement damage, but very high levels of helium and hydrogen. Unlike fission flux-spectra where most helium and hydrogen are generated by transmutation in nickel and only secondarily in iron or chromium, gas production in ADS flux-spectra are rather insensitive to alloy composition, such that Fe-Cr base ferritic alloys also generate very large gas levels. While ferritic alloys are known to swell less than austenitic alloys in fission spectra, there is a concern that high gas levels in fusion and especially ADS facilities may strongly accelerate void swelling in ferritic alloys. In this study of void swelling in response to helium and hydrogen generation, irradiation was conducted on three ferritic-martensitic steels using the Electrostatic Accelerator with External Injector (ESUVI) facility that can easily produce any combination of helium to dpa and/or hydrogen to dpa ratios. Irradiation was conducted under single, dual and triple beam modes using 1.8 MeV Cr+3, 40 keV He+, and 20 keV H+. In the first part of this study we investigated the response of dual-phase EP-450 to variations in He/dpa and H/dpa ratio, focusing first on dual ion studies and then triple ion studies, showing that there is a diminishing influence on swelling with increasing total gas content. In the second part we investigated the relative response of three alloys spanning a range of starting microstructure and composition. In addition to observing various synergisms between He and H, the most important conclusion was that the tempered martensite phase, known to lag behind the ferrite phase in swelling in the absence of gases, loses much of its resistance to void nucleation when irradiated at large gas/dpa levels.

  7. Response of austenitic steels to radiation damage

    SciTech Connect

    Rowcliffe, A.F.; Grossbeck, M.L.

    1983-01-01

    Austenitic stainless steels are prominent contenders as first wall and blanket structural materials for early fusion power reactors. Properties affecting the performance of this class of alloys in the fusion irradiation environment, such as swelling, tensile elongation, irradiation creep, fatigue, and crack growth, have been identified. These properties and the effects of neutron irradiation on them are discussed in this paper. Emphasis is placed on the present status of understanding of irradiation effects.

  8. Validation Analyses of IEAF-2001 Activation Cross-Section Data for SS-316 and F82H Steels Irradiated in a White d-Li Neutron Field

    NASA Astrophysics Data System (ADS)

    Simakov, S. P.; Fischer, U.; v. Möllendorff, U.; Schmuck, I.; Tsige-Tamirat, H.; Wilson, P. P. H.

    2005-05-01

    The evaluated intermediate-energy activation cross-section library IEAF-2001 has been tested against integral experiments with SS-316 and F82H steels exposed to a white neutron flux spectrum extending up to 55 MeV. By making use of the ALARA inventory code the expected γ-active product nuclide inventories were calculated and compared with the measured one. It was found that IEAF-2001 reasonably agrees with experimental data for most of the detected radioisotopes. The reasons for some larger disagreements were found to be the uncertainty of the sample elemental composition, non-validated neutron activation reaction cross sections, and sequential charge particle reactions.

  9. Orientation dependency of mechanical properties of 1950`s vintage Type 304 stainless steel weldment components before and after low temperature neutron irradiation

    SciTech Connect

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

    1992-12-31

    Databases of mechanical properties for both the piping and reactor vessels at the Savannah River Site (SRS) were developed from weldment components (base, weld, and weld heat-affected-zone (HAZ)) of archival piping specimens in the unirradiated and irradiated conditions. Tensile, Charpy V-notch (CVN), and Compact Tension C(T) specimens were tested at 25 and 125C before and after irradiation at low temperatures (90 to 150C) to levels of 0.065 to 2.1 dpa. irradiation hardened the weldment components and reduced the absorbed energy and toughness properties from the unirradiated values. A marked difference in the Charpy V-notch absorbed energy and the elastic-plastic fracture toughness (J{sub IC}) was observed for both the base and HAZ components with the C-L orientation being lower in toughness than the L-C orientation in both the unirradiated and irradiated conditions. Fracture surface examination of the base and HAZ components of unirradiated C(T) specimens showed a ``channel`` morphology in the fracture surfaces of the C-L specimens, whereas equiaxed ductile rupture occurred in the L-C specimens. Chromium carbide precipitation in the HAZ component reduced the fracture toughness of the C-L and L-C specimens compared to the respective base component C-L and L-C specimens. Optical metallography of the piping materials showed stringers of second phase particles parallel to the rolling direction along with a banding or modulation in the microchemistry perpendicular to the pipe axis or rolling direction of the plate material.

  10. Orientation dependency of mechanical properties of 1950's vintage Type 304 stainless steel weldment components before and after low temperature neutron irradiation

    SciTech Connect

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

    1992-01-01

    Databases of mechanical properties for both the piping and reactor vessels at the Savannah River Site (SRS) were developed from weldment components (base, weld, and weld heat-affected-zone (HAZ)) of archival piping specimens in the unirradiated and irradiated conditions. Tensile, Charpy V-notch (CVN), and Compact Tension C(T) specimens were tested at 25 and 125C before and after irradiation at low temperatures (90 to 150C) to levels of 0.065 to 2.1 dpa. irradiation hardened the weldment components and reduced the absorbed energy and toughness properties from the unirradiated values. A marked difference in the Charpy V-notch absorbed energy and the elastic-plastic fracture toughness (J[sub IC]) was observed for both the base and HAZ components with the C-L orientation being lower in toughness than the L-C orientation in both the unirradiated and irradiated conditions. Fracture surface examination of the base and HAZ components of unirradiated C(T) specimens showed a channel'' morphology in the fracture surfaces of the C-L specimens, whereas equiaxed ductile rupture occurred in the L-C specimens. Chromium carbide precipitation in the HAZ component reduced the fracture toughness of the C-L and L-C specimens compared to the respective base component C-L and L-C specimens. Optical metallography of the piping materials showed stringers of second phase particles parallel to the rolling direction along with a banding or modulation in the microchemistry perpendicular to the pipe axis or rolling direction of the plate material.

  11. Welding tritium exposed stainless steel

    SciTech Connect

    Kanne, W.R. Jr.

    1994-11-01

    Stainless steels that are exposed to tritium become unweldable by conventional methods due to buildup of decay helium within the metal matrix. With longer service lives expected for tritium containment systems, methods for welding on tritium exposed material will become important for repair or modification of the systems. Solid-state resistance welding and low-penetration overlay welding have been shown to mitigate helium embrittlement cracking in tritium exposed 304 stainless steel. These processes can also be used on stainless steel containing helium from neutron irradiation, such as occurs in nuclear reactors.

  12. Characteristics of radiation porosity and structural phase state of reactor austenitic 07C-16Cr-19Ni-2Mo-2Mn-Ti-Si-V-P-B Steel after neutron irradiation at a temperature of 440-600°C to damaging doses of 36-94 dpa

    NASA Astrophysics Data System (ADS)

    Portnykh, I. A.; Panchenko, V. L.

    2016-06-01

    The phase composition and the characteristics of vacancy voids in cold-worked steel 07C-16Cr-19Ni-2Mo-2Mn-Ti-Si-V-P-B (CW EK164-ID) after neutron irradiation at damaging doses of 36-94 dpa and temperatures of 440-600°C are investigated. In the entire range of damaging doses and temperatures, voids with different sizes are observed in the material. The maximum void size increases with irradiation temperature up to ~550°C, whereas their concentration decreases. At higher irradiation temperatures, almost no coarse voids are observed. The concentration of fine voids (to 10 nm in size) sharply increases with temperature from 440 to 480°C. Further increases in the temperature do not result in the noticeable concentration growth. In the irradiation temperature range of 440-515°C, second phases precipitate ( G phase, γ' phase, and complex fcc carbides). At higher irradiation temperatures, there are Laves-phase particles, fine second carbides of the MC type, and needle shape precipitates identified as phosphides in the material.

  13. Irradiation creep of dispersion strengthened copper alloy

    SciTech Connect

    Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

    1997-04-01

    Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.

  14. Design of YCF-1 mobile γ irradiator

    NASA Astrophysics Data System (ADS)

    Hehu, Zhang; Chuanzhen, Wang

    1993-07-01

    YCF-1 Mobile irradiator is designed by BINE of China. It has been put into running in YanJi city of Jilin province. It is able to be moved to border and distance places and area lumped and spreading out of agricultural products to service. It can play a important role in demonstration and extending irradiation technology in food irradiation, disinfestation, sterilization and quarantine, etc. This paper describes the features and design considerations of mobile irradiator. This irradiator adopted Cesium-137 source. The design capacity of loading source is 9.25PBq (250kCi), A half-time of Cs- 137 is 30.2 years long, exchanging source is not needed utilization rate of energy is higher, and the shielding is thinner, The Weight is lighter, The dose rate on the surface of it is 0.0025mSv/h in accordance with national standard. The internal size of irradiation room is 1800×1800×900mm (L×W×H), The sheilding of irradiation room is a steel shell filled with lead. The thickness of lead is 18cm. The irradiator is installed on a special flat truck. The size of the truck is 7000×3400×4200mm (L×W×H). The weight of irradiator is more than 80 150kw. The main components and parts of irradiator are: source, source racks and hoist, irradiation chamber, storage source chamber, the product's transport system, dose monitoring system, ventilation system and safety interlock system, etc.

  15. 10 CFR 36.33 - Irradiator pools.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... purification system designed to be capable of maintaining the water during normal operation at a conductivity..., irradiator pools must either: (1) Have a water-tight stainless steel liner or a liner metallurgically... water level that could allow water to drain out of the pool. Pipes that have intakes more than 0.5...

  16. 10 CFR 36.33 - Irradiator pools.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... purification system designed to be capable of maintaining the water during normal operation at a conductivity..., irradiator pools must either: (1) Have a water-tight stainless steel liner or a liner metallurgically... water level that could allow water to drain out of the pool. Pipes that have intakes more than 0.5...

  17. Tissue irradiator

    DOEpatents

    Hungate, F.P.; Riemath, W.F.; Bunnell, L.R.

    1975-12-16

    A tissue irradiator is provided for the in-vivo irradiation of body tissue. The irradiator comprises a radiation source material contained and completely encapsulated within vitreous carbon. An embodiment for use as an in- vivo blood irradiator comprises a cylindrical body having an axial bore therethrough. A radioisotope is contained within a first portion of vitreous carbon cylindrically surrounding the axial bore, and a containment portion of vitreous carbon surrounds the radioisotope containing portion, the two portions of vitreous carbon being integrally formed as a single unit. Connecting means are provided at each end of the cylindrical body to permit connections to blood- carrying vessels and to provide for passage of blood through the bore. In a preferred embodiment, the radioisotope is thulium-170 which is present in the irradiator in the form of thulium oxide. A method of producing the preferred blood irradiator is also provided, whereby nonradioactive thulium-169 is dispersed within a polyfurfuryl alcohol resin which is carbonized and fired to form the integral vitreous carbon body and the device is activated by neutron bombardment of the thulium-169 to produce the beta-emitting thulium-170.

  18. Effect of neutron irradiation on vanadium alloys

    SciTech Connect

    Braski, D.N.

    1986-01-01

    Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600/sup 0/C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520/sup 0/C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys.

  19. Development of ferritic steels for fusion reactor applications

    SciTech Connect

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

    1988-08-01

    Chromium-molybdenum ferritic (martensitic) steels are leading candidates for the structural components for future fusion reactors. However, irradiation of such steels in a fusion environment will produce long-lived radioactive isotopes that will lead to difficult waste-disposal problems. Such problems could be reduced by replacing the elements in the steels (i.e., Mo, Nb, Ni, N, and Cu) that lead to long-lived radioactive isotopes. We have proposed the development of ferritic steels analogous to conventional Cr-Mo steels, which contain molybdenum and niobium. It is proposed that molybdenum be replaced by tungsten and niobium be replaced by tantalum. Eight experimental steels were produced. Chromium concentrations of 2.25, 5, 9, and 12% were used (all concentrations are in wt %). Steels with these chromium compositions, each containing 2% W and 0.25% V, were produced. To determine the effect of tungsten and vanadium, 2.25 Cr steels were produced with 2% W and no vanadium and with 0.25% V and O and 1% W. A 9Cr steel containing 2% W, 0.25 V, and 0.07% Ta was also studied. For all alloys, carbon was maintained at 0.1%. Tempering studies on the normalized steels indicated that the tempering behavior of the new Cr-W steels was similar to that of the analogous Cr-Mo steels. Microscopy studies indicated that 2% tungsten was required in the 2.25 Cr steels to produce 100% bainite in 15.9-mm-thick plate during normalization. The 5Cr and 9Cr steels were 100% martensite, but the 12 Cr steel contained about 75% martensite with the balance delta-ferrite. 33 refs., 35 figs., 5 tabs.

  20. Decontaminating and Melt Recycling Tritium Contaminated Stainless Steel

    SciTech Connect

    Clark, E.A.

    1995-04-03

    The Westinghouse Savannah River Company, Idaho National Engineering Laboratory, and several university and industrial partners are evaluating recycling radioactively contaminated stainless steel. The goal of this program is to recycle contaminated stainless steel scrap from US Department of Energy national defense facilities. There is a large quantity of stainless steel at the DOE Savannah River Site from retired heavy water moderated Nuclear material production reactors (for example heat exchangers and process water piping), that will be used in pilot studies of potential recycle processes. These parts are contaminated by fission products, activated species, and tritium generated by neutron irradiation of the primary reactor coolant, which is heavy (deuterated) water. This report reviews current understanding of tritium contamination of stainless steel and previous studies of decontaminating tritium exposed stainless steel. It also outlines stainless steel refining methods, and proposes recommendations based on this review.

  1. Ion-stimulated gas desorption yields of electropolished, chemically etched, and coated (Au, Ag, Pd, TiZrV) stainless steel vacuum chambers and St707 getter strips irradiated with 4.2 MeV/u lead ions

    NASA Astrophysics Data System (ADS)

    Mahner, E.; Hansen, J.; Küchler, D.; Malabaila, M.; Taborelli, M.

    2005-05-01

    The ion-induced desorption experiment, installed in the CERN Heavy-Ion Accelerator LINAC 3, has been used to measure molecular desorption yields for 4.2 MeV/u lead ions impacting under grazing incidence on different accelerator-type vacuum chambers. Desorption yields for H2, CH4, CO, and CO2, which are of fundamental interest for future accelerator applications, are reported for different stainless steel surface treatments. In order to study the effect of the surface oxide layer on the gas desorption, gold-, silver-, palladium-, and getter-coated 316 LN stainless steel chambers and similarly prepared samples were tested for desorption at LINAC 3 and analyzed for chemical composition by x-ray photoemission spectroscopy. The large effective desorption yield of 2×104 molecules /Pb53+ ion, previously measured for uncoated, vacuum fired stainless steel, was reduced after noble-metal coating by up to 2 orders of magnitude. In addition, pressure rise measurements, the effectiveness of beam scrubbing with lead ions, and the consequence of a subsequent venting on the desorption yields of a beam-scrubbed vacuum chamber are described. Practical consequences for the vacuum system of the future Low Energy Ion Ring are discussed.

  2. Irradiation subassembly

    DOEpatents

    Seim, O.S.; Filewicz, E.C.; Hutter, E.

    1973-10-23

    An irradiation subassembly for use in a nuclear reactor is described which includes a bundle of slender elongated irradiation -capsules or fuel elements enclosed by a coolant tube and having yieldable retaining liner between the irradiation capsules and the coolant tube. For a hexagonal bundle surrounded by a hexagonal tube the yieldable retaining liner may consist either of six segments corresponding to the six sides of the tube or three angular segments each corresponding in two adjacent sides of the tube. The sides of adjacent segments abut and are so cut that metal-tometal contact is retained when the volume enclosed by the retaining liner is varied and Springs are provided for urging the segments toward the center of the tube to hold the capsules in a closely packed configuration. (Official Gazette)

  3. Temperature Effects on the Mechanical Properties of Candidate SNS Target Container Materials after Proton and Neutron Irradiation

    SciTech Connect

    Byun, T.S.

    2001-11-09

    This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54 to 2.53 dpa. Irradiation temperatures were in the range 30 to 100 C. Tensile testing was performed at room temperature (20 C) and 164 C to study the effects of test temperature on the tensile properties. Test materials displayed significant radiation-induced hardening and loss of ductility due to irradiation. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative strain hardening. In the EC316LN stainless steel, increasing the test temperature from 20 C to 164 C decreased the strength by 13 to 18% and the ductility by 8 to 36%. The tensile data for the EC316LN stainless steel irradiated in spallation conditions were in line with the values in a database for 316 stainless steels for doses up to 1 dpa irradiated in fission reactors at temperatures below 200 C. However, extra strengthening induced by helium and hydrogen contents is evident in some specimens irradiated to above about 1 dpa. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. It was estimated that the 316 stainless steels would retain more than 1% true stains to necking at 164 C after irradiation to 5 dpa. A calculation using reduction of area (RA) measurements and stress-strain data predicted positive strain hardening during plastic instability.

  4. Irradiated foods

    MedlinePlus

    ... it reduces the risk of food poisoning . Food irradiation is used in many countries. It was first approved in the U.S. to prevent sprouts on white potatoes, and to control insects on wheat and in certain spices and seasonings.

  5. Effect of re-irradiation by neutrons on mechanical properties of un-irradiated/irradiated SS316LN weldments

    NASA Astrophysics Data System (ADS)

    Tsuchiya, K.; Shimizu, M.; Kawamura, H.; Kalinin, G.

    2008-02-01

    Stainless steel of type SS316LN-IG (ITER Grade) is used for the branch pipeline connecting of the module coolant system and for other structures of ITER. One of the most important requirements for the branch pipeline connection is to recover various defects by welding. In the present study, characteristics of irradiated weldments were evaluated. SS316LN-IG specimens irradiated to helium contents of 3 and 10 appm He were prepared by the first neutron irradiation. Thereafter, the SS316LN-IG specimens with three different combinations of un-irradiation and irradiation were welded by a tungsten inert-gas (TIG) welding method. These weldments were re-irradiated at 150 °C up to a fast neutron fluence of about 7.5 × 10 24 n/m 2 ( E > 1 MeV). Tensile tests of the weldments and the base material were carried out at 20 and 150 °C after the re-irradiation. The results of the comparison before and after the re-irradiation showed that tensile properties of all weldment specimens with the different combinations were almost the same as those of the base materials.

  6. Development of a triple beam irradiation facility

    NASA Astrophysics Data System (ADS)

    Hamada, S.; Miwa, Y.; Yamaki, D.; Katano, Y.; Nakazawa, T.; Noda, K.

    1998-10-01

    A triple beam ion irradiation facility has been developed to study the synergistic effects of displacement damage, helium and hydrogen atoms on microstructural changes of materials under irradiation environments simulating a fusion reactor. The system consists of a vacuum chamber and three beamlines, which are connected with each electrostatic accelerator. Samples can be irradiated in the wide temperature range from liquid nitrogen to 1273 K in the chamber by replacing two kinds of sample stages alternatively. An austenitic stainless steel was simultaneously irradiated with triple beam of nickel, helium and hydrogen ions at 573-673 K using this facility and TEM observations were carried out from a cross sectional view normal to the incident surface. It was shown that the number density of dislocation loops decreased in the region where hydrogen and helium were deposited in comparison with ones in the region where only displacement damage was induced to a similar damage level.

  7. Embrittlement of RPV steels; An atom probe tomography perspective

    SciTech Connect

    Miller, Michael K; Russell, Kaye F

    2007-01-01

    Atom probe tomography has played a key role in the understanding of the embrittlement of neutron irradiated reactor pressure vessel steels through the atomic level characterization of the microstructure. Atom probe tomography has been used to demonstrate the importance of the post weld stress relief treatment in reducing the matrix copper content in high copper alloys, the formation of {approx}-nm-diameter copper-, nickel-, manganese- and silicon-enriched precipitates during neutron irradiation in copper containing RPV steels, and the coarsening of these precipitates during post irradiation heat treatments. Atom probe tomography has been used to detect {approx}2-nm-diameter nickel-, silicon- and manganese-enriched clusters in neutron irradiated low copper and copper free alloys. Atom probe tomography has also been used to quantify solute segregation to, and precipitation on, dislocations and grain boundaries.

  8. Ultrahigh carbon steels, Damascus steels, and superplasticity

    SciTech Connect

    Sherby, O.D.; Wadsworth, J.

    1997-04-01

    The processing properties of ultrahigh carbon steels (UHCSs) have been studied at Stanford University over the past twenty years. These studies have shown that such steels (1 to 2.1% C) can be made superplastic at elevated temperature and can have remarkable mechanical properties at room temperature. It was the investigation of these UHCSs that eventually brought us to study the myths, magic, and metallurgy of ancient Damascus steels, which in fact, were also ultrahigh carbon steels. These steels were made in India as castings, known as wootz, possibly as far back as the time of Alexander the Great. The best swords are believed to have been forged in Persia from Indian wootz. This paper centers on recent work on superplastic UHCSs and on their relation to Damascus steels. 32 refs., 6 figs.

  9. First principle-based AKMC modelling of the formation and medium-term evolution of point defect and solute-rich clusters in a neutron irradiated complex Fe-CuMnNiSiP alloy representative of reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Ngayam-Happy, R.; Becquart, C. S.; Domain, C.

    2013-09-01

    The formation and medium-term evolution of point defect and solute-rich clusters under neutron irradiation have been modelled in a complex Fe-CuMnNiSiP alloy representative of RPV steels, by means of first principle-based atomistic kinetic Monte Carlo simulations. The results obtained reproduce most features observed in available experimental studies, highlighting the very good agreement between both series. According to simulation, solute-rich clusters form and develop via an induced segregation mechanism on either the vacancy or interstitial clusters, and these point defect clusters are efficiently generated only in cascade debris and not Frenkel pair flux. The results have revealed the existence of two distinct populations of clusters with different characteristic features. Solute-rich clusters in the first group are bound essentially to interstitial clusters and they are enriched in Mn mostly, but also Ni to a lesser extent. Over the low dose regime, their density increases in the alloy as a result of the accumulation of highly stable interstitial clusters. In the second group, the solute-rich clusters are merged with vacancy clusters, and they contain mostly Cu and Si, but also substantial amount of Mn and Ni. The formation of a sub-population of pure solute clusters has been observed, which results from annihilation of the low stable vacancy clusters on sinks. The results indicate finally that the Mn content in clusters is up to 50%, Cu, Si, and Ni sharing the other half in more or less equivalent amounts. This composition has not demonstrated any noticeable modification with increasing dose over irradiation.

  10. Quantitative TEM analysis of precipitation and grain boundary segregation in neutron irradiated EUROFER 97

    NASA Astrophysics Data System (ADS)

    Dethloff, Christian; Gaganidze, Ermile; Aktaa, Jarir

    2014-11-01

    Characterization of irradiation induced microstructural defects is essential for assessing the applicability of structural steels like the Reduced Activation Ferritic/Martensitic steel EUROFER 97 in upcoming fusion reactors. In this work Transmission Electron Microscopy (TEM) is used to analyze the types and structure of precipitates, and the evolution of their size distributions and densities caused by neutron irradiation to a dose of 32 displacements per atom (dpa) at 330-340 °C in the irradiation experiment ARBOR 1. A significant growth of MX and M23C6 type precipitates is observed after neutron irradiation, while the precipitate density remains unchanged. Hardening caused by MX and M23C6 precipitate growth is assessed by applying the Dispersed Barrier Hardening (DBH) model, and shown to be of minor importance when compared to other irradiation effects like dislocation loop formation. Additionally, grain boundary segregation of chromium induced by neutron irradiation was investigated and detected in irradiated specimens.

  11. Advances in the Hopkinson bar testing of irradiated/non-irradiated nuclear materials and large specimens.

    PubMed

    Albertini, Carlo; Cadoni, Ezio; Solomos, George

    2014-05-13

    A brief review of the technological advances of the Hopkinson bar technique in tension for the study of irradiated/non-irradiated nuclear materials and the development of this technology for large specimens is presented. Comparisons are made of the dynamic behaviour of non-irradiated and irradiated materials previously subjected to creep, low cycle fatigue and irradiation (2, 10 and 30 displacements per atom). In particular, complete results of the effect of irradiation on the dynamic mechanical properties of AISI304L steel, tested at 20, 400 and 550°C are presented. These high strain rate tests have been performed with a modified Hopkinson bar (MHB), installed inside a hot cell. Examples of testing large nuclear steel specimens with a very large Hopkinson bar are also shown. The results overall demonstrate the capability of the MHB to efficiently reproduce the material stress conditions in case of accidental internal and external dynamic loadings in nuclear reactors, thus contributing to the important process of their structural assessment.

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

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

    1998-03-01

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

  13. Current status and recent research achievements in ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Tavassoli, A.-A. F.; Diegele, E.; Lindau, R.; Luzginova, N.; Tanigawa, H.

    2014-12-01

    When the austenitic stainless steel 316L(N) was selected for ITER, it was well known that it would not be suitable for DEMO and fusion reactors due to its irradiation swelling at high doses. A parallel programme to ITER collaboration already had been put in place, under an IEA fusion materials implementing agreement for the development of a low activation ferritic/martensitic steel, known for their excellent high dose irradiation swelling resistance. After extensive screening tests on different compositions of Fe-Cr alloys, the chromium range was narrowed to 7-9% and the first RAFM was industrially produced in Japan (F82H: Fe-8%Cr-2%W-TaV). All IEA partners tested this steel and contributed to its maturity. In parallel several other RAFM steels were produced in other countries. From those experiences and also for improving neutron efficiency and corrosion resistance, European Union opted for a higher chromium lower tungsten grade, Fe-9%Cr-1%W-TaV steel (Eurofer), and in 1997 ordered the first industrial heats. Other industrial heats have been produced since and characterised in different states, including irradiated up to 80 dpa. China, India, Russia, Korea and US have also produced their grades of RAFM steels, contributing to overall maturity of these steels. This paper reviews the work done on RAFM steels by the fusion materials community over the past 30 years, in particular on the Eurofer steel and its design code qualification for RCC-MRx.

  14. The effects of low dose rate irradiation and thermal aging on reactor structural alloys

    NASA Astrophysics Data System (ADS)

    Allen, T. R.; Trybus, C. L.; Cole, J. I.

    As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10 -8 dpa/s) irradiation at 380-410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.

  15. Alloy development for irradiation performance. Quarterly progress report for period ending December 31, 1980

    SciTech Connect

    Not Available

    1981-04-01

    Progress is reported in eight sections: analysis and evaluation studies, test matrices and test methods development, Path A Alloy Development (austenitic stainless steels), Path C Alloy Development (Ti and V alloys), Path D Alloy Development (Fe alloys), Path E Alloy Development (ferritic steels), irradiation experiments and materials inventory, and materials compatibility and hydrogen permeation studies. (DLC)

  16. Alloy development for irradiation performance. Quarterly progress report for period ending December 31, 1979

    SciTech Connect

    Ashdown, B.G.

    1980-04-01

    Progress is reported concerning preparation of a materials handbook for fusion, creep-fatigue of first-wall structural materials, test results on miniature compact tension fracture toughness specimens, austenitic stainless steels, Fe-Ni-Cr alloys, iron-base alloys with long-range crystal structure, ferritic steels, irradiation experiments, corrosion testing, and hydrogen permeation studies. (FS)

  17. Aging and Embrittlement of High Fluence Stainless Steels

    SciTech Connect

    Was, gary; Jiao, Zhijie; der ven, Anton Van; Bruemmer, Stephen; Edwards, Dan

    2012-12-31

    Irradiation of austenitic stainless steels results in the formation of dislocation loops, stacking fault tetrahedral, Ni-Si clusters and radiation-induced segregation (RIS). Of these features, it is the formation of precipitates which is most likely to impact the mechanical integrity at high dose. Unlike dislocation loops and RIS, precipitates exhibit an incubation period that can extend from 10 to 46 dpa, above which the cluster composition changes and a separate phase, (G-phase) forms. Both neutron and heavy ion irradiation showed that these clusters develop slowly and continue to evolve beyond 100 dpa. Overall, this work shows that the irradiated microstructure features produced by heavy ion irradiation are remarkably comparable in nature to those produced by neutron irradiation at much lower dose rates. The use of a temperature shift to account for the higher damage rate in heavy ion irradiation results in a fairly good match in the dislocation loop microstructure and the precipitate microstructure in austenitic stainless steels. Both irradiations also show segregation of the same elements and in the same directions, but to achieve comparable magnitudes, heavy ion irradiation must be conducted at a much higher temperature than that which produces a match with loops and precipitates. First-principles modeling has confirmed that the formation of Ni-Si precipitates under irradiation is likely caused by supersaturation of solute to defect sinks caused by highly correlated diffusion of Ni and Si. Thus, the formation and evolution of Ni-Si precipitates at high dose in austenitic stainless steels containing Si is inevitable.

  18. Phase Stability under Irradiation of Precipitates and Solid Solutions in Model ALloys and in ODS Alloys Relevant for Gen IV

    SciTech Connect

    Arthur T. Motta; Robert C. Birtcher

    2007-10-17

    The overall objective of this program is to investigate the irradiation-altered phase stability of oxide precipitates in ODS steels and of model alloy solid solutions of associated systems. This information can be used to determine whether the favorable mechanical propertiies of these steels are maintained under irradiation, thus addressing one of the main materials research issues for this class of steels as identified by the GenIV working groups. The research program will also create fundamental understanding of the irradiation precipitation/dissolution problem by studying a "model" system in which the variables can be controlled and their effects understood individually.

  19. Mechanical behavior of AISI 304SS determined by miniature test methods after neutron irradiation to 28 dpa

    SciTech Connect

    Ellen M. Rabenberg; Brian J. Jaques; Bulent H. Sencer; Frank A. Garner; Paula D. Freyer; Taira Okita; Darryl P. Butt

    2014-05-01

    The mechanical properties of AISI 304 stainless steel irradiated for over a decade in the Experimental Breeder Reactor (EBR-II) were measured using miniature mechanical testing methods. The shear punch method was used to evaluate the shear strengths of the neutron-irradiated steel and a correlation factor was empirically determined to predict its tensile strength. The strength of the stainless steel slightly decreased with increasing irradiation temperature, and significantly increased with increasing dose until it saturated above approximately 5 dpa. Ferromagnetic measurements were used to observe and deduce the effects of the stress-induced austenite to martensite transformation as a result of shear punch testing.

  20. The steel scrap age.

    PubMed

    Pauliuk, Stefan; Milford, Rachel L; Müller, Daniel B; Allwood, Julian M

    2013-04-01

    Steel production accounts for 25% of industrial carbon emissions. Long-term forecasts of steel demand and scrap supply are needed to develop strategies for how the steel industry could respond to industrialization and urbanization in the developing world while simultaneously reducing its environmental impact, and in particular, its carbon footprint. We developed a dynamic stock model to estimate future final demand for steel and the available scrap for 10 world regions. Based on evidence from developed countries, we assumed that per capita in-use stocks will saturate eventually. We determined the response of the entire steel cycle to stock saturation, in particular the future split between primary and secondary steel production. During the 21st century, steel demand may peak in the developed world, China, the Middle East, Latin America, and India. As China completes its industrialization, global primary steel production may peak between 2020 and 2030 and decline thereafter. We developed a capacity model to show how extensive trade of finished steel could prolong the lifetime of the Chinese steelmaking assets. Secondary steel production will more than double by 2050, and it may surpass primary production between 2050 and 2060: the late 21st century can become the steel scrap age.

  1. The steel scrap age.

    PubMed

    Pauliuk, Stefan; Milford, Rachel L; Müller, Daniel B; Allwood, Julian M

    2013-04-01

    Steel production accounts for 25% of industrial carbon emissions. Long-term forecasts of steel demand and scrap supply are needed to develop strategies for how the steel industry could respond to industrialization and urbanization in the developing world while simultaneously reducing its environmental impact, and in particular, its carbon footprint. We developed a dynamic stock model to estimate future final demand for steel and the available scrap for 10 world regions. Based on evidence from developed countries, we assumed that per capita in-use stocks will saturate eventually. We determined the response of the entire steel cycle to stock saturation, in particular the future split between primary and secondary steel production. During the 21st century, steel demand may peak in the developed world, China, the Middle East, Latin America, and India. As China completes its industrialization, global primary steel production may peak between 2020 and 2030 and decline thereafter. We developed a capacity model to show how extensive trade of finished steel could prolong the lifetime of the Chinese steelmaking assets. Secondary steel production will more than double by 2050, and it may surpass primary production between 2050 and 2060: the late 21st century can become the steel scrap age. PMID:23442209

  2. Methods of forming steel

    DOEpatents

    Branagan, Daniel J.; Burch, Joseph V.

    2001-01-01

    In one aspect, the invention encompasses a method of forming a steel. A metallic glass is formed and at least a portion of the glass is converted to a crystalline steel material having a nanocrystalline scale grain size. In another aspect, the invention encompasses another method of forming a steel. A molten alloy is formed and cooled the alloy at a rate which forms a metallic glass. The metallic glass is devitrified to convert the glass to a crystalline steel material having a nanocrystalline scale grain size. In yet another aspect, the invention encompasses another method of forming a steel. A first metallic glass steel substrate is provided, and a molten alloy is formed over the first metallic glass steel substrate to heat and devitrify at least some of the underlying metallic glass of the substrate.

  3. Corrosion behavior of carbon steels under tuff repository environmental conditions

    SciTech Connect

    McCright, R.D.; Weiss, H.

    1984-10-01

    Carbon steels may be used for borehole liners in a potential high-level nuclear waste repository in tuff in Nevada. Borehole liners are needed to facilitate emplacement of the waste packages and to facilitate retrieval of the packages, if required. Corrosion rates of low carbon structural steels AISI 1020 and ASTM A-36 were determined in J-13 well water and in saturated steam at 100{sup 0}C. Tests were conducted in air-sparged J-13 water to attain more oxidizing conditions representative of irradiated aqueous environments. A limited number of irradiation corrosion and stress corrosion tests were performed. Chromium-molybdenum alloy steels and cast irons were also tested. These materials showed lower general corrosion but were susceptible to stress corrosion cracking when welded. 4 references, 4 tables.

  4. Irradiation effect on mechanical properties in structural materials of fast breeder reactor plant

    NASA Astrophysics Data System (ADS)

    Nagae, Yuji; Takaya, Shigeru; Wakai, Eiichi; Aoto, Kazumi

    2011-07-01

    The effects of displacement per atom (dpa) level, helium content, and the ratio of helium content to dpa level on the tensile and creep properties have been investigated in the assumed irradiation damage range of FBR structural materials. The assumed irradiation damage range is up to about 1 dpa and about 30 appm for helium content. Austenitic stainless steel and high-chromium martensitic steel are considered as FBR structural materials. As a result, it is shown that the dpa level is a promising index for evaluating neutron irradiation damage.

  5. The compositional dependence of irradiation creep of austenitic alloys irradiated in PFR at 420{degrees}C

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Munro, B.

    1997-04-01

    Irradiation creep data are expensive and often difficult to obtain, especially when compared to swelling data. This requires that maximum use be made of available data sources in order to elucidate the parametric dependencies of irradiation creep for application to new alloys and to new environments such as those of proposed fusion environments. One previously untapped source of creep data is that of a joint U.S./U.K. experiment conducted in the Prototype Fast Reactor (PFR) in Dounreay, Scotland. In this experiment, five austenitic steels were irradiated in a variety of starting conditions. In particular, these steels spanned a large range (15-40%) of nickel contents, and contained strong variations in Mo, Ti, Al, and Nb. Some alloys were solution-strengthened and some were precipitation-strengthened. Several were cold-worked. These previously unanalyzed data show that at 420{degrees}C all austenitic steels have a creep compliance that is roughly independent of the composition of the steel at 2{+-}1 x 10{sup {minus}6}MPa{sup {minus}1} dpa{sup {minus}1}. The variation within this range may arise from the inability to completely separate the non-creep strains arising from precipitation reactions and the stress-enhancement of swelling. Each of these can be very sensitive to the composition and starting treatment of a steel.

  6. Use of Irradiated Foods

    NASA Technical Reports Server (NTRS)

    Brynjolfsson, A.

    1985-01-01

    The safety of irradiated foods is reviewed. Guidelines and regulations for processing irradiated foods are considered. The radiolytic products formed in food when it is irradiated and its wholesomeness is discussed. It is concluded that food irradiation processing is not a panacea for all problems in food processing but when properly used will serve the space station well.

  7. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures: Summary report

    SciTech Connect

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

    1996-04-01

    Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 250{degrees}C. These specimens have been tested over a temperature range from 20 to 250{degrees}C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenic stainless steels, but the toughness remains quite high. The toughness decreases as the temperature increases. Irradiation at 250{degrees}C is more damaging that at 90{degrees}C, causing larger decreases in the fracture toughness. The ferritic-martensitic steels HT-9 and F82H show significantly greater reductions in fracture toughness that the austenitic stainless steels.

  8. Reweldability test of irradiated SS316 by the TIG welding method

    NASA Astrophysics Data System (ADS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Oyamada, Rokuro

    1996-10-01

    Stainless steel is a candidate material for the structural material in fusion reactors. Rewelding of irradiated materials will have a large impact on the design and the maintenance of in-vessel components. In the present work, the welding specimens made of type 316 stainless steel were irradiated in JMTR (Japan materials testing reactor) to a fast neutron fluence of ˜2.0 × 10 20 n/cm 2 ( E > 1 MeV) at a temperature of ˜200°C. The rewelding of unirradiated and/or irradiated stainless steel was performed by the tungsten inert gas (TIG) welding method and the weldments of unirradiated and/or irradiated SS316 were characterized by tensile testing (test temp.: 20°C and 200°C), hardness, metallographical observation and SEM/XMA analyses.

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

    SciTech Connect

    Dai, Y.

    1996-06-01

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

  10. Detection of irradiated liquor

    NASA Astrophysics Data System (ADS)

    Shengchu, Qi; Jilan, Wu; Rongyao, Yuan

    D-2,3-butanediol is formed by irradiation processes in irradiated liquors. This radiolytic product is not formed in unirradiated liquors and its presence can therefore be used to identify whether a liquor has been irradiated or not. The relation meso/dl≈1 for 2,3-butanediol and the amount present in irradiated liquors may therefore be used as an indication of the dose used in the irradiation.

  11. In-service irradiated and aged material evaluations

    SciTech Connect

    Haggag, F.M.; Nanstad, R.K.; Alexander, D.J.

    1995-10-01

    The objective of this task is to provide a direct assessment of actual material properties in irradiated components of nuclear reactors, including the effects of irradiation and aging. Four activities are currently in progress: (1) establishing a machining capability for contaminated or activated materials by completing procurement and installation of a computer-based milling machine in a hot cell; (2) machining and testing specimens from cladding materials removed from the Gundremmingen reactor to establish their fracture properties; (3) preparing an interpretive report on the effects of neutron irradiation on cladding; and (4) continuing the evaluation of long-term aging of austenitic structural stainless steel weld metal by metallurgically examining and testing specimens aged at 288 and 343{degrees}C and reporting the results, as well as by continuing the aging of the stainless steel cladding toward a total time of 50,000 h.

  12. Multipulse nanosecond laser modification of steel surface

    NASA Astrophysics Data System (ADS)

    Chumakov, A. N.; Nikonchuk, I. S.; Gaković, B.; Petrović, S.; Trtica, M.

    2014-09-01

    Results of surface modification are presented for MnNiCrMo-steel samples exposed to a Nd:YAG laser operating in a pulse-periodic mode (10 Hz frequency, 532 nm wavelength and 17 ns pulse duration). The steel samples were irradiated in air by a series of laser pulses at a fluence of 10.7 J cm-2 close to a plasma formation threshold. Surface structures were examined by optical, scanning electron and confocal optical microscopy. The appearance of the detected surface structures strongly depends on the number of laser pulses and power density of laser radiation. Significant differences were found between laser-induced structures in the center of the laser spot, at its edges and in the nearest surrounding of the laser spot. The reasons for such differences are discussed.

  13. Thermal annealing as a method to predict results of high temperature irradiation embrittlement

    NASA Astrophysics Data System (ADS)

    Kryukov, A.; Debarberis, L.; Hähner, P.; Gillemot, F.; Oszvald, F.

    2013-01-01

    In order to assess the validity of post-irradiation annealing as a method to predict results of high temperature irradiation a new analysis of experimental data has been performed revealing the combined influence of annealing temperature and impurities content on residual embrittlement after annealing. For 2CrMoV (WWER-440 reactor pressure vessel) steel with low contents of copper and phosphorus, the comparison of two embrittlement dependencies has been done: on irradiation temperature and post-irradiation annealing temperature. It is demonstrated that data for both the transition temperature shift after irradiation, ΔTk, and the residual transition temperature shift after post-irradiation annealing, ΔTres, fall within the same scatter band. A similarly close correlation is observed by comparison of yield strength increases after irradiation and after post-irradiation annealing.

  14. Final Report on MEGAPIE Target Irradiation and Post-Irradiation Examination

    SciTech Connect

    Yong, Dai

    2015-06-30

    Megawatt pilot experiment (MEGAPIE) was successfully performed in 2006. One of the important goals of MEGAPIE is to understand the behaviour of structural materials of the target components exposed to high fluxes of high-energy protons and spallation neutrons in flowing LBE (liquid lead-bismuth eutectic) environment by conducting post-irradiation examination (PIE). The PIE includes four major parts: non-destructive test, radiochemical analysis of production and distribution of radionuclides produced by spallation reaction in LBE, analysis of LBE corrosion effects on structural materials, T91 and SS 316L steels, and mechanical testing of the T91 and SS 316L steels irradiated in the lower part of the target. The non-destructive test (NDT) including visual inspection and ultrasonic measurement was performed in the proton beam window area of the T91 calotte of the LBE container, the most intensively irradiated part of the MEGAPIE target. The visual inspection showed no visible failure and the ultrasonic measurement demonstrated no detectable change in thickness in the beam window area. Gamma mapping was also performed in the proton beam window area of the AlMg3 safety-container. The gamma mapping results were used to evaluate the accumulated proton fluence distribution profile, the input data for determining irradiation parameters. Radiochemical analysis of radionuclides produced by spallation reaction in LBE is to improve the understanding of the production and distribution of radionuclides in the target. The results demonstrate that the radionuclides of noble metals, 207Bi, 194Hg/Au are rather homogeneously distributed within the target, while radionuclides of electropositive elements are found to be deposited on the steel-LBE interface. The corrosion effect of LBE on the structural components under intensive irradiation was investigated by metallography. The results show that no evident corrosion damages. However, unexpected deep

  15. Emulation of reactor irradiation damage using ion beams

    SciTech Connect

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide, irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.

  16. Emulation of reactor irradiation damage using ion beams

    DOE PAGES

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide,more » irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

  17. DOE uses transportable irradiator for demonstration and testing

    SciTech Connect

    Not Available

    1988-12-01

    The U.S. Dept. of Energy's Pacific Northwest Laboratory (PNL), Richland, Washington (operated by Battelle Memorial Institute), has a transportable irradiator that was built to travel to various locations to demonstrate the benefits of low-dose irradiation for the processing of food. Part of a DOE program designed to establish irradiation facilities in Alaska, Florida, Hawaii, Iowa, Oklahoma, and Washington, the mobile unit can also be used for research, pilot-scale processing, operator training, and education. The irradiation unit consists of two lead-lined cylindrical chambers-an irradiation chamber and a source chamber-inside a steel casing. During operation, the item to be irradiated is placed inside the irradiation chamber, which is then rotated until a window in the chamber lines up with a screened window in the source chamber. The source chamber contains the transportation cask containing the four capsules of cesium-137 that are used as the source of gamma radiation. During operation, the lid of the cask is raised, pulling the capsules into operating position. In this alignment, the product is irradiated for a predetermined length of time. Then the lid of the cask is lowered and the irradiation chamber is rotated back to its original position for removal of the product.

  18. Development status of CLAM steel for fusion application

    NASA Astrophysics Data System (ADS)

    Huang, Qunying

    2014-12-01

    The China low activation martensitic (CLAM) steel is being developed at the Institute of Nuclear Energy Safety Technology (INEST) under wide collaboration within China. Significant R&D work on CLAM steel was carried out to help make it suitable for industrial applications. The effect of refining processes and thermal aging on composition, microstructures and mechanical properties were investigated. Material properties before irradiation including impact, fracture toughness, thermal aging, creep and fatigue properties etc. were assessed. A series of irradiation tests in the fission reactor HFETR in Chengdu up to 2 dpa and in the spallation neutron source SINQ in Paul Scherrer Institute up to 20 dpa were performed. PbLi corrosion tests for more than 10,000 h were done in the DRAGON-I and PICOLO loops. Fabrication techniques for a test blanket module (TBM) are being developed and a 1/3 scale TBM prototype is being fabricated with CLAM steel. Recent progresses on the development status of this steel are presented here. The code qualification of CLAM steel is under plan for its final application in ITER-TBM and DEMO in the future.

  19. Modern Steel Framed Schools.

    ERIC Educational Resources Information Center

    American Inst. of Steel Construction, Inc., New York, NY.

    In view of the cost of structural framing for school buildings, ten steel-framed schools are examined to review the economical advantages of steel for school construction. These schools do not resemble each other in size, shape, arrangement or unit cost; some are original in concept and architecture, and others are conservative. Cost and…

  20. Steel Industry Wastes.

    ERIC Educational Resources Information Center

    Schmidtke, N. W.; Averill, D. W.

    1978-01-01

    Presents a literature review of wastes from steel industry, covering publications of 1976-77. This review covers: (1) coke production; (2) iron and steel production; (3) rolling operations; and (4) surface treatment. A list of 133 references is also presented. (NM)

  1. The Steel Band.

    ERIC Educational Resources Information Center

    Weil, Bruce

    1996-01-01

    Describes studying the steel drum, an import from Trinidad, as an instrument of intellectual growth. Describes how developing a steel drum band provided Montessori middle school students the opportunity to experience some important feelings necessary to emotional growth during this difficult age: competence, usefulness, independence, and…

  2. Microstructure evolution in austenitic Fe-Cr-Ni alloys irradiated with rotons: comparison with neutron-irradiated microstructures

    NASA Astrophysics Data System (ADS)

    Gan, J.; Was, G. S.

    2001-08-01

    Irradiation-induced microstructures of high purity and commercial purity austenitic stainless steels were investigated using proton-irradiation. For high purity alloys, Fe-20Cr-9Ni (HP 304 SS), Fe-20Cr-24Ni and Ni-18Cr-9Fe were irradiated using 3.2 MeV protons between 300°C and 600°C at a dose rate of 7×10 -6 dpa/ s to doses up to 3.0 dpa. The commercial purity alloys, CP 304 SS and CP 316 SS were irradiated at 360°C to doses between 0.3 and 5.0 dpa. The dose, temperature and composition dependence of the number density and size of dislocation loops and voids were characterized. The changes in yield strength due to irradiation were estimated from Vickers hardness measurements and compared to calculations using a dispersed-barrier-hardening (DBH) model. The dose and temperature dependence of proton-irradiated microstructure (loops, voids) and the irradiation hardening are consistent with the neutron-data trend. Results indicate that proton-irradiation can accurately reproduce the microstructure of austenitic alloys irradiated in LWR cores.

  3. EAST ELEVATION, LTV STEEL (FORMERLY REPUBLIC STEEL), 8" BAR MILL, ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    EAST ELEVATION, LTV STEEL (FORMERLY REPUBLIC STEEL), 8" BAR MILL, BUFFALO PLANT. VIEW LOOKING SOUTHWEST FROM ROLL SHOP. 8" BAR MILL DESIGNED AND BUILT BY DONNER STEEL CO. (PREDECESSOR OF REPUBLIC), 1919-1920. FOR DESCRIPTION OF ORIGINAL MILL SEE "IRON AGE", 116\\4 (23 JULY 1925): 201-204. - LTV Steel, 8-inch Bar Mill, Buffalo Plant, Buffalo, Erie County, NY

  4. Irradiation creep relaxation of void swelling-driven stresses

    NASA Astrophysics Data System (ADS)

    Hall, M. M.

    2013-01-01

    Swelling-driven-creep test specimens are used to measure the compressive stresses that develop due to constraint of irradiation void swelling. These specimens use a previously non-irradiated 20% CW Type 316 stainless steel holder to axially restrain two Type 304 stainless steel tubular specimens that were previously irradiated in the US Experimental Breeder Reactor (EBR-II) at 490 °C. One specimen was previously irradiated to fluence levels in the void nucleation regime (9 dpa) and the other in the quasi-steady void growth regime (28 dpa). A lift-off compliance measurement technique was used post-irradiation to determine compressive stresses developed during reirradiation of the two specimen assemblies in Row 7 of EBR-II at temperatures of 547 °C and 504 °C, respectively, to additional damage levels each of about 5 dpa. Results obtained on the higher fluence swelling-driven-creep specimen show that compressive stress due to constraint of swelling retards void swelling to a degree that is consistent with active load uniaxial compression specimens that were irradiated as part of a previously reported multiaxial in-reactor creep experiment. Swelling results obtained on the lower fluence swelling-driven creep specimen show a much larger effect of compressive stress in reducing swelling, demonstrating that the larger effect of stress on swelling is on void nucleation as compared to void growth. Test results are analyzed using a recently proposed multiaxial creep-swelling model.

  5. Tailoring plasticity of austenitic stainless steels for nuclear applications: Review of mechanisms controlling plasticity of austenitic steels below 400 °C

    NASA Astrophysics Data System (ADS)

    Meric de Bellefon, G.; van Duysen, J. C.

    2016-07-01

    AISI 304 and 316 austenitic stainless steels were invented in the early 1900s and are still trusted by materials and mechanical engineers in numerous sectors because of their good combination of strength, ductility, and corrosion resistance, and thanks to decades of experience and data. This article is part of an effort focusing on tailoring the plasticity of both types of steels to nuclear applications. It provides a synthetic and comprehensive review of the plasticity mechanisms in austenitic steels during tensile tests below 400 °C. In particular, formation of twins, extended stacking faults, and martensite, as well as irradiation effects and grain rotation are discussed in details.

  6. Tailoring plasticity of austenitic stainless steels for nuclear applications: Review of mechanisms controlling plasticity of austenitic steels below 400 °C

    NASA Astrophysics Data System (ADS)

    Meric de Bellefon, G.; van Duysen, J. C.

    2016-07-01

    AISI 304 and 316 austenitic stainless steels were invented in the early 1900s and are still trusted by materials and mechanical engineers in numerous sectors because of their good combination of strength, ductility, and corrosion resistance, and thanks to decades of experience and data. This article is part of an effort focusing on tailoring the plasticity of both types of steels to nuclear applications. It provides a synthetic and comprehensive review of the plasticity mechanisms in austenitic steels during tensile tests below 400 °C. In particular, formation of twins, extended stacking faults, and martensite, as well as irradiation effects and grain rotation are discussed in details.

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

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

    1997-04-01

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

  8. Commercial food irradiation

    SciTech Connect

    Black, E.F.; Libby, L.M.

    1983-06-01

    Food irradiation is discussed. Irradiation exposes food to gamma rays from a cobalt-60 or a cesium-137 source, or to high-energy electrons emitted by an electron accelerator. A major advantage is that food can be packaged either before or after treatment. FDA regulations with regard to irradiation are discussed. Comments on an 'Advance Notice' on irradiation, published by the FDA in 1981 are summarized.

  9. X-ray attenuation properties of stainless steel (u)

    SciTech Connect

    Wang, Lily L; Berry, Phillip C

    2009-01-01

    Stainless steel vessels are used to enclose solid materials for studying x-ray radiolysis that involves gas release from the materials. Commercially available stainless steel components are easily adapted to form a static or a dynamic condition to monitor the gas evolved from the solid materials during and after the x-ray irradiation. Experimental data published on the x-ray attenuation properties of stainless steel, however, are very scarce, especially over a wide range of x-ray energies. The objective of this work was to obtain experimental data that will be used to determine how a poly-energetic x-ray beam is attenuated by the stainless steel container wall. The data will also be used in conjunction with MCNP (Monte Carlos Nuclear Particle) modeling to develop an accurate method for determining energy absorbed in known solid samples contained in stainless steel vessels. In this study, experiments to measure the attenuation properties of stainless steel were performed for a range of bremsstrahlung x-ray beams with a maximum energy ranging from 150 keV to 10 MeV. Bremsstrahlung x-ray beams of these energies are commonly used in radiography of engineering and weapon components. The weapon surveillance community has a great interest in understanding how the x-rays in radiography affect short-term and long-term properties of weapon materials.

  10. Multiscale Modeling of the Deformation of Advanced Ferritic Steels for Generation IV Nuclear Energy

    SciTech Connect

    Nasr M. Ghoniem; Nick Kioussis

    2009-04-18

    The objective of this project is to use the multi-scale modeling of materials (MMM) approach to develop an improved understanding of the effects of neutron irradiation on the mechanical properties of high-temperature structural materials that are being developed or proposed for Gen IV applications. In particular, the research focuses on advanced ferritic/ martensitic steels to enable operation up to 650-700°C, compared to the current 550°C limit on high-temperature steels.

  11. Irradiation creep of candidate materials for advanced nuclear plants

    NASA Astrophysics Data System (ADS)

    Chen, J.; Jung, P.; Hoffelner, W.

    2013-10-01

    In the present paper, irradiation creep results of an intermetallic TiAl alloy and two ferritic oxide dispersion strengthened (ODS) steels are summarized. In situ irradiation creep measurements were performed using homogeneous implantation with α- and p-particles to maximum doses of 0.8 dpa at displacement damage rates of 2-8 × 10-6 dpa/s. The strains of miniaturized flat dog-bone specimens were monitored under uniaxial tensile stresses ranging from 20 to 400 MPa at temperatures of 573, 673 and 773 K, respectively. The effects of material composition, ODS particle size, and bombarding particle on the irradiation creep compliance was studied and results are compared to literature data. Evolution of microstructure during helium implantation was investigated in detail by TEM and is discussed with respect to irradiation creep models.

  12. Pulsed Magnetic Welding for Advanced Core and Cladding Steel

    SciTech Connect

    Cao, Guoping; Yang, Yong

    2013-12-19

    To investigate a solid-state joining method, pulsed magnetic welding (PMW), for welding the advanced core and cladding steels to be used in Generation IV systems, with a specific application for fuel pin end-plug welding. As another alternative solid state welding technique, pulsed magnetic welding (PMW) has not been extensively explored on the advanced steels. The resultant weld can be free from microstructure defects (pores, non-metallic inclusions, segregation of alloying elements). More specifically, the following objectives are to be achieved: 1. To design a suitable welding apparatus fixture, and optimize welding parameters for repeatable and acceptable joining of the fuel pin end-plug. The welding will be evaluated using tensile tests for lap joint weldments and helium leak tests for the fuel pin end-plug; 2 Investigate the microstructural and mechanical properties changes in PMW weldments of proposed advanced core and cladding alloys; 3. Simulate the irradiation effects on the PWM weldments using ion irradiation.

  13. Glass Stronger than Steel

    DOE R&D Accomplishments Database

    Yarris, Lynn

    2011-03-28

    A new type of damage-tolerant metallic glass, demonstrating a strength and toughness beyond that of steel or any other known material, has been developed and tested by a collaboration of researchers from Berkeley Lab and Caltech.

  14. Structural Amorphous Steels

    NASA Astrophysics Data System (ADS)

    Lu, Z. P.; Liu, C. T.; Thompson, J. R.; Porter, W. D.

    2004-06-01

    Recent advancement in bulk metallic glasses, whose properties are usually superior to their crystalline counterparts, has stimulated great interest in fabricating bulk amorphous steels. While a great deal of effort has been devoted to this field, the fabrication of structural amorphous steels with large cross sections has remained an alchemist’s dream because of the limited glass-forming ability (GFA) of these materials. Here we report the discovery of structural amorphous steels that can be cast into glasses with large cross-section sizes using conventional drop-casting methods. These new steels showed interesting physical, magnetic, and mechanical properties, along with high thermal stability. The underlying mechanisms for the superior GFA of these materials are discussed.

  15. Radiation damage studies on stainless steel, Ni, Cu, Mo for nuclear fusion reactors

    NASA Astrophysics Data System (ADS)

    Constantinescu, B.; Sarbu, C.; Simionescu, Luiza

    1997-04-01

    Studies on dose and energy dependence of blistering and flaking on stainless steels, Ni, Cu, Mo produced by 3.0, 4.7 and 6.8 MeV He + ions irradiation are presented. Using SEM and TEM techniques, irradiation phenomena such as sponge- and wave-like structures, submicronic cracks, microcraters, helium bubbles on matrix, grain boundaries, loops and TiC precipitates are illustrated. The appearance of an amorphous phase in the Ti-modified austenitic steel 12KH18N10T is discussed. QTMD preliminary results on H reemission in He preimplanted Ni samples are reported.

  16. Radiation Stability of Nanoclusters in Nano-structured Oxide Dispersion Strengthened (ODS) Steels

    SciTech Connect

    Certain, Alicia G.; Kuchibhatla, Satyanarayana V N T; Shutthanandan, V.; Hoelzer, D. T.; Allen, T. R.

    2013-03-01

    Nanostructured oxide dispersion strengthened (ODS) steels are considered candidates for nuclear fission and fusion applications at high temperature and dose. The complex oxide nanoclusters in these alloys provide high-temperature strength and are expected to afford better radiation resistance. Proton, heavy ion, and neutron irradiations have been performed to evaluate cluster stability in 14YWT and 9CrODS steel under a range of irradiation conditions. Energy-filtered transmission electron microscopy and atom probe tomography were used in this work to analyze the evolution of the oxide population.

  17. Helium-induced weld cracking in irradiated 304 stainless steel

    SciTech Connect

    Birchenall, A.K. )

    1989-01-01

    This report consists of slide notes for presentation to The Metallurgical Society of the American Institute of Mining, Metallurgical and Petroleum Engineers (AIME). The meeting in question will be held October 3, 1989 in Indianapolis. This presentation will be the second of three consecutive talks contributed by SRL personnel dealing with helium-induced weld cracking.

  18. Life after Steel

    ERIC Educational Resources Information Center

    Mangan, Katherine

    2013-01-01

    Bobby Curran grew up in a working-class neighborhood in Baltimore, finished high school, and followed his grandfather's steel-toed bootprints straight to Sparrows Point, a 3,000-acre sprawl of industry on the Chesapeake Bay. College was not part of the plan. A gritty but well-paying job at the RG Steel plant was Mr. Curran's ticket to a secure…

  19. Summary of the U.S. specimen matrix for the HFIR 13J varying temperature irradiation capsule

    SciTech Connect

    Zinkle, S.J.

    1998-03-01

    The US specimen matrix for the collaborative DOE/Monbusho HFIR 13J varying temperature irradiation capsule contains two ceramics and 29 different metals, including vanadium alloys, ferritic/martensitic steels, pure iron, austenitic stainless steels, nickel alloys, and copper alloys. This experiment is designed to provide fundamental information on the effects of brief low-temperature excursions on the tensile properties and microstructural evolution of a wide range of materials irradiated at nominal temperatures of 350 and 500 C to a dose of {approximately}5 dpa. A total of 340 miniature sheet tensile specimens and 274 TEM disks are included in the US-supplied matrix for the irradiation capsule.

  20. Irradiation creep behavior of V-4Cr-4Ti alloys irradiated in a liquid sodium environment at the JOYO fast reactor

    NASA Astrophysics Data System (ADS)

    Fukumoto, Ken-ichi; Matsui, Hideki; Narui, Minoru; Yamazaki, Masanori

    2013-06-01

    Irradiation experiments on V-4Cr-4Ti alloys with sodium-enclosed irradiation capsules in the JOYO fast reactor were conducted using pressurized creep tubes (PCTs). The irradiation creep strain was significantly larger than the thermal creep strain below 686 °C, but there was no swelling of the neutron-irradiated V-4Cr-4Ti alloys. At temperatures below 500 °C, the irradiation creep was found to be proportional to the square root of the neutron dose and linear with the stress level. Above 500 °C, it was expected to be proportional to the stress level to a power greater than unity, because the irradiation creep mechanism could change from the stress-induced preferred absorption mechanism (SIPA) to the preferred absorption glide mechanism (PGA). By comparing annealed PCT specimens with cold-worked specimens, the cold-worked V-4Cr-4Ti alloys exhibited a larger irradiation creep strain compared with the annealed alloys. The irradiation creep compliance of the V-4Cr-4Ti alloys were ˜10 × 10-6 MPa-1 dpa-1 below 500 °C and 50-200 × 10-6 MPa-1 dpa-1 above 500 °C, a value greater than that of commercial V-4Cr-4Ti alloys, austenitic steels and ferritic steels.

  1. Waste product profile: Steel cans

    SciTech Connect

    Miller, C.

    1996-07-01

    Steel cans are made from tinplate steel, which is produced in basic oxygen furnaces. A thin layer of tin is applied to the can`s inner and outer surfaces to prevent rusting and protect food and beverage flavors. As a result, steel cans are often called tin cans. Steel mills are the largest market for steel cans. Integrated mills use the basic oxygen process to manufacture tinplate, appliances, car bodies, and steel framing. Electric arc furnaces use 100% scrap to produce steel shapes such as railroad ties and bridge spans. Electric arc furnaces are more geographically diverse and tend to have smaller capacities than basic oxygen furnaces. Detinners remove the tin from steel cans for resale to tin using industries. Continued decreases in the amount of tin used in steel cans has lessened the importance of this market. Foundries use scrap as a raw material in making castings and molds for industrial users.

  2. Articles comprising ferritic stainless steels

    DOEpatents

    Rakowski, James M.

    2016-06-28

    An article of manufacture comprises a ferritic stainless steel that includes a near-surface region depleted of silicon relative to a remainder of the ferritic stainless steel. The article has a reduced tendency to form an electrically resistive silica layer including silicon derived from the steel when the article is subjected to high temperature oxidizing conditions. The ferritic stainless steel is selected from the group comprising AISI Type 430 stainless steel, AISI Type 439 stainless steel, AISI Type 441 stainless steel, AISI Type 444 stainless steel, and E-BRITE.RTM. alloy, also known as UNS 44627 stainless steel. In certain embodiments, the article of manufacture is a fuel cell interconnect for a solid oxide fuel cell.

  3. Preliminary report on the irradiation conditions of the HFIR JP-23 experiment

    SciTech Connect

    Ermi, A.M.; Gelles, D.S.

    1995-04-01

    The objective of this effort was to irradiate a series of alloys over the temperature range 300 to 600{degrees}C to approximately 10 dpa in the High Flux Isotope Reactor (HFIR). The alloys covered a wide range of materials and treatments. The Japanese specimen matrix consisted of ferritic steels, vanadium alloys, copper alloys, molybdenum alloys, and titanium-aluminum compounds. The US specimen matrix consisted of vanadium alloys, 316 stainless steels, and isotopically tailored ferritic and austenitic alloys.

  4. Recent Progress of R&D Activities on Reduced Activation Ferritic/Martensitic Steels

    SciTech Connect

    Huang, Q.; Baluc, N.; Dai, Y.; Jitsukawa, S.; Kimura, A.; Konys, J.; Kurtz, Richard J.; Lindau, R.; Muroga, T.; Odette, George R.; Raj, B.; Stoller, Roger E.; Tan, L.; Tanigawa, Hiroyasu; Tavassoli, A,-A.F.; Yamamoto, Takuya; Wan, F.; Wu, Y.

    2013-01-03

    Several types of reduced activation ferritic/martensitic (RAFM) steel have been developed over the past 30 years in China, Europe, India, Japan, Russia and the USA for application in ITER TBM and future fusion DEMO and power reactors. The progress has been particularly important during the past few years with evaluation of mechanical porperties of these steels before and after irradiation and in contact with different cooling media. This paper presents recent RAFM steel results obtained in ITER partner countries in relation with different TBM and DEMO options

  5. Effect of Ni content on thermal and radiation resistance of VVER RPV steel

    NASA Astrophysics Data System (ADS)

    Shtrombakh, Ya. I.; Gurovich, B. A.; Kuleshova, E. A.; Frolov, A. S.; Fedotova, S. V.; Zhurko, D. A.; Krikun, E. V.

    2015-06-01

    In this paper thermal stability and radiation resistance of VVER-type RPV steels for pressure vessels of advanced reactors with different nickel content were studied. A complex of microstructural studies and mechanical tests of the steels in different states (after long thermal exposures, provoking embrittling heat treatment and accelerated neutron irradiation) was carried out. It is shown that nickel content (other things being equal) determines the extent of materials degradation under influence of operational factors: steels with a lower nickel concentration demonstrate a higher thermal stability and radiation resistance.

  6. Current status and future R&D for reduced-activation ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Hishinuma, A.; Kohyama, A.; Klueh, R. L.; Gelles, D. S.; Dietz, W.; Ehrlich, K.

    1998-10-01

    International research and development programs on reduced-activation ferritic/martensitic steels, the primary candidate-alloys for a DEMO fusion reactor and beyond, are briefly summarized, along with some information on conventional steels. An International Energy Agency (IEA) collaborative test program to determine the feasibility of reduced-activation ferritic/martensitic steels for fusion is in progress and will be completed within this century. Baseline properties including typical irradiation behavior for Fe-(7-9)%Cr reduced-activation ferritic steels are shown. Most of the data are for a heat of modified F82H steel, purchased for the IEA program. Experimental plans to explore possible problems and solutions for fusion devices using ferromagnetic materials are introduced. The preliminary results show that it should be possible to use a ferromagnetic vacuum vessel in tokamak devices.

  7. Profiles in garbage: Steel cans

    SciTech Connect

    Miller, C.

    1998-02-01

    Steel mills are the largest market for steel cans. Integrated mills use the basic oxygen process to manufacture tinplate, appliances, car bodies, and steel framing. Electric arc furnaces use 100% scrap to produce steel shapes such as railroad ties and bridge spans. Electric arc furnaces are more geographically diverse and tend to have smaller capacities than basic oxygen furnaces. Detinners remove the tin from steel cans for resale to tin using industries. With less tin use in steel cans, the importance of the detinning market has declined substantially. Foundries use scrap as a raw material in making castings and molds for industrial users.

  8. Neutron irradiation and compatibility testing of Li 2O

    NASA Astrophysics Data System (ADS)

    Porter, D. L.; Krsul, J. R.; Laug, M. T.; Walters, L. C.; Tetenbaum, M.

    1984-05-01

    A study was made of the neutron irradiation behavior of 6Li-enriched Li 2O in EBR-II. In addition, a stress corrosion study was performed ex-reactor to test the compatibility of Li 2O with a variety of stainless steels. The irradiation tests showed that tritium and helium retention in the Li 2O (˜ 89% dense) lessened with neutron exposure, and the retentions appear to approach a steady-state after ˜ 1% 6Li burnup. The stress corrosion studies, using 316 stainless steel (Ti-modified) and a 35% Ni alloy, showed that stress does not enhance the corrosion, and that dry Li 2O is not significantly corrosive, the LiOH content producing the corrosive effects. Corrosion, in general, was not severe because a passivation in sealed capsules seemed to occur after a time which greatly reduced corrosion rates.

  9. Irradiation creep and swelling of various austenitic alloys irradiated in PFR and FFTF

    SciTech Connect

    Garner, F.A.; Toloczko, M.B.

    1996-10-01

    In order to use data from surrogate neutron spectra for fusion applications, it is necessary to analyze the impact of environmental differences on property development. This is of particular importance in the study of irradiation creep and its interactions with void swelling, especially with respect to the difficulty of separation of creep strains from various non-creep strains. As part of an on-going creep data rescue and analysis effort, the current study focuses on comparative irradiations conducted on identical gas-pressurized tubes produced and constructed in the United States from austenitic steels (20% CW 316 and 20% CW D9), but irradiated in either the Prototype Fast Reactor (PFR) in the United Kingdom or the Fast Flux Test Facility in the United States. In PFR, Demountable Subassemblies (DMSA) serving as heat pipes were used without active temperature control. In FFTF the specimens were irradiated with active ({+-}{degrees}5C) temperature control. Whereas the FFTF irradiations involved a series of successive side-by-side irradiation, measurement and reinsertion of the same series of tubes, the PFR experiment utilized simultaneous irradiation at two axial positions in the heat pipe to achieve different fluences at different flux levels. The smaller size of the DMSA also necessitated a separation of the tubes at a given flux level into two groups (low-stress and high-stress) at slightly different axial positions, where the flux between the two groups varied {le}10%. Of particular interest in this study was the potential impact of the two types of separation on the derivation of creep coefficients.

  10. View northwest, wharf A, sheet steel bulkhead, steel lift tower ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    View northwest, wharf A, sheet steel bulkhead, steel lift tower - U.S. Coast Guard Sandy Hook Station, Western Docking Structure, West of intersection of Canfield Road & Hartshorne Drive, Highlands, Monmouth County, NJ

  11. Mechanical properties of irradiated multi-phase polycrystalline BCC materials

    NASA Astrophysics Data System (ADS)

    Song, Dingkun; Xiao, Xiazi; Xue, Jianming; Chu, Haijian; Duan, Huiling

    2015-04-01

    Structure materials under severe irradiations in nuclear environments are known to degrade because of irradiation hardening and loss of ductility, resulting from irradiation-induced defects such as vacancies, interstitials and dislocation loops, etc. In this paper, we develop an elastic-viscoplastic model for irradiated multi-phase polycrystalline BCC materials in which the mechanical behaviors of individual grains and polycrystalline aggregates are both explored. At the microscopic grain scale, we use the internal variable model and propose a new tensorial damage descriptor to represent the geometry character of the defect loop, which facilitates the analysis of the defect loop evolutions and dislocation-defect interactions. At the macroscopic polycrystal scale, the self-consistent scheme is extended to consider the multiphase problem and used to bridge the individual grain behavior to polycrystal properties. Based on the proposed model, we found that the work-hardening coefficient decreases with the increase of irradiation-induced defect loops, and the orientation/loading dependence of mechanical properties is mainly attributed to the different Schmid factors. At the polycrystalline scale, numerical results for pure Fe match well with the irradiation experiment data. The model is further extended to predict the hardening effect of dispersoids in oxide-dispersed strengthened steels by the considering the Orowan bowing. The influences of grain size and irradiation are found to compete to dominate the strengthening behaviors of materials.

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

    NASA Astrophysics Data System (ADS)

    Frazer, D.; Qvist, S.; Parker, S.; Krumwiede, D. L.; Caro, M.; Tesmer, J.; Maloy, S. A.; Wang, Y. Q.; Hosemann, P.

    2016-10-01

    A potentially promising coolant/structural material pair for a liquid-metal-cooled fast reactors is lead bismuth eutectic (LBE) coolant with the ferritic/martensitic steel HT9. The challenge of deploying LBE, however, is the corrosive environment it creates for structural materials. This corrosion can be mitigated with precise oxygen content control in the LBE to allow for the growth of passive protective oxide layers on the surface of the steel. In this paper, results are reported from the Irradiation Corrosion Experiment II (ICE-II), which allowed the simultaneous irradiation of a sample while in contact with LBE. It was found that a characteristic multilayer structure with an outer Fe3O4 oxide and inner FeCr2O4 spinel was grown and the oxidation was significantly larger in the irradiated region when compared to the region that was only exposed to LBE corrosion. Possible mechanisms are discussed to help understand this irradiation enhanced corrosion behavior.

  13. Packaging and shipment of irradiated spent fuel. Final report

    SciTech Connect

    Kohli, R.; Lawrence, A.

    1988-10-01

    Irradiated spent fuel rods, rod sections, and other loose fuel were retrieved from various storage locations at the Battelle hot cells, packaged in stainless steel tubes, and inserted in a new basket assembly in preparation for shipment to EG&G Idaho. Few assemblies Connecticut Yankee S004 and Turkey Point 817 were also retrieved and prepared for shipment. All three fuel assemblies were loaded in shipping cask TN8-L and shipped to EG&G Idaho for storage.

  14. Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Anderoglu, Osman; Byun, Thak Sang; Toloczko, Mychailo; Maloy, Stuart A.

    2013-01-01

    Ferritic/martensitic (F/M) steels are considered for core applications and pressure vessels in Generation IV reactors as well as first walls and blankets for fusion reactors. There are significant scientific data on testing and industrial experience in making this class of alloys worldwide. This experience makes F/M steels an attractive candidate. In this article, tensile behavior, fracture toughness and impact property, and creep behavior of the F/M steels under neutron irradiations to high doses with a focus on high Cr content (8 to 12) are reviewed. Tensile properties are very sensitive to irradiation temperature. Increase in yield and tensile strength (hardening) is accompanied with a loss of ductility and starts at very low doses under irradiation. The degradation of mechanical properties is most pronounced at <0.3 T M ( T M is melting temperature) and up to 10 dpa (displacement per atom). Ferritic/martensitic steels exhibit a high fracture toughness after irradiation at all temperatures even below 673 K (400 °C), except when tested at room temperature after irradiations below 673 K (400 °C), which shows a significant reduction in fracture toughness. Creep studies showed that for the range of expected stresses in a reactor environment, the stress exponent is expected to be approximately one and the steady state creep rate in the absence of swelling is usually better than austenitic stainless steels both in terms of the creep rate and the temperature sensitivity of creep. In short, F/M steels show excellent promise for high dose applications in nuclear reactors.

  15. Superclean steel development

    SciTech Connect

    Richman, R.H.; McNaughton, W.P. )

    1989-12-01

    The Electric Power Research Institute has actively encouraged and sponsored a number of research projects to develop a superclean 3.5NiCrMoV steel for low pressure turbine rotors. Such steel is highly resistant to temper embrittlement and will thus facilitate increased efficiency in electricity generation through the use of higher operating temperatures and improvements in design. The objective of this interim report was to integrate the results that have been generated to date worldwide in the pursuit of superclean steel. The report contains detailed findings that enable the interested utility to evaluate how the results affect utility decision making. A companion document has been written to summarize the findings from this technical report. The results indicate that steels with impurity contents typical of the superclean specification can be manufactured for production rotors with properties that equal or exceed those for conventional 3.5NiCrMoV rotors in every detail. Of particular interest are the results that the superclean steels appear to be virtually resistant to temper embrittlement to a temperature of 500 {degrees}C. 109 refs., 51 figs., 9 tabs.

  16. Brazing titanium to stainless steel

    NASA Technical Reports Server (NTRS)

    Batista, R. I.

    1980-01-01

    Titanium and stainless-steel members are usually joined mechanically for lack of any other effective method. New approach using different brazing alloy and plating steel member with nickel resolves problem. Process must be carried out in inert atmosphere.

  17. Continuous steel production and apparatus

    DOEpatents

    Peaslee, Kent D.; Peter, Jorg J.; Robertson, David G. C.; Thomas, Brian G.; Zhang, Lifeng

    2009-11-17

    A process for continuous refining of steel via multiple distinct reaction vessels for melting, oxidation, reduction, and refining for delivery of steel continuously to, for example, a tundish of a continuous caster system, and associated apparatus.

  18. Neutron dosimetry and damage calculations for the EBRII COBRA-1A irradiations

    SciTech Connect

    Greenwood, L.R.; Ratner, R.T.

    1997-04-01

    Neutron fluence measurements and radiation damage calculations are reported for the joint U.S. and Japanese COBRA-1A1 and 1A2 irradiations in the Experimental Breeder Reactor II. The maximum total neutron fluences at midplane were 2.0E+22 and 7.5E+22 n/cm{sup 2}, for the 1A1 and 1A2 irradiations, respectively, resulting in about 8.0 and 30.3 dpa in stainless steel.

  19. Radiation-induced segregation and precipitation behaviours around cascade clusters under electron irradiation.

    PubMed

    Sueishi, Yuichiro; Sakaguchi, Norihito; Shibayama, Tamaki; Kinoshita, Hiroshi; Takahashi, Heishichiro

    2003-01-01

    We have investigated the formation of cascade clusters and structural changes in them by means of electron irradiation following ion irradiation in an austenitic stainless steel. Almost all of the cascade clusters, which were introduced by the ion irradiation, grew to form interstitial-type dislocation loops or vacancy-type stacking fault tetrahedra after electron irradiation at 623 K, whereas a few of the dot-type clusters remained in the matrix. It was possible to recognize the concentration of Ni and Si by radiation-induced segregation around the dot-type clusters. After electron irradiation at 773 K, we found that some cascade clusters became precipitates (delta-Ni2Si) due to radiation-induced precipitation. This suggests that the cascade clusters could directly become precipitation sites during irradiation.

  20. The weldability of low activation Cr-W steels

    NASA Astrophysics Data System (ADS)

    Wang, C. A.; Klueh, R. L.; Chin, B. A.

    1992-09-01

    A series of chromium-tungsten ferritic steels patterned on the chromium-molybdenum alloys, case 2 14 Cr-1Mo, 9Cr-1MoVNb and 12Cr-1MoVW, were tested for weldability. These steels are being developed as candidates for the first wall and blanket structure of fusion reactors. Use of these materials will minimize the long-term radioactive hazards associated with disposal after service. In these low activation alloys, long half-life elements (Mo and Nb), which become activated during irradiation, are replaced. Autogenous bead-on-plate welds were performed using the gas tungsten arc welding process. Experimental results showed that all welds were free of cracks. Sound welds were achieved in case 2 14 and 5% Cr-W low activation steels while loss of ductility was observed in 9 and 12% Cr-W steels. This results suggests that post-weld heat treatment is necessary to restore toughness to the 9-12% Cr-W steels.

  1. A-3 steel work completed

    NASA Technical Reports Server (NTRS)

    2009-01-01

    Stennis Space Center engineers celebrated a key milestone in construction of the A-3 Test Stand on April 9 - completion of structural steel work. Workers with Lafayette (La.) Steel Erector Inc. placed the last structural steel beam atop the stand during a noon ceremony attended by more than 100 workers and guests.

  2. Microbial-Influenced Corrosion of Corten Steel Compared with Carbon Steel and Stainless Steel in Oily Wastewater by Pseudomonas aeruginosa

    NASA Astrophysics Data System (ADS)

    Mansouri, Hamidreza; Alavi, Seyed Abolhasan; Fotovat, Meysam

    2015-07-01

    The microbial corrosion behavior of three important steels (carbon steel, stainless steel, and Corten steel) was investigated in semi petroleum medium. This work was done in modified nutrient broth (2 g nutrient broth in 1 L oily wastewater) in the presence of Pseudomonas aeruginosa and mixed culture (as a biotic media) and an abiotic medium for 2 weeks. The behavior of corrosion was analyzed by spectrophotometric and electrochemical methods and at the end was confirmed by scanning electron microscopy. The results show that the degree of corrosion of Corten steel in mixed culture, unlike carbon steel and stainless steel, is less than P. aeruginosa inoculated medium because some bacteria affect Corten steel less than other steels. According to the experiments, carbon steel had less resistance than Corten steel and stainless steel. Furthermore, biofilm inhibits separated particles of those steels to spread to the medium; in other words, particles get trapped between biofilm and steel.

  3. MASSIVE LEAKAGE IRRADIATOR

    DOEpatents

    Wigner, E.P.; Szilard, L.; Christy, R.F.; Friedman, F.L.

    1961-05-30

    An irradiator designed to utilize the neutrons that leak out of a reactor around its periphery is described. It avoids wasting neutron energy and reduces interference with the core flux to a minimum. This is done by surrounding all or most of the core with removable segments of the material to be irradiated within a matrix of reflecting material.

  4. Perspective on food irradiation

    SciTech Connect

    Not Available

    1987-02-01

    Recent US Food and Drug Administration approval of irradiation treatment for fruit, vegetables and pork has stimulated considerable discussion in the popular press on the safety and efficacy of irradiation processing of food. This perspective is designed to summarize the current scientific information available on this issue.

  5. Irradiation Creep in Graphite

    SciTech Connect

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  6. Braze alloy spreading on steel

    NASA Technical Reports Server (NTRS)

    Siewert, T. A.; Heine, R. W.; Lagally, M. G.

    1978-01-01

    Scanning electron microscopy (SEM) and Auger electron microscopy (AEM) were employed to observe elemental surface decomposition resulting from the brazing of a copper-treated steel. Two types of steel were used for the study, stainless steel (treated with a eutectic silver-copper alloy), and low-carbon steel (treated with pure copper). Attention is given to oxygen partial pressure during the processes; a low enough pressure (8 x 10 to the -5th torr) was found to totally inhibit the spreading of the filler material at a fixed heating cycle. With both types of steel, copper treatment enhanced even spreading at a decreased temperature.

  7. Deformation in metals after low temperature irradiation: Part II - Irradiation hardening, strain hardening, and stress ratios

    SciTech Connect

    Byun, Thak Sang; Li, Meimei

    2008-03-01

    Effects of irradiation at temperatures 200oC on tensile stress parameters are analyzed for dozens of bcc, fcc, and hcp pure metals and alloys, focusing on irradiation hardening, strain hardening, and relationships between the true stress parameters. Similar irradiation-hardening rates are observed for all the metals irrespective of crystal type; typically, the irradiation-hardening rates are large, in the range 100 - 1000 GPa/dpa, at the lowest dose of <0.0001 dpa and decrease with dose to a few tens of MPa/dpa or less at about 10 dpa. However, average irradiation-hardening rates over the dose range of 0 dpa − (the dose to plastic instability at yield) are considerably lower for stainless steels due to their high uniform ductility. It is shown that whereas low temperature irradiation increases the yield stress, it does not significantly change the strain-hardening rate of metallic materials; it decreases the fracture stress only when non-ductile failure occurs. Such dose independence in strain hardening behavior results in strong linear relationships between the true stress parameters. Average ratios of plastic instability stress to unirradiated yield stress are about 1.4, 3.9, and 1.3 for bcc metals (and precipitation hardened IN718 alloy), annealed fcc metals (and pure Zr), and Zr-4 alloy, respectively. Ratios of fracture stress to plastic instability stress are calculated to be 2.2, 1.7, and 2.1, respectively. Comparison of these values confirms that the annealed fcc metals and other soft metals have larger uniform ductility but smaller necking ductility when compared to other materials.

  8. Sensitization of stainless steel

    NASA Technical Reports Server (NTRS)

    Nagy, James P.

    1990-01-01

    The objective of this experiment is to determine the corrosion rates of 18-8 stainless steels that have been sensitized at various temperatures and to show the application of phase diagrams. The laboratory instructor will assign each student a temperature, ranging from 550 C to 1050 C, to which the sample will be heated. Further details of the experimental procedure are detailed.

  9. Alloy development for irradiation performance in fusion reactors

    NASA Astrophysics Data System (ADS)

    Harling, O. K.; Grant, N. J.

    1980-12-01

    The development of improved structural alloys for the fusion reactor first wall application is addressed. Several new alloys were produced by rapid solidification. Emphasis in alloy design and production was placed on producing austenitic Type 316SS with fine dispersions of TiC and Al2O3 particles. Results of mechanical and microstructural tests are presented. A number of neutron irradiations were initiated on samples fabricated from alloys produced. A dual beam, heavy ion, and helium ion, irradiation was completed using several alloys and a range of temperatures, damage rates, and total doses. Modeling of irradiation phenomena was continued with emphasis on understanding the effect of recoil resolution on relatively stable second phase particles. The microstructure of several ZrB2 doped stainless steels was characterized.

  10. High Temperature Irradiation Effects in Selected Generation IV Structural Alloys

    SciTech Connect

    Nanstad, Randy K; McClintock, David A; Hoelzer, David T; Tan, Lizhen; Allen, Todd R.

    2009-01-01

    In the Generation IV Materials Program cross-cutting task, irradiation and testing were carried out to address the issue of high temperature irradiation effects with selected current and potential candidate metallic alloys. The materials tested were (1) a high-nickel iron-base alloy (Alloy 800H); (2) a nickel-base alloy (Alloy 617); (3) two advanced nano-structured ferritic alloys (designated 14YWT and 14WT); and (4) a commercial ferritic-martensitic steel (annealed 9Cr-1MoV). Small tensile specimens were irradiated in rabbit capsules in the High-Flux Isotope Reactor at temperatures from about 550 to 700 C and to irradiation doses in the range 1.2 to 1.6 dpa. The Alloy 800H and Alloy 617 exhibited significant hardening after irradiation at 580 C; some hardening occurred at 660 C as well, but the 800H showed extremely low tensile elongations when tested at 700 C. Notably, the grain boundary engineered 800H exhibited even greater hardening at 580 C and retained a high amount of ductility. Irradiation effects on the two nano-structured ferritic alloys and the annealed 9Cr-1MoV were relatively slight at this low dose.

  11. Special steel production on common carbon steel production line

    NASA Astrophysics Data System (ADS)

    Pi, Huachun; Han, Jingtao; Hu, Haiping; Bian, Ruisheng; Kang, Jianjun; Xu, Manlin

    2004-06-01

    The equipment and technology of small bar tandem rolling line of Shijiazhuang Iron & Steel Co. in China has reached the 90's international advanced level in the 20th century, but products on the line are mostly of common carbon steel. Currently there are few steel plants in China to produce 45 steel bars for cold drawing, which is a kind of shortage product. Development of 45 steel for cold drawing has a wide market outlook in China. In this paper, continuous cooling transformation (CCT) curve of 45 steel for cold drawing used for rolling was set out first. According to the CCT curve, we determined some key temperature points such as Ac3 temperature and Ac1 temperature during the cooling procedure and discussed the precipitation microstructure at different cooling rate. Then by studying thermal treatment process of 45 steel bars for cold drawing, the influence of cooling time on microstructure was analyzed and the optimum cooling speed has been found. All results concluded from the above studies are the basis of regulating controlled cooling process of 45 steel bars for cold drawing. Finally, the feasible production process of 45 steel bars for cold drawing on common carbon steel production line combined with the field condition was recommended.

  12. Collaborative investigations of in-service irradiated material from the Japan Power Demonstration Reactor pressure vessel

    SciTech Connect

    Corwin, W.R.; Broadhead, B.L.; Suzuki, M.; Kohsaka, A.

    1997-02-01

    There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

  13. Test reactor irradiation coordination

    SciTech Connect

    Heartherly, D.W.; Siman Tov, I.I.; Sparks, D.W.

    1995-10-01

    This task was established to supply and coordinate irradiation services needed by NRC contractors other than ORNL. These services include the design and assembly of irradiation capsules as well as arranging for their exposure, disassembly, and return of specimens. During this period, the final design of the facility and specimen baskets was determined through an iterative process involving the designers and thermal analysts. The resulting design should permit the irradiation of all test specimens to within 5{degrees}C of their desired temperature. Detailing of all parts is ongoing and should be completed during the next reporting period. Procurement of the facility will also be initiated during the next review period.

  14. Alaskan Commodities Irradiation Project

    SciTech Connect

    Zarling, J.P.; Swanson, R.B.; Logan, R.R.; Das, D.K.; Lewis, C.E.; Workman, W.G.; Tumeo, M.A.; Hok, C.I.; Birklid, C.A.; Bennett, F.L.

    1988-12-01

    The ninety-ninth US Congress commissioned a six-state food irradiation research and development program to evaluate the commercial potential of this technology. Hawaii, Washington, Iowa, Oklahoma and Florida as well as Alaska have participated in the national program; various food products including fishery products, red meats, tropical and citrus fruits and vegetables have been studied. The purpose of the Alaskan study was to review and evaluate those factors related to the technical and economic feasibility of an irradiator in Alaska. This options analysis study will serve as a basis for determining the state's further involvement in the development of food irradiation technology. 40 refs., 50 figs., 53 tabs.

  15. Ferritic steel melt and FLiBe/steel experiment : melting ferritic steel.

    SciTech Connect

    Troncosa, Kenneth P.; Smith, Brandon M.; Tanaka, Tina Joan

    2004-11-01

    In preparation for developing a Z-pinch IFE power plant, the interaction of ferritic steel with the coolant, FLiBe, must be explored. Sandia National Laboratories Fusion Technology Department was asked to drop molten ferritic steel and FLiBe in a vacuum system and determine the gas byproducts and ability to recycle the steel. We tried various methods of resistive heating of ferritic steel using available power supplies and easily obtained heaters. Although we could melt the steel, we could not cause a drop to fall. This report describes the various experiments that were performed and includes some suggestions and materials needed to be successful. Although the steel was easily melted, it was not possible to drip the molten steel into a FLiBe pool Levitation melting of the drop is likely to be more successful.

  16. Development of corrosion-resistant improved Al-doped austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Kondo, Keietsu; Miwa, Yukio; Okubo, Nariaki; Kaji, Yoshiyuki; Tsukada, Takashi

    2011-10-01

    Aluminum-doped type 316L SS (316L/Al) has been developed for the purpose of suppressing the degradation of corrosion resistance induced by irradiation in austenitic stainless steels (SSs). The electrochemical corrosion properties of this material were estimated after Ni-ion irradiation at a temperature range from 330 °C to 550 °C. When irradiated at 550 °C up to 12 dpa, 316L/Al showed high corrosion resistance in the vicinity of grain boundaries (GBs) and in grains, while severe GB etching and local corrosion in grains were observed in irradiated 316L and 316 SS. It is supposed that aluminum enrichment was enhanced by high-temperature irradiation at GBs and in grains, to compensate for lost corrosion resistance induced by chromium depletion.

  17. Insight on the inconsistencies of Barkhausen signal measurements for radiation damage on nuclear reactor steel

    SciTech Connect

    Barroso, Soraia Pirfo; Fitzpatrick, Michael E.; Gillemot, Ferenc; Horváth, Marta; Horváth, Ákos; Szekely, Richard

    2014-02-18

    This paper focuses on the use of magnetic measurements, using Barkhausen signals to determine the irradiation effects, attempting to predict fracture toughness changes on nuclear reactor structural materials and correlating these measurements to mechanical testing and microstructure. For this study, two types of nuclear reactor materials were investigated: one sensitive to irradiation effects, the JRQ IAEA's reference material (A533B- -type); and one resistant material, 15KH2MFA WWER's reactor pressure vessel steel. The samples were carefully identified within the original heat block, i.e. forged or rolled plate. These calibrated samples were irradiated at different neutron fluences up to 10{sup 23} n/m{sup 2}. We show how microstructural anisotropy can mask the irradiation effects in the magnetic measurements. A correlation between irradiation effects and the magnetic measurements is explained based on this study.

  18. Insight on the inconsistencies of Barkhausen signal measurements for radiation damage on nuclear reactor steel

    NASA Astrophysics Data System (ADS)

    Barroso, Soraia Pirfo; Fitzpatrick, Michael E.; Gillemot, Ferenc; Horváth, Marta; Horváth, Ákos; Szekely, Richard

    2014-02-01

    This paper focuses on the use of magnetic measurements, using Barkhausen signals to determine the irradiation effects, attempting to predict fracture toughness changes on nuclear reactor structural materials and correlating these measurements to mechanical testing and microstructure. For this study, two types of nuclear reactor materials were investigated: one sensitive to irradiation effects, the JRQ IAEA's reference material (A533B- -type); and one resistant material, 15KH2MFA WWER's reactor pressure vessel steel. The samples were carefully identified within the original heat block, i.e. forged or rolled plate. These calibrated samples were irradiated at different neutron fluences up to 1023 n/m2. We show how microstructural anisotropy can mask the irradiation effects in the magnetic measurements. A correlation between irradiation effects and the magnetic measurements is explained based on this study.

  19. 46 CFR 59.20-1 - Carbon-steel or alloy-steel castings.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... BOILERS, PRESSURE VESSELS AND APPURTENANCES Welding Repairs to Castings § 59.20-1 Carbon-steel or alloy-steel castings. Defects in carbon-steel or alloy-steel castings may be repaired by welding. The...

  20. 46 CFR 59.20-1 - Carbon-steel or alloy-steel castings.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... BOILERS, PRESSURE VESSELS AND APPURTENANCES Welding Repairs to Castings § 59.20-1 Carbon-steel or alloy-steel castings. Defects in carbon-steel or alloy-steel castings may be repaired by welding. The...

  1. 46 CFR 59.20-1 - Carbon-steel or alloy-steel castings.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... BOILERS, PRESSURE VESSELS AND APPURTENANCES Welding Repairs to Castings § 59.20-1 Carbon-steel or alloy-steel castings. Defects in carbon-steel or alloy-steel castings may be repaired by welding. The...

  2. 46 CFR 59.20-1 - Carbon-steel or alloy-steel castings.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... BOILERS, PRESSURE VESSELS AND APPURTENANCES Welding Repairs to Castings § 59.20-1 Carbon-steel or alloy-steel castings. Defects in carbon-steel or alloy-steel castings may be repaired by welding. The...

  3. 46 CFR 59.20-1 - Carbon-steel or alloy-steel castings.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... BOILERS, PRESSURE VESSELS AND APPURTENANCES Welding Repairs to Castings § 59.20-1 Carbon-steel or alloy-steel castings. Defects in carbon-steel or alloy-steel castings may be repaired by welding. The...

  4. Ion implantations of oxide dispersion strengthened steels

    NASA Astrophysics Data System (ADS)

    Sojak, S.; Simeg Veternikova, J.; Slugen, V.; Petriska, M.; Stacho, M.

    2015-12-01

    This paper is focused on a study of radiation damage and thermal stability of high chromium oxide dispersion strengthened steel MA 956 (20% Cr), which belongs to the most perspective structural materials for the newest generation of nuclear reactors - Generation IV. The radiation damage was simulated by the implantation of hydrogen ions up to the depth of about 5 μm, which was performed at a linear accelerator owned by Slovak University of Technology. The ODS steel MA 956 was available for study in as-received state after different thermal treatments as well as in ions implanted state. Energy of the hydrogen ions chosen for the implantation was 800 keV and the implantation fluence of 6.24 × 1017 ions/cm2. The investigated specimens were measured by non-destructive technique Positron Annihilation Lifetime Spectroscopy in order to study the defect behavior after different thermal treatments in the as-received state and after the hydrogen ions implantation. Although, different resistance to defect production was observed in individual specimens of MA 956 during the irradiation, all implanted specimens contain larger defects than the ones in as-received state.

  5. PLEPS study of ions implanted RAFM steels

    NASA Astrophysics Data System (ADS)

    Sojak, S.; Slugeň, V.; Egger, W.; Ravelli, L.; Petriska, M.; Veterníková, J.; Stacho, M.; Sabelová, V.

    2014-04-01

    Current nuclear power plants (NPP) require radiation, heat and mechanical resistance of their structural materials with the ability to stay operational during NPP planned lifetime. Radiation damage much higher, than in the current NPP, is expected in new generations of nuclear power plants, such as Generation IV and fusion reactors. Investigation of perspective structural materials for new generations of nuclear power plants is among others focused on study of reduced activation ferritic/martensitic (RAFM) steels. These steels have good characteristics as reduced activation, good resistance to volume swelling, good radiation, and heat resistance. Our experiments were focused on the study of microstructural changes of binary Fe-Cr alloys with different chromium content after irradiation, experimentally simulated by ion implantations. Fe-Cr alloys were examined, by Pulsed Low Energy Positron System (PLEPS) at FRM II reactor in Garching (Munich), after helium ion implantations at the dose of 0.1 C/cm2. The investigation was focused on the chromium effect and the radiation defects resistivity. In particular, the vacancy type defects (monovacancies, vacancy clusters) have been studied. Based on our previous results achieved by conventional lifetime technique, the decrease of the defects size with increasing content of chromium is expected also for PLEPS measurements.

  6. Flux effect analysis in WWER-440 reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Kryukov, A.; Blagoeva, D.; Debarberis, L.

    2013-11-01

    The results of long term research programme concerning the determination of irradiation embrittlement dependence on fast neutron flux for WWER-440 reactor pressure vessel steels before and after annealing are presented in this paper. The study of flux effect was carried out on commercial WWER-440 steels which differ significantly in phosphorous (0.013-0.036 wt%) and copper (0.08-0.20 wt%) contents. All specimens were irradiated in surveillance channel positions under similar conditions at high ˜4 × 1012 сm-2 s-1 and low ˜6 × 1011 сm-2 s-1 fluxes (E > 0.5 MeV) at a temperature of 270 °С. The radiation embrittlement was evaluated by transition temperature shift on the basis of Charpy specimens test results. In case of low flux, the measured Tk shifts could be 25-50 °C bigger than the Tk shifts obtained from high flux data. A significant flux effect is observed in WWER-440 reactor pressure vessel steels with higher copper content (>0.13 wt%).

  7. History of ultrahigh carbon steels

    SciTech Connect

    Wadsworth, J.; Sherby, O.D.

    1997-06-20

    The history and development of ultrahigh carbon steels (i.e., steels containing between 1 and 2.l percent C and now known as UHCS) are described. The early use of steel compositions containing carbon contents above the eutectoid level is found in ancient weapons from around the world. For example, both Damascus and Japanese sword steels are hypereutectoid steels. Their manufacture and processing is of interest in understanding the role of carbon content in the development of modern steels. Although sporadic examples of UHCS compositions are found in steels examined in the early part of this century, it was not until the mid-1970s that the modern study began. This study had its origin in the development of superplastic behavior in steels and the recognition that increasing the carbon content was of importance in developing that property. The compositions that were optimal for superplasticity involved the development of steels that contained higher carbon contents than conventional modern steels. It was discovered, however, that the room temperature properties of these compositions were of interest in their own right. Following this discovery, a period of intense work began on understanding their manufacture, processing, and properties for both superplastic forming and room temperature applications. The development of superplastic cast irons and iron carbides, as well as those of laminated composites containing UHCS, was an important part of this history.

  8. Respiratory status of stainless steel and mild steel welders.

    PubMed

    Kalliomäki, P L; Kalliomäki, K; Korhonen, O; Nordman, H; Rahkonen, E; Vaaranen, V

    1982-01-01

    Eighty-three full-time stainless steel and 29 mild steel welders from one shipyard were examined clinically, and their lung function was measured. The stainless steel welders had used both tungsten inert-gas (low-fume concentration) and manual metal-arc (MMA) (high-fume concentration) welding methods. The individual exposure of the welders was estimated based on the time spent doing MMA welding, the amount of retained contaminants in the lungs (magnetopulmography), and urinary chromium excretion. The results suggest that there is a greater prevalence of small airway disease among shipyard mild steel MMA welders than among stainless steel welders. Among the stainless steel welders the impairment of lung function parameters was associated with the MMA welding method. The type of welding, then, is important when the health hazards of welders are studied, and welders cannot be regarded as a single, homogeneous group. PMID:7100838

  9. Microstructure evolution in proton-irradiated austenitic Fe-Cr-Ni alloys under LWR core conditions

    NASA Astrophysics Data System (ADS)

    Gan, Jian

    1999-11-01

    Irradiation-induced microstructure of austenitic stainless steel was investigated using proton irradiation. High-purity alloys of Fe-20Cr-9Ni (UHP 304 SS), Fe-20Cr-24Ni and Ni-18Cr-9Fe were irradiated using 3.2 MeV protons at a dose rate of 7 × 10-6 dpa/s between 300°C and 600°C. The irradiation produced a microstructure consisting of dislocation loops and voids. The dose and temperature dependence of the number density and size of dislocation loops and voids were investigated. The changes in yield strength due to irradiation were estimated from Vickers hardness measurements and compared to calculations using a dispersed-barrier hardening model. The dose and temperature dependence of microstructure and hardness change for proton irradiation follows the same trend as that for neutron irradiation at comparable irradiation conditions. Commercial purity alloys of CP 304 SS and CP 316 SS were irradiated at 360°C to doses between 0.3 and 3.0 dpa. The irradiated microstructure consists of dislocation loops. No voids were detected at doses up to 3.0 dpa. Loop size distributions are in close agreement with that in the same alloys neutron-irradiated in a LWR core. The loop density also agrees with neutron irradiation data. The yield strength as a function of dose in proton irradiated commercial purity alloys is consistent with the neutron- data trend. A fast-reactor microstructure model was adapted for light water reactor (LWR) irradiation conditions (275°C, 7 × 10 -8 dpa/s) and then applied to proton irradiation under conditions (360°C, 7 × 10-6 dpa/s) relevant to LWRs. The original model was modified by including in-cascade interstitial clustering and the loss of interstitial clusters to sinks by cluster diffusion. It was demonstrated that loop nucleation for both LWR irradiation condition and proton irradiation are driven by in-cascade interstitial clustering. One important result from this modeling work is that the difference in displacement cascade between

  10. Economics of food irradiation

    SciTech Connect

    Deitch, J.

    1982-01-01

    This article examines the cost competitiveness of the food irradiation process. An analysis of the principal factors--the product, physical plant, irradiation source, and financing--that impact on cost is made. Equations are developed and used to calculate the size of the source for planned product throughput, efficiency factors, power requirements, and operating costs of sources, radionuclides, and accelerators. Methods of financing and capital investment are discussed. A series of tables show cost breakdowns of sources, buildings, equipment, and essential support facilities for both a cobalt-60 and a 10-MeV electron accelerator facility. Additional tables present irradiation costs as functions of a number of parameters--power input, source size, dose, and hours of annual operation. The use of the numbers in the tables are explained by examples of calculations of the irradiation costs for disinfestation of grains and radicidation of feed.

  11. Stress corrosion cracking of type 304L stainless steel core shroud welds.

    SciTech Connect

    Chung, H. M.; Park, J.-H.; Sanecki, J. E.; Zaluzec, N. J.; Yu, M. S.; Yang, T. T.

    1999-10-26

    Microstructural analyses by advanced metallographic techniques were conducted on mockup welds and a cracked BWR core shroud weld fabricated from Type 304L stainless steel. heat-affected zones of the shroud weld and mockup shielded-metal-arc welds were free of grain-boundary carbide, martensite, delta ferrite, or Cr depletion near grain boundaries. However, as a result of exposure to welding fumes, the heat-affected zones of the welds were significantly contaminated by fluorine and oxygen which migrate to grain boundaries. Significant oxygen contamination promotes fluorine contamination and suppresses classical thermal sensitization, even in Type 304 steels. Results of slow-strain-rate tensile tests indicate that fluorine exacerbates the susceptibility of irradiated steels to intergranular stress corrosion cracking. These observations, combined with previous reports on the strong influence of weld flux, indicate that oxygen and fluorine contamination and fluorine-catalyzed stress corrosion play a major role in cracking of Type 304L stainless steel core shroud welds.

  12. Recent improvements in size effects correlations for DBTT and upper shelf energy of ferritic steels

    SciTech Connect

    Kumar, A.S.; Louden, B.S. ); Garner, F.A.; Hamilton, M.L. )

    1992-01-01

    Currently available correlations for the effects of specimen size on the USE were developed for relatively ductile steels and will not serve as well when the steels become embrittled. Size effects correlations were developed recently for the impact properties of less ductile HT9 to be applied to other initially more ductile steels as they lose their ductility during irradiation. These new correlations successfully predict the ductile brittle transition temperature (DBTT) and the upper shelf energy (USE) of full size Charpy specimens based on subsize specimen data. The new DBTT and the USE correlations were tested against published experimental data on other ferritic steels and shown to perform successfully at lower USE particularly when both precracked and notched only specimens were employed.

  13. Reliability-based condition assessment of steel containment and liners

    SciTech Connect

    Ellingwood, B.; Bhattacharya, B.; Zheng, R.

    1996-11-01

    Steel containments and liners in nuclear power plants may be exposed to aggressive environments that may cause their strength and stiffness to decrease during the plant service life. Among the factors recognized as having the potential to cause structural deterioration are uniform, pitting or crevice corrosion; fatigue, including crack initiation and propagation to fracture; elevated temperature; and irradiation. The evaluation of steel containments and liners for continued service must provide assurance that they are able to withstand future extreme loads during the service period with a level of reliability that is sufficient for public safety. Rational methodologies to provide such assurances can be developed using modern structural reliability analysis principles that take uncertainties in loading, strength, and degradation resulting from environmental factors into account. The research described in this report is in support of the Steel Containments and Liners Program being conducted for the US Nuclear Regulatory Commission by the Oak Ridge National Laboratory. The research demonstrates the feasibility of using reliability analysis as a tool for performing condition assessments and service life predictions of steel containments and liners. Mathematical models that describe time-dependent changes in steel due to aggressive environmental factors are identified, and statistical data supporting the use of these models in time-dependent reliability analysis are summarized. The analysis of steel containment fragility is described, and simple illustrations of the impact on reliability of structural degradation are provided. The role of nondestructive evaluation in time-dependent reliability analysis, both in terms of defect detection and sizing, is examined. A Markov model provides a tool for accounting for time-dependent changes in damage condition of a structural component or system. 151 refs.

  14. [The industrial environment in the electric-furnace steel smelting, converter and open-hearth furnace methods of manufacturing manganese-alloyed steels].

    PubMed

    Karnaukh, N G; Petrov, G A; Gapon, V A; Poslednichenko, I P; Shmidt, S E

    1992-01-01

    Inspection of the environment in manganese-alloyed steel production showed inadequate hygienic conditions of the technological processes employed. Air was more polluted by manganese oxides during the oxygen-converter process though their highest concentrations, 38 times exceeding the MAS, appeared during the casting of steel. An electric furnace coated by dust-noise-proof material and gas cleaning is preferable from a hygienic point of view. The influence of unfavourable microclimate, intensive infrared irradiation and loud noise on workers necessitates automation and mechanization of the process in order to improve the working conditions.

  15. Total lymphoid irradiation

    SciTech Connect

    Sutherland, D.E.; Ferguson, R.M.; Simmons, R.L.; Kim, T.H.; Slavin, S.; Najarian, J.S.

    1983-05-01

    Total lymphoid irradiation by itself can produce sufficient immunosuppression to prolong the survival of a variety of organ allografts in experimental animals. The degree of prolongation is dose-dependent and is limited by the toxicity that occurs with higher doses. Total lymphoid irradiation is more effective before transplantation than after, but when used after transplantation can be combined with pharmacologic immunosuppression to achieve a positive effect. In some animal models, total lymphoid irradiation induces an environment in which fully allogeneic bone marrow will engraft and induce permanent chimerism in the recipients who are then tolerant to organ allografts from the donor strain. If total lymphoid irradiation is ever to have clinical applicability on a large scale, it would seem that it would have to be under circumstances in which tolerance can be induced. However, in some animal models graft-versus-host disease occurs following bone marrow transplantation, and methods to obviate its occurrence probably will be needed if this approach is to be applied clinically. In recent years, patient and graft survival rates in renal allograft recipients treated with conventional immunosuppression have improved considerably, and thus the impetus to utilize total lymphoid irradiation for its immunosuppressive effect alone is less compelling. The future of total lymphoid irradiation probably lies in devising protocols in which maintenance immunosuppression can be eliminated, or nearly eliminated, altogether. Such protocols are effective in rodents. Whether they can be applied to clinical transplantation remains to be seen.

  16. Irradiation Embritlement in Alloy HT-­9

    SciTech Connect

    Serrano De Caro, Magdalena

    2012-08-27

    HT-9 steel is a candidate structural and cladding material for high temperature lead-bismuth cooled fast reactors. In typical advanced fast reactor designs fuel elements will be irradiated for an extended period of time, reaching up to 5-7 years. Significant displacement damage accumulation in the steel is expected (> 200 dpa) when exposed to dpa-rates of 20-30 dpa{sub Fe}/y and high fast flux (E > 0.1 MeV) {approx}4 x 10{sup 15} n/cm{sup 2}s. Core temperatures could reach 400-560 C, with coolant temperatures at the inlet as low as 250 C, depending on the reactor design. Mechanical behavior in the presence of an intense fast flux and high dose is a concern. In particular, low temperature operation could be limited by irradiation embrittlement. Creep and corrosion effects in liquid metal coolants could set a limit to the upper operating temperature. In this report, we focus on the low temperature operating window limit and describe HT-9 embrittlement experimental findings reported in the literature that could provide supporting information to facilitate the consideration of a Code Case on irradiation effects for this class of steels in fast reactor environments. HT-9 has an extensive database available on irradiation performance, which makes it the best choice as a possible near-term candidate for clad, and ducts in future fast reactors. Still, as it is shown in this report, embrittlement data for very low irradiation temperatures (< 200 C) and very high radiation exposure (> 150 dpa) is scarce. Experimental findings indicate a saturation of DBTT shifts as a function of dose, which could allow for long lifetime cladding operation. However, a strong increase in DBTT shift with decreasing irradiation temperature could compromise operation at low service temperatures. Development of a deep understanding of the physics involved in the radiation damage mechanisms, together with multiscale computer simulation models of irradiation embrittlement will provide the basis to

  17. Stainless Steel Permeability

    SciTech Connect

    Buchenauer, Dean A.; Karnesky, Richard A.

    2015-09-01

    An understanding of the behavior of hydrogen isotopes in materials is critical to predicting tritium transport in structural metals (at high pressure), estimating tritium losses during production (fission environment), and predicting in-vessel inventory for future fusion devices (plasma driven permeation). Current models often assume equilibrium diffusivity and solubility for a class of materials (e.g. stainless steels or aluminum alloys), neglecting trapping effects or, at best, considering a single population of trapping sites. Permeation and trapping studies of the particular castings and forgings enable greater confidence and reduced margins in the models. For FY15, we have continued our investigation of the role of ferrite in permeation for steels of interest to GTS, through measurements of the duplex steel 2507. We also initiated an investigation of the permeability in work hardened materials, to follow up on earlier observations of unusual permeability in a particular region of 304L forgings. Samples were prepared and characterized for ferrite content and coated with palladium to prevent oxidation. Issues with the poor reproducibility of measurements at low permeability were overcome, although the techniques in use are tedious. Funding through TPBAR and GTS were secured for a research grade quadrupole mass spectrometer (QMS) and replacement turbo pumps, which should improve the fidelity and throughput of measurements in FY16.

  18. Characterization of mechanical properties and microstructure of highly irradiated SS 316

    NASA Astrophysics Data System (ADS)

    Karthik, V.; Kumar, RanVijay; Vijayaragavan, A.; Venkiteswaran, C. N.; Anandaraj, V.; Parameswaran, P.; Saroja, S.; Muralidharan, N. G.; Joseph, Jojo; Kasiviswanathan, K. V.; Jayakumar, T.; Raj, Baldev

    2013-08-01

    Cold worked austenitic stainless steel type AISI 316 is used as the material for fuel cladding and wrapper of the Fast Breeder Test Reactor (FBTR), India. The evaluation of mechanical properties of these core structurals is very essential to assess its integrity and ensure safe and productive operation of FBTR to very high burn-ups. The changes in the mechanical properties of these core structurals are associated with microstructural changes caused by high fluence neutron irradiation and temperatures of 673-823 K. Remote tensile testing has been used for evaluating the tensile properties of irradiated clad tubes and shear punch test using small disk specimens for evaluating the properties of irradiated hexagonal wrapper. This paper will highlight the methods employed for evaluating the mechanical properties of the irradiated cladding and wrapper and discuss the trends in properties as a function of dpa (displacement per atom) and irradiation temperature.

  19. Principles and practice of a bellows-loaded compact irradiation vehicle

    NASA Astrophysics Data System (ADS)

    Byun, Thak Sang; Li, Meimei; Snead, Lance L.; Katoh, Yutai; Burchell, Timothy D.; McDuffee, Joel L.

    2013-08-01

    This article describes the key design principles and application of a mini-bellows loaded irradiation creep frame technology developed for use in the high flux isotope reactor (HFIR). For this irradiation vehicle, the bellows, frame, sample, and temperature monitor are contained within a hydraulic or fixed "rabbit" capsule of a few inches in length. Of critical importance and key to this technology is the viability and stability of the metallic bellows under the elevated temperature irradiation environment. Conceptual design and supporting analysis have been performed for tension and compression specimens. Benchtop verification has substantiated the modeling regarding the ability of the bellows to produce sufficient stress to induce irradiation creep in subsize specimens. Discussion focuses on the possible stress ranges in specimens induced by the miniature gas-pressurized bellows and the limitations imposed by the size and structure of thin-walled bellows. A brief discussion of pre- and post-irradiation measurement of the integrity of load frames irradiated to 4.4 × 1025 n/m2 (E > 0.1 MeV) is presented. Following this protocol, the pre-irradiation loading to a sample is determined and post-irradiation loading inferred. An in-reactor creep testing technology using pressurized mini-bellows has been established for irradiation creep tests in tensile or compressive mode using the HFIR rabbit capsule design. Results from theoretical calculation and in-furnace tests confirmed that the pressurized bellows-loaded miniature creep frame can produce enough thrust force to induce irradiation creep in subsize specimens. Bellows materials, types and dimensions were selected considering in-reactor integrity, load transferring function, weldability, and in-reactor stability. Both stainless steel and IN 718 mini-bellows were proven to be capable of irradiation creep testing. A practical process for the testing and evaluation of applied stress has been developed and applied to in

  20. Process for dezincing galvanized steel

    DOEpatents

    Morgan, William A.; Dudek, Frederick J.; Daniels, Edward J.

    1998-01-01

    A process for removing zinc from galvanized steel. The galvanized steel is immersed in an electrolyte containing at least about 15% by weight of sodium or potassium hydroxide and having a temperature of at least about 75.degree. C. and the zinc is galvanically corroded from the surface of the galvanized steel. The material serving as the cathode is principally a material having a standard electrode potential which is intermediate of the standard electrode potentials of zinc and cadmium in the electrochemical series. The corrosion rate may be accelerated by (i) increasing the number density of corrosion sites in the galvanized steel by mechanically abrading or deforming the galvanized steel, (ii) heating the galvanized steel to form an alloy of zinc on the surface of the galvanized steel, (iii) mixing the galvanized steel with a material having a standard electrode potential which is intermediate of the standard electrode potentials of zinc and cadmium in the electrochemical series, or (iv) moving the galvanized steel relative to itself and to the electrolyte while immersed in the electrolyte.

  1. Process for dezincing galvanized steel

    DOEpatents

    Morgan, W.A.; Dudek, F.J.; Daniels, E.J.

    1998-07-14

    A process is described for removing zinc from galvanized steel. The galvanized steel is immersed in an electrolyte containing at least about 15% by weight of sodium or potassium hydroxide and having a temperature of at least about 75 C and the zinc is galvanically corroded from the surface of the galvanized steel. The material serving as the cathode is principally a material having a standard electrode potential which is intermediate of the standard electrode potentials of zinc and cadmium in the electrochemical series. The corrosion rate may be accelerated by (1) increasing the number density of corrosion sites in the galvanized steel by mechanically abrading or deforming the galvanized steel, (2) heating the galvanized steel to form an alloy of zinc on the surface of the galvanized steel, (3) mixing the galvanized steel with a material having a standard electrode potential which is intermediate of the standard electrode potentials of zinc and cadmium in the electrochemical series, or (4) moving the galvanized steel relative to itself and to the electrolyte while immersed in the electrolyte. 1 fig.

  2. Induction heat treatment of steel

    SciTech Connect

    Semiatin, S.L.; Stutz, D.E.

    1985-01-01

    This book discusses the induction heating. After reviewing heat treating operations for steel and the principles of the heat treatment of steel, an overview of induction heat treating is provided. Next, consideration is given to equipment and equipment selection, coil design, power requirements and temperature control. A discussion of surface and through hardening of steel is provided, including information on frequency and power selection and quenching apparatus. Tempering is considered, followed by information on control of residual stresses, cracking, temper brittleness and the important metallurgical and hardness differences between induction and furnace treated steel.

  3. High strength, tough alloy steel

    DOEpatents

    Thomas, Gareth; Rao, Bangaru V. N.

    1979-01-01

    A high strength, tough alloy steel is formed by heating the steel to a temperature in the austenite range (1000.degree.-1100.degree. C.) to form a homogeneous austenite phase and then cooling the steel to form a microstructure of uniformly dispersed dislocated martensite separated by continuous thin boundary films of stabilized retained austenite. The steel includes 0.2-0.35 weight % carbon, at least 1% and preferably 3-4.5% chromium, and at least one other substitutional alloying element, preferably manganese or nickel. The austenite film is stable to subsequent heat treatment as by tempering (below 300.degree. C.) and reforms to a stable film after austenite grain refinement.

  4. Irradiation and food processing.

    PubMed

    Sigurbjörnsson, B; Loaharanu, P

    1989-01-01

    After more than four decades of research and development, food irradiation has been demonstrated to be safe, effective and versatile as a process of food preservation, decontamination or disinfection. Its various applications cover: inhibition of sprouting of root crops; insect disinfestation of stored products, fresh and dried food; shelf-life extension of fresh fruits, vegetables, meat and fish; destruction of parasites and pathogenic micro-organisms in food of animal origin; decontamination of spices and food ingredients, etc. Such applications provide consumers with the increase in variety, volume and value of food. Although regulations on food irradiation in different countries are largely unharmonized, national authorities have shown increasing recognition and acceptance of this technology based on the Codex Standard for Irradiated Foods and its associated Code of Practice. Harmonization of national legislations represents an important prerequisite to international trade in irradiated food. Consumers at large are still not aware of the safety and benefits that food irradiation has to offer. Thus, national and international organizations, food industry, trade associations and consumer unions have important roles to play in introducing this technology based on its scientific values. Public acceptance of food irradiation may be slow at the beginning, but should increase at a faster rate in the foreseeable future when consumers are well informed of the safety and benefits of this technology in comparison with existing ones. Commercial applications of food irradiation has already started in 18 countries at present. The volume of food or ingredients treated on a commercial scale varies from country to country ranging from several tons of spices to hundreds of thousands of tons of grains per annum. With the increasing interest of national authorities and the food industry in applying the process, it is anticipated that some 25 countries will use some 55 commercial

  5. Silicon carbide tritium permeation barrier for steel structural components.

    SciTech Connect

    Causey, Rion A.; Garde, Joseph Maurico; Buchenauer, Dean A.; Calderoni, Pattrick; Holschuh, Thomas, Jr.; Youchison, Dennis Lee; Wright, Matt; Kolasinski, Robert D.

    2010-09-01

    Chemical vapor deposited (CVD) silicon carbide (SiC) has superior resistance to tritium permeation even after irradiation. Prior work has shown Ultrametfoam to be forgiving when bonded to substrates with large CTE differences. The technical objectives are: (1) Evaluate foams of vanadium, niobium and molybdenum metals and SiC for CTE mitigation between a dense SiC barrier and steel structure; (2) Thermostructural modeling of SiC TPB/Ultramet foam/ferritic steel architecture; (3) Evaluate deuterium permeation of chemical vapor deposited (CVD) SiC; (4) D testing involved construction of a new higher temperature (> 1000 C) permeation testing system and development of improved sealing techniques; (5) Fabricate prototype tube similar to that shown with dimensions of 7cm {theta} and 35cm long; and (6) Tritium and hermeticity testing of prototype tube.

  6. Corrosion testing of stainless steel-zirconium metal waste forms

    SciTech Connect

    Abraham, D.P.; Simpson, L.J.; Devries, M.J.; McDeavitt, S.M.

    1999-07-01

    Stainless steel-zirconium (SS-Zr) alloys have been developed as waste forms for the disposal of metallic waste generated during the electrometallurgical treatment of spent nuclear fuel. The waste forms incorporate irradiated cladding hulls, components of the alloy fuel, noble metal fission products, and actinide elements. The baseline waste form is a stainless steel-15 wt% zirconium (SS-15Zr) alloy. This article presents microstructures and some of the corrosion studies being conducted on the waste form alloys. Electrochemical corrosion, immersion corrosion, and vapor hydration tests have been performed on various alloy compositions to evaluate corrosion behavior and resistance to selective leaching of simulated fission products. The SS-Zr waste forms immobilize and retain fission products very effectively and show potential for acceptance as high-level nuclear waste forms.

  7. Application of laser in seam welding of dissimilar steel to aluminium joints for thick structural components

    NASA Astrophysics Data System (ADS)

    Meco, S.; Pardal, G.; Ganguly, S.; Williams, S.; McPherson, N.

    2015-04-01

    Laser welding-brazing technique, using a continuous wave (CW) fibre laser with 8000 W of maximum power, was applied in conduction mode to join 2 mm thick steel (XF350) to 6 mm thick aluminium (AA5083-H22), in a lap joint configuration with steel on the top. The steel surface was irradiated by the laser and the heat was conducted through the steel plate to the steel-aluminium interface, where the aluminium melts and wets the steel surface. The welded samples were defect free and the weld micrographs revealed presence of a brittle intermetallic compounds (IMC) layer resulting from reaction of Fe and Al atoms. Energy Dispersive Spectroscopy (EDS) analysis indicated the stoichiometry of the IMC as Fe2Al5 and FeAl3, the former with maximum microhardness measured of 1145 HV 0.025/10. The IMC layer thickness varied between 4 to 21 μm depending upon the laser processing parameters. The IMC layer showed an exponential growth pattern with the applied specific point energy (Esp) at a constant power density (PD). Higher PD values accelerate the IMC layer growth. The mechanical shear strength showed a narrow band of variation in all the samples (with the maximum value registered at 31.3 kN), with a marginal increase in the applied Esp. This could be explained by the fact that increasing the Esp results into an increase in the wetting and thereby the bonded area in the steel-aluminium interface.

  8. Neutron induced damage in reactor pressure vessel steel: An X-ray absorption fine structure study

    NASA Astrophysics Data System (ADS)

    Kuri, G.; Cammelli, S.; Degueldre, C.; Bertsch, J.; Gavillet, D.

    2009-03-01

    The radiation damage produced in reactor pressure vessel (RPV) steels during neutron irradiation is a long-standing problem of considerable practical interest. In this study, an extended X-ray absorption fine structure (EXAFS) spectroscopy has been applied at Cu, Ni and Mn K-edges to systematically investigate neutron induced radiation damage to the metal-site bcc structure of RPV steels, irradiated with neutrons in the fluence range from 0.85 to 5.0 × 1019 cm-2. An overall similarity of Cu, Ni and Mn atomic environment in the iron matrix is observed. The radial distribution functions (RDFs), derived from EXAFS data have been found to evolve continuously as a function of neutron fluence describing the atomic-scale structural modifications in RPVs by neutron irradiations. From the pristine data, long range order beyond the first- and second-shell is apparent in the RDF spectra. In the irradiated specimens, all near-neighbour peaks are greatly reduced in magnitude, typical of damaged material. Prolonged annealing leads annihilation of point defects to give rise to an increase in the coordination numbers of near-neighbour atomic shells approaching values close to that of non-irradiated material, but does not suppress the formation of nano-sized Cu and/or Ni-rich-precipitates. Total amount of radiation damage under a given irradiation condition has been determined. The average structural parameters estimated from the EXAFS data are presented and discussed.

  9. NERVA irradiation program. GTR 23, volume 1: Combined effects of reactor radiation and cryogenic temperature on NERVA structural materials

    NASA Technical Reports Server (NTRS)

    Mcdaniel, R. H.; Bradford, E. W.; Lewis, J. H.; Wattier, J. B.

    1973-01-01

    Specimens fabricated from structural materials that were candidates for certain NERVA applications were irradiated in liquid nitrogen (LN2), liquid hydrogen (LH2), water, and air. The specimens irradiated in LN2 were stored in LN2 and finally tested in LN2, or at some higher temperature in a few instances. The specimens irradiated in LH2 underwent an unplanned warmup while in storage so this portion of the test was lost; some specimens were tested in LN2 but none were tested in LH2. The Ground Test Reactor was the radiation source. The test specimens consisted mainly of tensile and fracture toughness specimens of several different materials, but other types of specimens such as tear, flexure, springs, and lubricant were also irradiated. Materials tested include Hastelloy X, Al, Ni steel, steel, Be, ZrC, Ti-6Al-4V, CuB, and Ti-5Al-2.5Sn.

  10. Unrestrained swelling of uranium-nitride fuel irradiated at temperatures ranging from 1100 to 1400 K (1980 to 2520 R)

    NASA Technical Reports Server (NTRS)

    Rohal, R. G.; Tambling, T. N.

    1973-01-01

    Six fuel pins were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a stainless steel (type 304L) clad. The pins were irradiated for approximately 4000 hours to burnups of about 2.0 atom percent uranium. The average clad surface temperature during irradiation was about 1100 K (1980 deg R). Since stainless steel has a very low creep strength relative to that of UN at this temperature, these tests simulated unrestrained swelling of UN. The tests indicated that at 1 percent uranium atom burnup the unrestrained diametrical swelling of UN is about 0.5, 0.8, and 1.0 percent at 1223, 1264, and 1306 K (2200, deg 2273 deg, and 2350 deg R), respectively. The tests also indicated that the irradiation induced swelling of unrestrained UN fuel pellets appears to be isotropic.

  11. HRTEM Study of the Role of Nanoparticles in ODS Ferritic Steel

    SciTech Connect

    Hsiung, L; Tumey, S; Fluss, M; Serruys, Y; Willaime, F

    2011-08-30

    Structures of nanoparticles and their role in dual-ion irradiated Fe-16Cr-4.5Al-0.3Ti-2W-0.37Y{sub 2}O{sub 3} (K3) ODS ferritic steel produced by mechanical alloying (MA) were studied using high-resolution transmission electron microscopy (HRTEM) techniques. The observation of Y{sub 4}Al{sub 2}O{sub 9} complex-oxide nanoparticles in the ODS steel imply that decomposition of Y{sub 2}O{sub 3} in association with internal oxidation of Al occurred during mechanical alloying. HRTEM observations of crystalline and partially crystalline nanoparticles larger than {approx}2 nm and amorphous cluster-domains smaller than {approx}2 nm provide an insight into the formation mechanism of nanoparticles/clusters in MA/ODS steels, which we believe involves solid-state amorphization and re-crystallization. The role of nanoparticles/clusters in suppressing radiation-induced swelling is revealed through TEM examinations of cavity distributions in (Fe + He) dual-ion irradiated K3-ODS steel. HRTEM observations of helium-filled cavities (helium bubbles) preferably trapped at nanoparticle/clusters in dual-ion irradiated K3-ODS are presented.

  12. ORNL irradiation creep facility

    SciTech Connect

    Reiley, T.C.; Auble, R.L.; Beckers, R.M.; Bloom, E.E.; Duncan, M.G.; Saltmarsh, M.J.; Shannon, R.H.

    1980-09-01

    A machine was developed at ORNL to measure the rates of elongation observed under irradiation in stressed materials. The source of radiation is a beam of 60 MeV alpha particles from the Oak Ridge Isochronous Cyclotron (ORIC). This choice allows experiments to be performed which simulate the effects of fast neutrons. A brief review of irradiation creep and experimental constraints associated with each measurement technique is given. Factors are presented which lead to the experimental choices made for the Irradiation Creep Facility (ICF). The ICF consists of a helium-filled chamber which houses a high-precision mechanical testing device. The specimen to be tested must be thermally stabilized with respect to the temperature fluctuations imposed by the particle beam which passes through the specimen. Electrical resistance of the specimen is the temperature control parameter chosen. Very high precision in length measurement and temperature control are required to detect the small elongation rates relevant to irradiation creep in the test periods available (approx. 1 day). The apparatus components and features required for the above are presented in some detail, along with the experimental procedures. The damage processes associated with light ions are discussed and displacement rates are calculated. Recent irradiation creep results are given, demonstrating the suitability of the apparatus for high resolution experiments. Also discussed is the suitability of the ICF for making high precision thermal creep measurements.

  13. Irradiation and post-irradiation examination of uranium-free nitride fuel

    NASA Astrophysics Data System (ADS)

    Hania, P. R.; Klaassen, F. C.; Wernli, B.; Streit, M.; Restani, R.; Ingold, F.; Fedorov, A. V.; Wallenius, J.

    2015-11-01

    Two identical Phénix-type 15-15Ti steel pinlets each containing a 70 mm Pu0.3Zr0.7N fuel stack in a 1-bar helium atmosphere have been irradiated in the HFR Petten at medium high linear power (46-47 kW/m at BOL) and an average cladding temperature of 505 °C. The pins were irradiated to a plutonium burn-up of 9.7% (88 MWd/kgHM) in 170 full power days. Both pins remained fully intact. Post-irradiation examination performed at NRG and PSI showed that the overall swelling rate of the fuel was 0.92 vol-%/%FIHMA. Fission gas release was 5-6%, while helium release was larger than 50%. No fuel restructuring was observed, and only mild cracking. EPMA measurements show a burn-up increase toward the pellet edge of up to 4 times. All investigated fission products except to some extent the noble metals were found to be evenly distributed over the matrix, indicating good solubility. Local formation of a secondary phase with high Pu content and hardly any Zr was observed. A general conclusion of this investigation is that ZrN is a suitable inert matrix for burning plutonium at high destruction rates.

  14. Femtosecond laser color marking stainless steel surface with different wavelengths

    NASA Astrophysics Data System (ADS)

    Li, Guoqiang; Li, Jiawen; Hu, Yanlei; Zhang, Chenchu; Li, Xiaohong; Chu, Jiaru; Huang, Wenhao

    2015-03-01

    The femtosecond laser color marking stainless steel surfaces with different incident wavelengths were investigated theoretically and experimentally. It indicates that the spectral regions of the colors firstly increase and then reduce with increasing spatial periods of the ripples induced by laser irradiation. Additionally, the colors are gradually changed from blue to red due to the elongation of the diffracted light wavelengths. As a result, the color effects are distinctly different. This study offers a new controllable parameter to produce diverse colors, which may find a wide range of applications in the laser color marking, art designing and so on.

  15. Influence of structural-phase state of ferritic-martensitic steels on the helium porosity development

    NASA Astrophysics Data System (ADS)

    Chernov, I. I.; Staltsov, M. S.; Kalin, B. A.; Bogachev, I. A.; Guseva, L. Yu; Dzhumaev, P. S.; Emelyanova, O. V.; Drozhzhina, M. V.; Manukovsky, K. V.; Nikolaeva, I. D.

    2016-04-01

    Transmission electron microscopy (TEM) has been used to study the effect of the initial structural-phase state (SPhS) of ferritic-martensitic steels EK-181, EP-450 and EP-450- ODS (with 0.5 wt.% nanoparticles of Y2O3) on the of helium porosity formation and gas swelling. Different SPhS of steel EK-181 was produced by water quenching, annealing, normalizing plus tempered, intensive plastic deformation by torsion (HPDT). Irradiation was carried out by He+-40 keV ions at 923 K up to fluence of 5-1020 He+/m2. It is shown that the water quenching causes the formation of uniformly distributed small bubbles (d¯ ∼ 2 nm) of the highest density (ρ∼ 1025 m-3). After normalization followed by tempering as well as after annealing bubbles distribution is highly non-uniform both by volume and in size. Very large faceted bubbles (pre-equilibrium gas-filled voids) are formed in ferrite grains resulting in high level of gas swelling of the irradiated layer with S = 4,9 ± 1,2 and 3.8 ± 0.9% respectively. Nano- and microcrystalline structure created by HPDT completely degenerate at irradiation temperature and ion irradiation formed bubbles of the same parameters as in the annealed steel. Bubbles formed in EP-450-ODS steel are smaller in size and density, which led to a decrease of helium swelling by 4 times (S = 0.8 ± 0.2%) as compared to the swelling of the matrix steel EP-450 (S = 3.1 ± 0.7%).

  16. Development of new ferritic steels as cladding material for metallic fuel fast breeder reactor

    NASA Astrophysics Data System (ADS)

    Tokiwai, Moriyasu; Horie, Masaaki; Kako, Kenji; Fujiwara, Masayuki

    1993-09-01

    The excellent thermal, chemical and neutronic properties of metallic fuel (U-Pu-Zr alloy) will lead to drastic improvements in fast reactor safety and the related fuel cycle economy. Some new high molybdenum 12Cr ferritic stainless steel candidate cladding alloys have been designed to achieve the mechanical properties required for high performance metallic fuel elements. These candidate claddings were irradiated by ion bombardment and tested to determine their strength and creep rupture properties. A 12Cr-8Mo and a 12Cr-8Mo-0.1Y 2O 3 steel were fabricated into cladding via a powder metallurgy process and by a mechanical alloying process, respectively. These claddings had two and three times the creep rupture strength (pressurized at 650°C for 10000 h) of a conventional 12Cr ferritic steel (HT-9). These two steels also showed no void formation up to 350 dpa by Ni 3+ irradiation. A zircaloy-2 lined steel cladding tube has also been fabricated for the purpose of reducing fuel-cladding interdiffusion and chemical interaction.

  17. Development of a robust modeling tool for radiation-induced segregation in austenitic stainless steels

    SciTech Connect

    Yang, Ying; Field, Kevin G; Allen, Todd R.; Busby, Jeremy T

    2015-09-01

    Irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels in Light Water Reactor (LWR) components has been linked to changes in grain boundary composition due to irradiation induced segregation (RIS). This work developed a robust RIS modeling tool to account for thermodynamics and kinetics of the atom and defect transportation under combined thermal and radiation conditions. The diffusion flux equations were based on the Perks model formulated through the linear theory of the thermodynamics of irreversible processes. Both cross and non-cross phenomenological diffusion coefficients in the flux equations were considered and correlated to tracer diffusion coefficients through Manning’s relation. The preferential atomvacancy coupling was described by the mobility model, whereas the preferential atom-interstitial coupling was described by the interstitial binding model. The composition dependence of the thermodynamic factor was modeled using the CALPHAD approach. Detailed analysis on the diffusion fluxes near and at grain boundaries of irradiated austenitic stainless steels suggested the dominant diffusion mechanism for chromium and iron is via vacancy, while that for nickel can swing from the vacancy to the interstitial dominant mechanism. The diffusion flux in the vicinity of a grain boundary was found to be greatly influenced by the composition gradient formed from the transient state, leading to the oscillatory behavior of alloy compositions in this region. This work confirms that both vacancy and interstitial diffusion, and segregation itself, have important roles in determining the microchemistry of Fe, Cr, and Ni at irradiated grain boundaries in austenitic stainless steels.

  18. The industrial ecology of steel

    SciTech Connect

    Considine, Timothy J.; Jablonowski, Christopher; Considine, Donita M.M.; Rao, Prasad G.

    2001-03-26

    This study performs an integrated assessment of new technology adoption in the steel industry. New coke, iron, and steel production technologies are discussed, and their economic and environmental characteristics are compared. Based upon detailed plant level data on cost and physical input-output relations by process, this study develops a simple mathematical optimization model of steel process choice. This model is then expanded to a life cycle context, accounting for environmental emissions generated during the production and transportation of energy and material inputs into steelmaking. This life-cycle optimization model provides a basis for evaluating the environmental impacts of existing and new iron and steel technologies. Five different plant configurations are examined, from conventional integrated steel production to completely scrap-based operations. Two cost criteria are used to evaluate technology choice: private and social cost, with the latter including the environmental damages associated with emissions. While scrap-based technologies clearly generate lower emissions in mass terms, their emissions of sulfur dioxide and nitrogen oxides are significantly higher. Using conventional damage cost estimates reported in the literature suggests that the social costs associated with scrap-based steel production are slightly higher than with integrated steel production. This suggests that adopting a life-cycle viewpoint can substantially affect environmental assessment of new technologies. Finally, this study also examines the impacts of carbon taxes on steel production costs and technology choice.

  19. FOOD IRRADIATION REACTOR

    DOEpatents

    Leyse, C.F.; Putnam, G.E.

    1961-05-01

    An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

  20. Economics of food irradiation

    NASA Astrophysics Data System (ADS)

    Kunstadt, Peter; Eng, P.; Steeves, Colyn; Beaulieu, Daniel; Eng, P.

    1993-07-01

    The number of products being radiation processed worldwide is constantly increasing and today includes such diverse items as medical disposables, fruits and vegetables, spices, meats, seafoods and waste products. This range of products to be processed has resulted in a wide range of irradiator designs and capital and operating cost requirements. This paper discusses the economics of low dose food irradiation applications and the effects of various parameters on unit processing costs. It provides a model for calculating specific unit processing costs by correlating known capital costs with annual operating costs and annual throughputs. It is intended to provide the reader with a general knowledge of how unit processing costs are derived.

  1. Fuel or irradiation subassembly

    DOEpatents

    Seim, O.S.; Hutter, E.

    1975-12-23

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

  2. The impact of transmutant helium on weldability of austenitic steel

    NASA Astrophysics Data System (ADS)

    Fabritsiev, S. A.; Pokrovsky, A. S.; Brovko, V. A.

    1996-10-01

    The results of the investigation of 0.05-0.15 dpa neutron irradiation impact on the Cr16Ni11Mo3Ti austenitic steel weldability are presented. Samples were irradiated to doses of 10 20 n/cm 2 and 3 × 10 20 n/cm 2 ( E > 0.1 MeV) in the RBT-10 reactor, thus providing helium accumulation of 1 appm and 2.5 appm, respectively. Flat samples, 1 mm in thickness, were welded by an automatic device for argon arc welding in a hot chamber. Low-cycle fatigue (LCF) testing in bending was used to assess impact of helium on the degradation of welded joint properties. LCF tests showed that the transmutant helium accumulation resulted in a decrease in the number of cycles to failure at Ttest = 20°C and 350°C. It is concluded that repeated welding will present in the repair of ITER materials.

  3. EFFECTS OF GAMMA IRRADIATION ON EPDM ELASTOMERS (REVISION 1)

    SciTech Connect

    Clark, E.

    2013-09-13

    Two formulations of EPDM elastomer, one substituting a UV stabilizer for the normal antioxidant in this polymer, and the other the normal formulation, were synthesized and samples of each were exposed to gamma irradiation in initially pure deuterium gas to compare their radiation stability. Stainless steel containers having rupture disks were designed for this task. After 130 MRad dose of cobalt-60 radiation in the SRNL Gamma Irradiation Facility, a significant amount of gas was created by radiolysis; however the composition indicated by mass spectroscopy indicated an unexpected increase in the total amount deuterium in both formulations. The irradiated samples retained their ductility in a bend test. No change of sample weight, dimensions, or density was observed. No change of the glass transition temperature as measured by dynamic mechanical analysis was observed, and most of the other dynamic mechanical properties remained unchanged. There appeared to be an increase in the storage modulus of the irradiated samples containing the UV stabilizer above the glass transition, which may indicate hardening of the material by radiation damage. Revision 1 adds a comparison with results of a study of tritium exposed EPDM. The amount of gas produced by the gamma irradiation was found to be equivalent to about 280 days exposure to initially pure tritium gas at one atmosphere. The glass transition temperature of the tritium exposed EPDM rose about 10°C. over 280 days, while no glass transition temperature change was observed for gamma irradiated EPDM. This means that gamma irradiation in deuterium cannot be used as a surrogate for tritium exposure.

  4. Characterization of an orthovoltage biological irradiator used for radiobiological research.

    PubMed

    Azimi, Rezvan; Alaei, Parham; Spezi, Emiliano; Hui, Susanta K

    2015-05-01

    Orthovoltage irradiators are routinely used to irradiate specimens and small animals in biological research. There are several reports on the characteristics of these units for small field irradiations. However, there is limited knowledge about use of these units for large fields, which are essential for emerging large-field irregular shape irradiations, namely total marrow irradiation used as a conditioning regimen for hematological malignancies. This work describes characterization of a self-contained Orthovoltage biological irradiator for large fields using measurements and Monte Carlo simulations that could be used to compute the dose for in vivo or in vitro studies for large-field irradiation using this or a similar unit. Percentage depth dose, profiles, scatter factors, and half-value layers were measured and analyzed. A Monte Carlo model of the unit was created and used to generate depth dose and profiles, as well as scatter factors. An ion chamber array was also used for profile measurements of flatness and symmetry. The output was determined according to AAPM Task Group 61 guidelines. The depth dose measurements compare well with published data for similar beams. The Monte Carlo-generated depth dose and profiles match our measured doses to within 2%. Scatter factor measurements indicate gradual variation of these factors with field size. Dose rate measured by placing the ion chamber atop the unit's steel plate or solid water indicate enhanced readings of 5 to 28% compared with those measured in air. The stability of output over a 5-year period is within 2% of the 5-year average. PMID:25694476

  5. Experimental atomic scale investigation of irradiation effects in CW 316SS and UFG-CW 316SS

    NASA Astrophysics Data System (ADS)

    Pareige, P.; Etienne, A.; Radiguet, B.

    2009-06-01

    Materials of the core internals of pressurized water reactor (austenitic stainless steels) are subject to neutron irradiation. To understand the ageing mechanisms associated with irradiation and propose life predictions of components or develop new materials, irradiation damage needs to be experimentally investigated. Atomic scale investigation of a neutron-irradiated CW316 SS with the laser pulsed atom probe gives a detailed description of the solute segregation in the austenitic grains. In order to understand the mechanism of solute segregation detected in the neutron-irradiated materials, ion irradiations were performed. These latest irradiations were realized on a CW 316SS as well as on a nanostructured CW 316SS. The study of irradiation effects in a nanograin material allows first, to easily analyse grain boundary segregation and second, to test the behaviour under irradiation of a new nanostructured material. The three aspects of this atomic scale investigation (neutron irradiation effect, model ion irradiation, new nanostructured CW 316 SS) are tackled in this paper.

  6. Structures and properties of rapidly solidified 9Cr-lMo steel

    NASA Astrophysics Data System (ADS)

    Megusar, J.; Lavernia, E.; Domalavage, P.; Harling, O. K.; Grant, N. J.

    1984-05-01

    Irradiation induced shifts of the DBTT and possible hydrogen embrittlement of ferritic steels are currently considered major problems for CTR applications. Rapid solidification and in particular liquid dynamic compaction (LDC) has been studied in developing 9Cr-1Mo steel as a candidate first wall material. Structural refinements such as reduction of segregation, fine grain size and fine size of second phase particles are retained in this process and this will have a favorable effect on fracture properties. LDC has also the potential of preparing first wall components directly from the melt and this would have an economic advantage over conventional ingot technology.

  7. Synergistic Effect of Triple Ion Beams on Radiation Damage in CLAM Steel

    NASA Astrophysics Data System (ADS)

    Yuan, Da-Qing; Zheng, Yong-Nan; Zuo, Yi; Fan, Ping; Zhou, Dong-Mei; Zhang, Qiao-Li; Ma, Xiao-Qiang; Cui, Bao-Qun; Chen, Li-Hua; Jiang, Wei-Sheng; Wu, Yi-Can; Huang, Qun-Ying; Peng, Lei; Cao, Xing-Zhong; Wang, Bao-Yi; Wei, Long; Zhu, Sheng-Yun

    2014-04-01

    The synergistic effect of triple ion beams is investigated by simultaneous and sequential irradiations of gold, hydrogen and helium ions on the low activation martensitic steel (CLAM) developed in China. The depth profile measurements of the positron annihilation Doppler broadening S parameter are carried out as a function of slow-positron beam energy to examine the produced radiation damage. The synergistic effect of displacement damage and hydrogen and helium on the formation of radiation damage is clearly observed. In the preset case, this effect suppresses the radiation damage in the CLAM steel due to the helium and/or hydrogen filling of vacancy clusters.

  8. EFFECTS OF GAMMA IRRADIATION ON EPDM ELASTOMERS

    SciTech Connect

    Clark, E.

    2011-09-22

    Two formulations of EPDM elastomer, one substituting a UV stabilizer for the normal antioxidant in this polymer, and the other the normal formulation, were synthesized and samples of each were exposed to gamma irradiation in initially pure deuterium gas to compare their radiation stability. Stainless steel containers having rupture disks were designed for this task. After 130 MRad dose of cobalt-60 radiation in the SRNL Gamma Irradiation Facility, a significant amount of gas was created by radiolysis; however the composition indicated by mass spectroscopy indicated an unexpected increase in the total amount deuterium in both formulations. The irradiated samples retained their ductility in a bend test. No change of sample weight, dimensions, or density was observed. No change of the glass transition temperature as measured by dynamic mechanical analysis was observed, and most of the other dynamic mechanical properties remained unchanged. There appeared to be an increase in the storage modulus of the irradiated samples containing the UV stabilizer above the glass transition, which may indicate hardening of the material by radiation damage. Polymeric materials become damaged by exposure over time to ionizing radiation. Despite the limited lifetime, polymers have unique engineering material properties and polymers continue to be used in tritium handling systems. In tritium handling systems, polymers are employed mainly in joining applications such as valve sealing surfaces (eg. Stem tips, valve packing, and O-rings). Because of the continued need to employ polymers in tritium systems, over the past several years, programs at the Savannah River National Laboratory have been studying the effect of tritium on various polymers of interest. In these studies, samples of materials of interest to the SRS Tritium Facilities (ultra-high molecular weight polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE, Teflon{reg_sign}), Vespel{reg_sign} polyimide, and the elastomer

  9. Connections: Superplasticity, Damascus Steels, Laminated Steels, and Carbon Dating

    NASA Astrophysics Data System (ADS)

    Wadsworth, Jeffrey

    2016-09-01

    In this paper, a description is given of the connections that evolved from the initial development of a family of superplastic plain carbon steels that came to be known as Ultra-High Carbon Steels (UHCS). It was observed that their very high carbon contents were similar, if not identical, to those of Damascus steels. There followed a series of attempts to rediscover how the famous patterns found on Damascus steels blades were formed. At the same time, in order to improve the toughness at room temperature of the newly-developed UHCS, laminated composites were made of alternating layers of UHCS and mild steel (and subsequently other steels and other metals). This led to a study of ancient laminated composites, the motives for their manufacture, and the plausibility of some of the claims relating to the number of layers in the final blades. One apparently ancient laminated composite, recovered in 1837 from the great pyramid of Giza which was constructed in about 2750 B.C., stimulated a carbon dating study of ancient steels. The modern interest in "Bladesmithing" has connections back to many of these ancient weapons.

  10. High-power laser applications in Nippon Steel Corporation

    NASA Astrophysics Data System (ADS)

    Minamida, Katsuhiro

    2000-02-01

    The laser, which was invented in 1960, has been developed using various substances of solids, liquids, gases and semiconductors as laser active media. Applications of laser utilizing the coherent properties of laser light and the high power density light abound in many industries and in heavy industries respectively. The full-scale use of lasers in the steel industry began nearly 23 years ago with their applications as controllable light sources. Its contribution to the increase in efficiency and quality of the steel making process has been important and brought us the saving of the energy, the resource and the labor. Laser applications in the steel making process generally require high input energy, so it is essential to consider the interaction between the laser beam and the irradiated material. In particular, the reflectivity of the laser beam on the surface of material and the quantity of the laser-induced plasma are critical parameters for high efficient processes with low energy losses. We have developed plenty of new laser systems for the steel making process with their considerations in mind. A review of the following high-power-laser applications is given in the present paper: (1) Use of plasma as a secondary heat source in CO2 laser welding for connecting steel sheets of various grades. (2) Laser-assisted electric resistance welding of pipes. (3) New type all-laser-welded honeycomb panels for high-speed transport. (4) Laser flying welder for continuous hot rolling mill using two 45 kW CO2 lasers.

  11. Microstructural observations of HFIR-irratiated austenitic stainless steels including welds from JP9-16

    SciTech Connect

    Sawai, T.; Shiba, K.; Hishinuma, A.

    1996-04-01

    Austenitic stainless steels, including specimens taken from various electron beam (EB) welds, have been irradiated in HFIR Phase II capsules, JP9-16. Fifteen specimens irradiated at 300, 400, and 500{degrees}C up to 17 dpa are so far examined by a transmission electron microscope (TEM). In 300{degrees}C irradiation, cavities were smaller than 2nm and different specimens showed little difference in cavity microstructure. At 400{degrees}C, cavity size was larger, but still very small (<8 nm). At 500{degrees}C, cavity size reached 30 nm in weld metal specimens of JPCA, while cold worked JPCA contained a small (<5 nm) cavities. Inhomogeneous microstructural evolution was clearly observed in weld-metal specimens irradiated at 500{degrees}C.

  12. NSUF Irradiated Materials Library

    SciTech Connect

    Cole, James Irvin

    2015-09-01

    The Nuclear Science User Facilities has been in the process of establishing an innovative Irradiated Materials Library concept for maximizing the value of previous and on-going materials and nuclear fuels irradiation test campaigns, including utilization of real-world components retrieved from current and decommissioned reactors. When the ATR national scientific user facility was established in 2007 one of the goals of the program was to establish a library of irradiated samples for users to access and conduct research through competitively reviewed proposal process. As part of the initial effort, staff at the user facility identified legacy materials from previous programs that are still being stored in laboratories and hot-cell facilities at the INL. In addition other materials of interest were identified that are being stored outside the INL that the current owners have volunteered to enter into the library. Finally, over the course of the last several years, the ATR NSUF has irradiated more than 3500 specimens as part of NSUF competitively awarded research projects. The Logistics of managing this large inventory of highly radioactive poses unique challenges. This document will describe materials in the library, outline the policy for accessing these materials and put forth a strategy for making new additions to the library as well as establishing guidelines for minimum pedigree needed to be included in the library to limit the amount of material stored indefinitely without identified value.

  13. Update on meat irradiation

    SciTech Connect

    Olson, D.G.

    1997-12-01

    The irradiation of meat and poultry in the United States is intended to eliminate pathogenic bacteria from raw product, preferably after packaging to prevent recontamination. Irradiation will also increase the shelf life of raw meat and poultry products approximately two to three times the normal shelf life. Current clearances in the United States are for poultry (fresh or frozen) at doses from 1.5 to 3.0 kGy and for fresh pork at doses from 0.3 to 1.0 kGy. A petition for the clearance of all red meat was submitted to the Food and Drug Administration (FDA) in July 1994. The petition is for clearances of fresh meat at doses from 1.5 to 4.5 kGy and for frozen meat at {approximately}2.5 to 7.5 kGy. Clearance for red meat is expected before the end of 1997. There are 28 countries that have food irradiation clearances, of which 18 countries have clearances for meat or poultry. However, there are no uniform categories or approved doses for meat and poultry among the countries that could hamper international trade of irradiated meat and poultry.

  14. Irradiating insect pests

    Technology Transfer Automated Retrieval System (TEKTRAN)

    This is a non-technical article focusing on phytosanitary uses of irradiation. In a series of interview questions, I present information on the scope of the invasive species problem and the contribution of international trade in agricultural products to the movement of invasive insects. This is foll...

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

    SciTech Connect

    Huang, F.H.

    1992-02-01

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

  16. Occupational Profiles in the European Steel Industry.

    ERIC Educational Resources Information Center

    Franz, Hans-Werner; And Others

    The steel industry in Europe has faced great changes, with resulting layoffs and restructuring. Now that the most basic changes seem to be over, it has become evident that the remaining steel industry requires more highly trained workers than was the case previously. Although steel maintenance employees were always highly skilled, steel production…

  17. Improving the toughness of ultrahigh strength steel

    SciTech Connect

    Soto, Koji

    2002-08-15

    The ideal structural steel combines high strength with high fracture toughness. This dissertation discusses the toughening mechanism of the Fe/Co/Ni/Cr/Mo/C steel, AerMet 100, which has the highest toughness/strength combination among all commercial ultrahigh strength steels. The possibility of improving the toughness of this steel was examined by considering several relevant factors.

  18. Influence of Atmospheric Pressure Torch Plasma Irradiation on Plant Growth

    NASA Astrophysics Data System (ADS)

    Akiyoshi, Yusuke; Hayashi, Nobuya; Kitazaki, Satoshi; Koga, Kazunori; Shiratani, Masaharu

    2011-10-01

    Growth stimulation characteristics of plants seeds are investigated by an atmospheric discharge irradiation into plasma seeds. Atmospheric pressure plasma torch is consisted of alumina ceramics tube and the steel mesh electrodes wind inside and outside of the tube. When AC high voltage (8 kHz) is applied to the electrode gap, the barrier discharge plasma is produced inside the alumina ceramics tube. The barrier discharge plasma is blown outside with the gas flow in ceramics tube. Radish sprouts seeds locate at 1 cm from the torch edge. The growth stimulation was observed in the length of a stem and a root after the plasma irradiation. The stem length increases approximately 2.8 times at the cultivation time of 24 h. And the growth stimulation effect is found to be maintained for 40 h, after sowing seeds. The mechanism of the growth stimulation would be the redox reaction inside plant cells induced by oxygen radicals.

  19. TEM Examination of Advanced Alloys Irradiated in ATR

    SciTech Connect

    Jian Gan, PhD

    2007-09-01

    Successful development of materials is critical to the deployment of advanced nuclear power systems. Irradiation studies of candidate materials play a vital role for better understanding materials performance under various irradiation environments of advanced system designs. In many cases, new classes of materials have to be investigated to meet the requirements of these advanced systems. For applications in the temperature range of 500 800ºC which is relevant to the fast neutron spectrum burner reactors for the Global Nuclear Energy Partnership (GNEP) program, oxide dispersion strengthened (ODS) and ferritic martensitic steels (e.g., MA957 and others) are candidates for advanced cladding materials. In the low temperature regions of the core (<600ºC), alloy 800H, HCM12A (also called T 122) and HT 9 have been considered.

  20. Validation of the shear punch-tensile correlation technique using irradiated materials

    SciTech Connect

    Hankin, G.L.; Faulkner, R.G.; Toloczko, M.B.; Hamilton, M.L.

    1998-03-01

    It was recently demonstrated that tensile data could be successfully related to shear punch data obtained on transmission electron microscopy (TEM) discs for a variety of irradiated alloys exhibiting yield strengths that ranged from 100 to 800 MPa. This implies that the shear punch test might be a viable alternative for obtaining tensile properties using a TEM disk, which is much smaller than even the smallest miniature tensile specimens, especially when irradiated specimens are not available or when they are too radioactive to handle easily. The majority of the earlier tensile-shear punch correlation work was done using a wide variety of unirradiated materials. The current work extends this correlation effort to irradiated materials and demonstrates that the same relationships that related shear punch tests remain valid for irradiated materials. Shear punch tests were performed on two sets of specimens. In the first group, three simple alloys from the {sup 59}Ni isotopic doping series in the solution annealed and cold worked conditions were irradiated at temperatures ranging from 365 to 495 C in the Fast Flux Test Facility. The corresponding tensile data already existed for tensile specimens fabricated from the same raw materials and irradiated side-by-side with the disks. In the second group, three variants of 316 stainless steel were irradiated in FFTF at 5 temperatures between 400 and 730 C to doses ranging from 12.5 to 88 dpa. The specimens were in the form of both TEM and miniature tensile specimens and were irradiated side-by-side.

  1. University of Wisconsin Ion Beam Laboratory: A facility for irradiated materials and ion beam analysis

    NASA Astrophysics Data System (ADS)

    Field, K. G.; Wetteland, C. J.; Cao, G.; Maier, B. R.; Dickerson, C.; Gerczak, T. J.; Field, C. R.; Kriewaldt, K.; Sridharan, K.; Allen, T. R.

    2013-04-01

    The University of Wisconsin Ion Beam Laboratory (UW-IBL) has recently undergone significant infrastructure upgrades to facilitate graduate level research in irradiated materials phenomena and ion beam analysis. A National Electrostatics Corp. (NEC) Torodial Volume Ion Source (TORVIS), the keystone upgrade for the facility, can produce currents of hydrogen ions and helium ions up to ˜200 μA and ˜5 μA, respectively. Recent upgrades also include RBS analysis packages, end station developments for irradiation of relevant material systems, and the development of an in-house touch screen based graphical user interface for ion beam monitoring. Key research facilitated by these upgrades includes irradiation of nuclear fuels, studies of interfacial phenomena under irradiation, and clustering dynamics of irradiated oxide dispersion strengthened steels. The UW-IBL has also partnered with the Advanced Test Reactor National Scientific User Facility (ATR-NSUF) to provide access to the irradiation facilities housed at the UW-IBL as well as access to post irradiation facilities housed at the UW Characterization Laboratory for Irradiated Materials (CLIM) and other ATR-NSUF partner facilities. Partnering allows for rapid turnaround from proposed research to finalized results through the ATR-NSUF rapid turnaround proposal system. An overview of the UW-IBL including CLIM and relevant research is summarized.

  2. A positron beam Doppler broadening analysis of formation and recovery of defects produced by ion irradiation in Fe‒C‒Cu alloys

    NASA Astrophysics Data System (ADS)

    Iwai, Takeo

    2013-04-01

    Mechanisms of radiation embrittlement of reactor pressure vessel steels remain to be fully understood, particularly the nature of so-called 'matrix defects'. One possible mechanism is vacancy cluster formation, probably assisted by cascade damage. In order to investigate the effect of copper on the formation and annealing processes of vacancy clusters, ion-irradiated Fe‒C and Fe‒C‒Cu were investigated using a variable energy positron beam. Doppler broadening analysis revealed that vacancy-type defects are produced by ion irradiation and that copper addition reduces the open volume of the defects. Post irradiation annealing suggested the vacancy clusters do not have a substantial role in irradiation hardening.

  3. Advanced steel reheat furnace

    SciTech Connect

    Moyeda, D.; Sheldon, M.; Koppang, R.; Lanyi, M.; Li, X.; Eleazer, B.

    1997-10-01

    Energy and Environmental Research Corp. (EER) under a contract from the Department of Energy is pursuing the development and demonstration of an Advanced Steel Reheating Furnace. This paper reports the results of Phase 1, Research, which has evaluated an advanced furnace concept incorporating two proven and commercialized technologies previously applied to other high temperature combustion applications: EER`s gas reburn technology (GR) for post combustion NOx control; and Air Product`s oxy-fuel enrichment air (OEA) for improved flame heat transfer in the heating zones of the furnace. The combined technologies feature greater production throughput with associated furnace efficiency improvements; lowered NOx emissions; and better control over the furnace atmosphere, whether oxidizing or reducing, leading to better control over surface finish.

  4. Heat transfer analysis of staphylococcus aureus on stainless steel with microwave radiation.

    PubMed

    Yeo, C B; Watson, I A; Stewart-Tull, D E; Koh, V H

    1999-09-01

    Staphylococcus aureus (NCTC 6571; Oxford strain) on stainless steel discs was exposed to microwave radiation at 2450 MHz and up to 800 W. Cell viability was reduced as the exposure time increased, with complete bacterial inactivation at 110 s, attaining a temperature of 61.4 degrees C. The low rate of temperature rise, RT, of the bacterial suspension as compared with sterile distilled water or nutrient broth suggests a significant influence of the microwave sterilization efficacy on the thermal properties of the micro-organisms. The heat transfer kinetics of thermal microwave irradiation suggest that the micro-organism has a power density at least 51-fold more than its surrounding liquid suspension. When the inoculum on the stainless steel disc was subjected to microwave radiation, heat conduction from the stainless steel to the inoculum was the cause of bacteriostasis with power absorbed at 23.8 W for stainless steel and 0.16 W for the bacteria-liquid medium. This report shows that the microwave killing pattern of Staph. aureus on stainless steel was mainly due to heat transfer from the stainless steel substrate and very little direct energy was absorbed from the microwaves. PMID:10540242

  5. 2169 steel waveform experiments.

    SciTech Connect

    Furnish, Michael David; Alexander, C. Scott; Reinhart, William Dodd; Brown, Justin L.

    2012-11-01

    In support of LLNL efforts to develop multiscale models of a variety of materials, we have performed a set of eight gas gun impact experiments on 2169 steel (21% Cr, 6% Ni, 9% Mn, balance predominantly Fe). These experiments provided carefully controlled shock, reshock and release velocimetry data, with initial shock stresses ranging from 10 to 50 GPa (particle velocities from 0.25 to 1.05 km/s). Both windowed and free-surface measurements were included in this experiment set to increase the utility of the data set, as were samples ranging in thickness from 1 to 5 mm. Target physical phenomena included the elastic/plastic transition (Hugoniot elastic limit), the Hugoniot, any phase transition phenomena, and the release path (windowed and free-surface). The Hugoniot was found to be nearly linear, with no indications of the Fe phase transition. Releases were non-hysteretic, and relatively consistent between 3- and 5-mmthick samples (the 3 mm samples giving slightly lower wavespeeds on release). Reshock tests with explosively welded impactors produced clean results; those with glue bonds showed transient releases prior to the arrival of the reshock, reducing their usefulness for deriving strength information. The free-surface samples, which were steps on a single piece of steel, showed lower wavespeeds for thin (1 mm) samples than for thicker (2 or 4 mm) samples. A configuration used for the last three shots allows release information to be determined from these free surface samples. The sample strength appears to increase with stress from ~1 GPa to ~ 3 GPa over this range, consistent with other recent work but about 40% above the Steinberg model.

  6. Corrosion of Steels in Steel Reinforced Concrete in Cassava Juice

    NASA Astrophysics Data System (ADS)

    Oluwadare, G. O.; Agbaje, O.

    The corrosion of two types of construction steels, ST60Mn and RST37-2♦, in a low cyanide concentration environment (cassava juice) and embedded in concrete had been studied. The ST60 Mn was found to be more corrosion resistant in both ordinary water and the cassava juice environment. The cyanide in cassava juice does not attack the steel but it provides an environment of lower pH around the steel in the concrete which leads to breakdown of the passivating film provided by hydroxyl ions from cement. Other factors such as the curing time of the concrete also affect the corrosion rates of the steel in the concrete. The corrosion rate of the steel directly exposed to cassava juice i.e., steel not embedded in concrete is about twice that in concrete. Long exposure of concrete structure to cassava processing effluent might result in deterioration of such structures. Careful attention should therefore be paid to disposal of cassava processing effluents, especially in a country like Nigeria where such processing is now on the increase.

  7. IN-SITU MEASUREMENT OF TRITIUM PERMEATION THROUGH STAINLESS STEEL

    SciTech Connect

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin; Longhurst, Glen R.

    2013-06-01

    The TMIST-2 irradiation experiment was conducted in the Advanced Test Reactor at Idaho National Laboratory to evaluate tritium permeation through Type 316 stainless steel (316 SS). The interior of a 316 SS seamless tube specimen was exposed to a 4He carrier gas mixed with a specified quantity of tritium (T2) to yield partial pressures of 0.1, 5, and 50 Pa at 292° and 330°C. In-situ tritium permeation measurements were made by passing a He-Ne sweep gas over the outer surface of the specimen to carry the permeated tritium to a bubbler column for liquid scintillation counting. An irradiation enhancement factor (IEF) was determined by comparing in-situ permeation data with a correlation for ex-reactor hydrogen permeation through austenitic stainless steel developed from literature data and reported by Le Claire. Nominal values for the IEF ranged between 3 and 5 for 316 SS. In-situ permeation data were also used to derive an in-reactor permeation correlation as a function of temperature and pressure. In addition, the triton recoil contribution to tritium permeation, which results from the transmutation of 3He to T, was also evaluated by introducing a 4He carrier gas mixed with 3He at a partial pressure of 1013 Pa at 330°C. Less than 3% of the tritium resulting from 3He transmutation contributed to tritium permeation.

  8. In Situ Measurement of Tritium Permeation Through Stainless Steel

    SciTech Connect

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin; Longhurst, Glen

    2013-06-01

    The TMIST-2 irradiation experiment was conducted in the Advanced Test Reactor at Idaho National Laboratory to evaluate tritium permeation through Type 316 stainless steel (316 SS). The interior of a 316 SS seamless tube specimen was exposed to a 4He carrier gas mixed with a specified quantity of tritium (T2) to yield partial pressures of 0.1, 5, and 50 Pa at 292° and 330°C. In-situ tritium permeation measurements were made by passing a He-Ne sweep gas over the outer surface of the specimen to carry the permeated tritium to a bubbler column for liquid scintillation counting. An irradiation enhancement factor (IEF) was determined by comparing in-situ permeation data with a correlation for ex-reactor hydrogen permeation through austenitic stainless steel developed from literature data and reported by Le Claire. Nominal values for the IEF ranged between 3 and 5 for 316 SS. In-situ permeation data were also used to derive an in-reactor permeation correlation as a function of temperature and pressure. In addition, the triton recoil contribution to tritium permeation, which results from the transmutation of 3He to T, was also evaluated by introducing a 4He carrier gas mixed with 3He at a partial pressure of 1013 Pa at 330°C. Less than 3% of the tritium resulting from 3He transmutation contributed to tritium permeation.

  9. Dosimetric evaluation of hybrid brass/stainless-steel apertures for proton therapy.

    PubMed

    Chen, Hao; Matysiak, Witold; Flampouri, Stella; Slopsema, Roelf; Li, Zuofeng

    2014-09-01

    In passive scattering proton therapy, patient specific collimators (apertures) are used to laterally shape the proton beam, and compensators are employed to distally conform proton dose to the target. Brass is a commonly used material for apertures and recently a hybrid brass/stainless-steel (BR/SST) aperture design has been introduced to reduce treatment cost without clinical flow change. We measured stopping power and leakage dose for apertures made of stainless steel and brass in the Proton Therapy system. The linear stopping power ratios for stainless steel (type 304) and brass to water were calculated to be 5.46 and 5.51, respectively. Measured stopping power ratios of SST and BR were 5.51  ±  0.04 and 5.56  ±  0.08, respectively, which agrees with the calculated values within 1%. Leakage dose on the downstream surface of two slabs of Ø18 cm stainless steel apertures (total thickness of 6.5 cm) for the maximum available proton energy (235 MeV) was 1.283% ± 0.004% of the prescription dose, and was smaller compared to the 1.358% ± 0.005% leakage dose measured for existing brass apertures of identical physical dimensions. Therefore, the existing beam range limits for brass aperture slabs used at our institution with safety margin allowances for material composition and delivered beam range uncertainties can be safely applied for the new BR/SST aperture design. Potential range differences in the brass and stainless steel interface regions of the hybrid design were further investigated using EBT3 GafChromic film. Film dosimetry revealed no discernible range variations across the brass and stainless steel interface regions. Neutron dose to the patient from brass and stainless steel apertures was simulated using the Monte Carlo method. The results indicate that stainless steel produces similar patient neutron dose compared to brass. Material activation dose rates of stainless steel were measured over a period of 7 d after irradiation. The

  10. ELECTRON IRRADIATION OF SOLIDS

    DOEpatents

    Damask, A.C.

    1959-11-01

    A method is presented for altering physical properties of certain solids, such as enhancing the usefulness of solids, in which atomic interchange occurs through a vacancy mechanism, electron irradiation, and temperature control. In a centain class of metals, alloys, and semiconductors, diffusion or displacement of atoms occurs through a vacancy mechanism, i.e., an atom can only move when there exists a vacant atomic or lattice site in an adjacent position. In the process of the invention highenergy electron irradiation produces additional vacancies in a solid over those normally occurring at a given temperature and allows diffusion of the component atoms of the solid to proceed at temperatures at which it would not occur under thermal means alone in any reasonable length of time. The invention offers a precise way to increase the number of vacancies and thereby, to a controlled degree, change the physical properties of some materials, such as resistivity or hardness.

  11. A polycrystal plasticity model of strain localization in irradiated iron

    NASA Astrophysics Data System (ADS)

    Barton, Nathan R.; Arsenlis, Athanasios; Marian, Jaime

    2013-02-01

    At low to intermediate homologous temperatures, the degradation of structural materials performance in nuclear environments is associated with high number densities of nanometric defects produced in irradiation cascades. In polycrystalline ferritic materials, self-interstitial dislocations loops are a principal signature of irradiation damage, leading to a mechanical response characterized by increased yield strengths, decreased total strain to failure, and decreased work hardening as compared to the unirradiated behavior. Above a critical defect concentration, the material deforms by plastic flow localization, giving rise to strain softening in terms of the engineering stress-strain response. Flow localization manifests itself in the form of defect-depleted crystallographic channels, through which all dislocation activity is concentrated. In this paper, we describe the formulation of a crystal plasticity model for pure Fe embedded in a finite element polycrystal simulator and present results of uniaxial tensile deformation tests up to 10% strain. We use a tensorial damage descriptor variable to capture the evolution of the irradiation damage loop subpopulation during deformation. The model is parameterized with detailed dislocation dynamics simulations of tensile tests up to 1.5% deformation of systems containing various initial densities of irradiation defects. The coarse-grained simulations are shown to capture the essential details of the experimental stress response observed in ferritic alloys and steels. Our methodology provides an effective linkage between the defect scale, of the order of one nanometer, and the continuum scale involving multiple grain orientations.

  12. Great Lakes Steel -- PCI facility

    SciTech Connect

    Eichinger, F.T.; Dake, S.H.; Wagner, E.D.; Brown, G.S.

    1997-12-31

    This paper discusses the planning, design, and start-up of the 90 tph PCI facility for National Steel`s Great Lakes Steel Division in River Rouge, MI. This project is owned and operated by Edison Energy Services, and was implemented on a fast-track basis by Raytheon Engineers and Constructors, Babcock Material Handling, and Babcock and Wilcox. This paper presents important process issues, basic design criteria, an the challenges of engineering and building a state-of-the-art PCI facility in two existing plants. Pulverized coal is prepared at the River Rouge Power Plant of Detroit Edison, is pneumatically conveyed 6,000 feet to a storage silo at Great Lakes Steel, and is injected into three blast furnaces.

  13. BIOLOGICAL IRRADIATION FACILITY

    DOEpatents

    McCorkle, W.H.; Cern, H.S.

    1962-04-24

    A facility for irradiating biological specimens with neutrons is described. It includes a reactor wherein the core is off center in a reflector. A high-exposure room is located outside the reactor on the side nearest the core while a low-exposure room is located on the opposite side. Means for converting thermal neutrons to fast neutrons are movably disposed between the reactor core and the high and low-exposure rooms. (AEC)

  14. Hydrogen embrittlement of structural steels.

    SciTech Connect

    Somerday, Brian P.

    2010-06-01

    Carbon-manganese steels are candidates for the structural materials in hydrogen gas pipelines, however it is well known that these steels are susceptible to hydrogen embrittlement. Decades of research and industrial experience have established that hydrogen embrittlement compromises the structural integrity of steel components. This experience has also helped identify the failure modes that can operate in hydrogen containment structures. As a result, there are tangible ideas for managing hydrogen embrittement in steels and quantifying safety margins for steel hydrogen containment structures. For example, fatigue crack growth aided by hydrogen embrittlement is a key failure mode for steel hydrogen containment structures subjected to pressure cycling. Applying appropriate structural integrity models coupled with measurement of relevant material properties allows quantification of safety margins against fatigue crack growth in hydrogen containment structures. Furthermore, application of these structural integrity models is aided by the development of micromechanics models, which provide important insights such as the hydrogen distribution near defects in steel structures. The principal objective of this project is to enable application of structural integrity models to steel hydrogen pipelines. The new American Society of Mechanical Engineers (ASME) B31.12 design code for hydrogen pipelines includes a fracture mechanics-based design option, which requires material property inputs such as the threshold for rapid cracking and fatigue crack growth rate under cyclic loading. Thus, one focus of this project is to measure the rapid-cracking thresholds and fatigue crack growth rates of line pipe steels in high-pressure hydrogen gas. These properties must be measured for the base materials but more importantly for the welds, which are likely to be most vulnerable to hydrogen embrittlement. The measured properties can be evaluated by predicting the performance of the pipeline

  15. Hyperparathyroidism after neck irradiation.

    PubMed

    Christmas, T J; Chapple, C R; Noble, J G; Milroy, E J; Cowie, A G

    1988-09-01

    A retrospective review of 1550 cases of hyperparathyroidism (HPT) treated surgically over a 30-year period reveals a past history of exposure to neck irradiation in 10 cases (0.7 per cent). The indication for radiotherapy was benign disease in nine and papillary thyroid carcinoma in one case. The mean interval between radiation exposure and the detection of HPT was 32 years (range 3-63 years). Patients treated with radioactive iodine alone developed HPT after a mean of 5 years while the interval for those treated with external beam therapy alone was a mean of 44 years. The parathyroid histology was adenoma in six cases, carcinoma in three cases and nodular hyperplasia in one case. All patients had coincident benign thyroid disease apart from one that had previously had papillary carcinoma and another with follicular carcinoma. Neck irradiation has been shown to confer an increased risk of HPT due to parathyroid adenoma and carcinoma. Radiotherapy for benign disease has generally been abandoned and these cases demonstrate a further contra-indication for the use of neck irradiation.

  16. Heat-to-heat variability of irradiation creep and swelling of HT9 irradiated to high neutron fluence at 400-600{degrees}C

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.

    1996-10-01

    Irradiation creep data on ferritic/martensitic steels are difficult and expensive to obtain, and are not available for fusion-relevant neutron spectra and displacement rates. Therefore, an extensive creep data rescue and analysis effort is in progress to characterize irradiation creep of ferritic/martensitic alloys in other reactors and to develop a methodology for applying it to fusion applications. In the current study, four tube sets constructed from three nominally similar heats of HT9 subjected to one of two heat treatments were constructed as helium-pressurized creep tubes and irradiated in FFTF-MOTA at four temperatures between 400 and 600{degrees}C. Each of the four heats exhibited a different stress-free swelling behavior at 400{degrees}C, with the creep rate following the swelling according to the familiar B{sub o} + DS creep law. No stress-free swelling was observed at the other three irradiation temperatures. Using a stress exponent of n = 1.0 as the defining criterion, {open_quotes}classic{close_quotes} irradiation creep was found at all temperatures, but, only over limited stress ranges that decreased with increasing temperature. The creep coefficient B{sub o} is a little lower ({approx}50%) than that observed for austenitic steel, but the swelling-creep coupling coefficient D is comparable to that of austenitic steels. Primary transient creep behavior was also observed at all temperatures except 400{degrees}C, and thermal creep behavior was found to dominate the deformation at high stress levels at 550 and 600{degrees}C.

  17. Tritium retention in reduced-activation ferritic/martensitic steels

    SciTech Connect

    Hatano, Y.; Abe, S.; Matsuyama, M.; Alimov, V.K.; Spitsyn, A.V.; Bobyr, N.P.; Cherkez, D.I.; Khripunov, B.I.; Golubeva, A.V.; Ogorodnikova, O.V.; Klimov, N.S.; Chernov, V.M.; Oyaidzu, M.; Yamanishi, T.

    2015-03-15

    Reduced-activation ferritic/martensitic (RAFM) steels are structural material candidates for breeding blankets of future fusion reactors. Therefore, tritium (T) retention in RAFM steels is an important problem in assessing the T inventory of blankets. In this study, specimens of RAFM steels were subjected to irradiation of 20 MeV W ions to 0.54 displacements per atom (dpa), exposure to high flux D plasmas at 400 and 600 K and that to pulsed heat loads. The specimens thus prepared were exposed to DT gas at 473 K. Despite severe modification in the surface morphology, heat loads had negligible effects on T retention. Significant increase in T retention at the surface and/or subsurface was observed after D plasma exposure. However, T trapped at the surface/subsurface layer was easily removed by maintaining the specimens in the air at about 300 K. Displacement damage led to increase in T retention in the bulk due to the trapping effects of defects, and T trapped was stable at 300 K. It was therefore concluded that displacement damages had the largest influence on T retention under the present conditions.

  18. Heavy-Section Steel Technology Program fracture issues

    SciTech Connect

    Pennell, W.E.

    1991-01-01

    Large-scale fracture mechanics tests have resulted in the identification of a number of fracture-technology issues. Identification of additional issues has come from the reactor vessel materials-irradiation test program and from reactor operating experience. This paper provides a review of fracture issues with an emphasis on their potential impact on a reactor vessel pressurized-thermal-shock (PTS) analysis. Mixed-mode crack propagation emerges as a major issue, due in large measure to the poor performance of existing models for the prediction of ductile tearing. Rectification of ductile tearing technology deficiencies may require extending the technology to include a more complete treatment of stress-state and loading history effects. The effect of cladding on vessel fracture remains uncertain to the point that it is not possible to determine at this time if the net effect will be positive or negative. Enhanced fracture toughness for shallow flaws has been demonstrated for low-strength structural steels. Demonstration of a similar effect in reactor pressure vessel steels could have a significant beneficial effect on the probabilistic analysis of reactor vessel fracture. Further development of existing fracture-mechanics models and concepts is required to meet the special requirements for fracture evaluation of circumferential flaws in the welds of ring-forged vessels. Fracture technology advances required to address the issues discussed in this paper are the major objective for the ongoing Heavy-Section Steel Technology (HSST) Program at Oak Ridge National Laboratory (ORNL). 22 refs., 18 figs.

  19. Laser Treatment on the Coating Surface Having Been Performed by Means of Plasma Surfacing With Powder Made of M2 Steel

    NASA Astrophysics Data System (ADS)

    Khaidarova, A. A.; Silantiev, S. A.

    2016-08-01

    In this study researches were carried out about the impact of pulsed laser irradiation on the surface of M2 steel which had been surfaced beforehand on steel 20 by poweder-pasma surfacing technique. The surface treatment was performed by the single point laser action and by successive imposition of overlapped impulses. During the surfacing the average irradiation power was varied from 15.8 to 21.0 W. The impulse duration in all points was constant and it was 7 msec. In this study it was researched the influence of laser-beam power change on the depth of penetration, changes in the structure and microhardness of treated areas.

  20. Wear of steel by rubber

    NASA Technical Reports Server (NTRS)

    Gent, A. N.; Pulford, C. T. R.

    1978-01-01

    Wear of a steel blade used as a scraper to abrade rubber surfaces has been found to take place much more rapidly on a cis-polyisoprene (natural rubber) surface than on a cis-polybutadiene surface, and much more rapidly in an inert atmosphere than in air. These observations are attributed to the direct attack upon steel of free-radical species generated by mechanical rupture of elastomer molecules during abrasion.