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

  1. Silicon's role in determining swelling in neutron-irradiated Fe-Cr-Ni-Si alloys

    SciTech Connect

    Sekimura, N. ); Garner, F. A. ); Newkirk, J.W. )

    1991-11-01

    Two silicon-modified alloy series, one based on Fe-15Cr-20Ni and another based on Fe-15Cr-25Ni were irradiated at target temperatures between 399 and 649{degree}C in EBR-II. The influence of silicon on swelling is more complex than previously envisioned and indicates that silicon plays two or more competing roles while in solution. Radiation-induced formation of {gamma}{prime} (Ni{sub 3}Si) precipitates is dependent on silicon and nickel content, as well as temperature. Precipitation of {gamma}{prime} appears to play only a minor role in void formation.

  2. Silicon`s role in determining swelling in neutron-irradiated Fe-Cr-Ni-Si alloys

    SciTech Connect

    Sekimura, N.; Garner, F. A.; Newkirk, J.W.

    1991-11-01

    Two silicon-modified alloy series, one based on Fe-15Cr-20Ni and another based on Fe-15Cr-25Ni were irradiated at target temperatures between 399 and 649{degree}C in EBR-II. The influence of silicon on swelling is more complex than previously envisioned and indicates that silicon plays two or more competing roles while in solution. Radiation-induced formation of {gamma}{prime} (Ni{sub 3}Si) precipitates is dependent on silicon and nickel content, as well as temperature. Precipitation of {gamma}{prime} appears to play only a minor role in void formation.

  3. 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.

  4. Welding irradiated stainless steel

    SciTech Connect

    Kanne, W.R. Jr.; Chandler, G.T.; Nelson, D.Z.; Franco-Ferreira, E.A.

    1993-12-31

    Conventional welding processes produced severe underbead cracking in irradiated stainless steel containing 1 to 33 appm helium from n,a reactions. A shallow penetration overlay technique was successfully demonstrated for welding irradiated stainless steel. The technique was applied to irradiated 304 stainless steel that contained 10 appm helium. Surface cracking, present in conventional welds made on the same steel at the same and lower helium concentrations, was eliminated. Underbead cracking was minimal compared to conventional welding methods. However, cracking in the irradiated material was greater than in tritium charged and aged material at the same helium concentrations. The overlay technique provides a potential method for repair or modification of irradiated reactor materials.

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

    SciTech Connect

    Yamamoto, Yukinori; Brady, Michael P; Lu, Zhao Ping; Liu, Chain T; Takeyama, Masao; Maziasz, Philip J; Pint, Bruce A

    2007-01-01

    Creep strengthening of Al-modified austenitic stainless steels by MC carbides or Fe{sub 2}Nb 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 Al{sub 2}O{sub 3} 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 {gamma} (fcc) + {alpha} (bcc) dual phase due to the strong {alpha}-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 {gamma}-Fe and still capable of alumina scale formation. Alloys that relied solely on Fe{sub 2}Nb 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 Fe{sub 2}Nb were of limited benefit to creep resistance due to the solution limit of NbC within the {gamma}-Fe matrix of the alloys studied. However, when combined with other MC-type strengtheners, such as V{sub 4}C{sub 3} or TiC, higher levels of creep resistance were obtained.

  6. Neutron Irradiation Resistance of RAFM Steels

    SciTech Connect

    Gaganidze, Ermile; Dafferner, Bernhard; Aktaa, Jarir

    2008-07-01

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

  7. Helium irradiation induced hardening in MNHS steels

    NASA Astrophysics Data System (ADS)

    Cui, Minghuan; Wang, Ji; Wang, Zhiguang; Shen, Tielong; Wei, Kongfang; Yao, Cunfeng; Sun, Jianrong; Gao, Ning; Zhu, Yabin; Pang, Lilong; Wang, Dong; Zhu, Huiping; Han, Yi; Fang, Xuesong

    2017-09-01

    A recently developed reduced activation martensitic MNHS steel was irradiated with 200 keV helium (He) ions to a fluence of 1.0 × 1020 ions/m2 at 300 °C and 1.0 × 1021 ions/m2 at 300 °C and 450 °C. After irradiation, transmission electron microscopy (TEM) and nano-indentation measurements were used to investigate the hardness change and defects induced by He irradiation. Two kinds of defects including He bubbles and dislocation loops are observed by TEM. Irradiation induces hardening of MNHS steels and peak hardness values occur in all irradiated samples. Hardness increments induced by He bubbles and dislocation loops are predicted and fitted with the experimental peak hardness increment, based on the dispersed barrier-hardening (DBH) model and the size and number density of the two defects. A good agreement is got between the predicted and experimental hardness increment and the obstacle strength factor of He bubbles is a little stronger than the obstacle strength of dislocation loops. Other possible contributions to irradiation induced hardening are also discussed.

  8. 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.

  9. 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%.

  10. 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.

  11. 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.

  12. Impact toughness of irradiated reduced-activation ferritic steels*1

    NASA Astrophysics Data System (ADS)

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

    1994-09-01

    Eight chromium-tungsten steels ranging from 2.25 to 12 wt% Cr were irradiated at 365°C to 13-14 dpa in the Fast Flux Test Facility. Post irradiation Charpy impact tests showed a loss of toughness for all steels, as measured by an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy. The most irradiation-resistant steels were two 9% Cr steels: the DBTT of a 9Cr-2W-0.25V-0.1C steel increased 29°C, and for the same composition with an addition of 0.07% Ta the DBTT increased only 15°C. This is the smallest shift ever observed for such a steel irradiated to these levels. The other steels developed shifts in DBTT of 100 to 300°C. A 2.25% Cr steel with 2% W, 0.25% V, and 0.1% C was less severely affected by irradiation than 2.25% Cr steels with 0.25% V and no tungsten, 2% W and no vanadium, and with 1% W and 0.25% V. Irradiation resistance appears to be associated with microstructure, and microstructural manipulation may lead to improved properties.

  13. Response of neutron-irradiated RPV steels to thermal annealing

    SciTech Connect

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

    1997-03-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the fracture toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results of work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response of several irradiated RPV steels.

  14. Irradiation behavior of Ti-stabilized 316L type steel

    NASA Astrophysics Data System (ADS)

    Rodchenkov, B. S.; Kalinin, G. M.; Strebkov, Yu. S.; Shamardin, V. K.; Prokhorov, V. I.; Bulanova, T. M.

    2009-04-01

    Type 316L austenitic steels are widely used for the in-vessel internal structures of fission reactors (core, core support, etc.) and for experimental irradiation facilities. The modifications of 316L Type steel (316L, 316L(N), US 316, J 316, JPCA, etc.) have been considered as structural material for International Thermonuclear Experimental Reactor (ITER). The results of investigation the irradiation behaviour of Ti-stabilized 316 L type steel (0.04 C-15 Cr-11 Ni-2.5 Mo-0.5 Ti) are presented in this work. The specimens cut out from 316L-Ti steel forging were irradiated in the SM-2 reactor up to a dose ˜4 and 10 dpa at 265 ± 15 °C. The tensile properties, fracture toughness and changes in resistance to intergranular stress corrosion cracking (IGSCC) have been investigated after irradiation. The results for Ti-stabilized 316L steel were compared with those for 316L(N)-IG steel irradiated at the same condition.

  15. 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).

  16. Effect of irradiation temperature on microstructural changes in self-ion irradiated austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Jin, Hyung-Ha; Ko, Eunsol; Lim, Sangyeob; Kwon, Junhyun; Shin, Chansun

    2017-09-01

    We investigated the microstructural and hardness changes in austenitic stainless steel after Fe ion irradiation at 400, 300, and 200 °C using transmission electron microscopy (TEM) and nanoindentation. The size of the Frank loops increased and the density decreased with increasing irradiation temperature. Radiation-induced segregation (RIS) was detected across high-angle grain boundaries, and the degree of RIS increases with increasing irradiation temperature. Ni-Si clusters were observed using high-resolution TEM in the sample irradiated at 400 °C. The results of this work are compared with the literature data of self-ion and proton irradiation at comparable temperatures and damage levels on stainless steels with a similar material composition with this study. Despite the differences in dose rate, alloy composition and incident ion energy, the irradiation temperature dependence of RIS and the size and density of radiation defects followed the same trends, and were very comparable in magnitude.

  17. 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.

  18. 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.

  19. 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.

  20. Microchemical evolution of neutron-irradiated stainless steel

    SciTech Connect

    Brager, H.R.; Garner, F.A.

    1980-04-01

    The precipitates that develop in AISI 316 stainless steel during irradiation play a dominant role in determining the dimensional and mechanical property changes of this alloy. This role is expressed primarily in a large change in matrix composition that alters the diffusional properties of the alloy matrix and also appears to alter the rate of acceptance of point defects at dislocations and voids. The major elemental participants in the evolution have been identified as nickel, silicon, and carbon. The exceptional sensitivity of this evolution to many variables accounts for much of the variability of response exhibited by this alloy in nominally similar irradiations.

  1. 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.

  2. 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.

  3. Changes in magnetic properties of neutron irradiated RPV steel

    SciTech Connect

    Park, D.G.; Hong, J.H.; Ok, C.I.; Kim, J.W.; Kim, H.C.

    1998-07-01

    Conventional magnetic parameters and Barkhausen noise have been measured in pressure vessel steel samples both as-received and irradiated with doses of up to 10{sup 18} n/cm{sup 2}. The conventional magnetic parameters, i.e., coercive force, remanence and maximum induction did not change significantly with irradiation, whereas the Barkhausen noise amplitude and energy during a magnetization cycle decreased markedly with irradiation dose. A three stage variation of Barkhausen noise with neutron dose was observed in the present work, namely an initial decrease, a near plateau and rapid decrease. The three stage variation with neutron dose is in qualitative agreement with computer simulations of the radiation damage process performed by Beeler. The hardness also varied in three stages in a reverse manner with transition at the same doses.

  4. Prediction of yield stress in highly irradiated ferritic steels

    NASA Astrophysics Data System (ADS)

    Windsor, Colin G.; Cottrell, Geoff; Kemp, Richard

    2008-03-01

    The design of any fusion power plant requires information on the irradiation hardening of low-activation ferritic/martensitic steels beyond the range of most present measurements. Neural networks have been used by Kemp et al (J. Nucl. Mater. 348 311-28) to model the yield stress of some 1811 irradiated alloys. The same dataset has been used in this study, but has been divided into a training set containing the majority of the dataset with low irradiation levels, and a test set which contains just those alloys which have been irradiated above a given level. For example some 4.5% of the alloys were irradiated above 30 displacements per atom. For this 'prediction' problem it is found that simpler networks with fewer inputs are advantageous. By using target-driven dimensionality reduction, linear combinations of the atomic inputs reduce the test residual below that achievable by adding inputs from single atoms. It is postulated that these combinations represent 'mechanisms' for the prediction of irradiated yield stress.

  5. a Study of Stress Relaxation Rate in Un-Irradiated and Neutron-Irradiated Stainless Steel

    NASA Astrophysics Data System (ADS)

    Ghauri, I. M.; Afzal, Naveed; Zyrek, N. A.

    Stress relaxation rate in un-irradiated and neutron-irradiated 303 stainless steel was investigated at room temperature. The specimens were exposed to 100 mC, Ra-Be neutron source of continuous energy 2-12 MeV for a period ranging from 4 to 16 days. The tensile deformation of the specimens was carried out using a Universal Testing Machine at 300 K. During the deformation, straining was frequently interrupted by arresting the cross head to observe stress relaxation at fixed load. Stress relaxation rate, s, was found to be stress dependent i.e. it increased with increasing stress levels σ0 both in un-irradiated and irradiated specimens, however the rate was lower in irradiated specimens than those of un-irradiated ones. A further decrease in s was observed with increase in exposure time. The experiential decrease in the relaxation rate in irradiated specimens is ascribed to strong interaction of glide dislocations with radiation induced defects. The activation energy for the movement of dislocations was found to be higher in irradiated specimens as compared with the un-irradiated ones.

  6. 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.

  7. Microstructural study of irradiated isotopically tailored F82H steel

    NASA Astrophysics Data System (ADS)

    Wakai, E.; Miwa, Y.; Hashimoto, N.; Robertson, J. P.; Klueh, R. L.; Shiba, K.; Abiko, K.; Furuno, S.; Jitsukawa, S.

    2002-12-01

    The synergistic effect of displacement damage and hydrogen or helium atoms on microstructures in F82H steel irradiated at 250-400 °C to 2.8-51 dpa in HFIR has been examined using isotopes of 54Fe or 10B. Hydrogen atoms increased slightly the formation of dislocation loops and changed the Burgers vector for some parts of dislocation loops, and they also affected on the formation of cavity at 250 °C to 2.8 dpa. Helium atoms also influenced them at around 300 °C, and the effect of helium atoms was enhanced at 400 °C. Furthermore, the relations between microstructures and radiation-hardening or ductile to brittle transition temperature (DBTT) shift in F82H steel were discussed. The cause of the shift increase of DBTT is thought to be due to the hardening of dislocation loops and the formation of α '-precipitates on dislocation loops.

  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. 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.

  10. Stress corrosion cracking of neutron irradiated type 304 stainless steels

    SciTech Connect

    Tsukada, Takashi; Miwa, Yukio; Nakajima, Hajime

    1995-12-31

    To study the effect of minor elements on the irradiation assisted stress corrosion cracking (IASCC), a high purity type 304L stainless steel and its heats doped minor elements, Si, P, S, C and Ti were irradiated at 513 K to 6.7 {times} 10{sup 24} n/m{sup 2} (E>1 MeV). After irradiation, susceptibility to the stress corrosion cracking (SCC) was evaluated by the slow strain rate tensile (SSRT) test in an oxygenated high purity water at 573 K, and the fracture surface of the specimens was examined by the scanning electron microscopy (SEM). The specimens showed high susceptibilities to SCC. Specimens without addition of C showed the intergranular type SCC (IGSCC), while C doped specimens generally failed by the transgranular type SCC (TGSCC). Addition of C into the hi purity alloy caused an enhancement of radiation hardening and a remarkable increase in maximum stress during SSRT test. Enrichment of Si changed specifically tensile properties after irradiation and decreased maximum stress and improved total elongation. Addition of S greatly enhanced the IASCC susceptibility and addition of P seemed to be beneficial for suppressing it. An effect of Ti was not prominent in the alloy with a high C concentration.

  11. Microchemical characterization of grain boundaries in irradiated steels

    SciTech Connect

    Walmsley, J.; Spellward, P.; Fisher, S.; Jenssen, A.

    1995-12-31

    Field Emission Gun Scanning Transmission Electron Microscopy and Auger Electron Spectroscopy have been used to characterize grain boundaries in unirradiated and neutron-irradiated type 304 stainless steel. Both techniques are used to give compositional information with nanometer-scale spatial resolution at and around grain boundaries. Irradiation induced changes in grain boundary nanochemistry from the solution treated starting condition are described. Initial segregation of Cr at boundaries is seen to develop through an intermediate ``side-lobe`` distribution, seen clearly at an intermediate dose of {approximately}10{sup 21}n/cm{sup 2}, to Cr depletion at higher dose of {approximately} 10{sup 22}n/cm{sup 2}. Thin foil analysis suggests a considerably higher grain boundary phosphorus level in the intermediate dose material than is measured by fracture surface analysis. For the high dose material the two techniques produce consistent phosphorus levels when comparison is made using experience gained from dual examinations of other steels. It is suggested that in the medium dose material fracture occurs along the plane of minimum chromium arising from the ``side-lobe`` Cr distribution so that the surface exposed by fracture is several nanometers away from the true grain boundary.

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

    NASA Astrophysics Data System (ADS)

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

    2006-10-01

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

  13. 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.

  14. 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.

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

    NASA Astrophysics Data System (ADS)

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

    1994-09-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-12-01

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

  17. 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.

  18. Stress corrosion cracking susceptibility of irradiated Type 304 stainless steels

    SciTech Connect

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

    1992-08-01

    Slow-strain-rate tensile tests and microstructural analysis by Auger electron spectroscopy were conducted on specimens of high- and commercial-purity (HP and CP) heats of Type 304 stainless steel obtained from neutron absorber tubes and a control blade sheath after irradiation up to 2.5 [times] 10[sup 2l] n[center dot]cm[sup [minus]2] (E > l MeV) in boiling water reactors (BWRs). The susceptibility of the HP absorber tubes to intergranular stress corrosion cracking (IGSCC) was higher than that of the CP absorber tubes or the CP control blade sheath. IGSCC susceptibilities of the BWR components could not be correlated to segregation impurities on grain boundaries. However for comparable fluence levels, Cr on grain-boundaries.

  19. Stress corrosion cracking susceptibility of irradiated Type 304 stainless steels

    SciTech Connect

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

    1992-08-01

    Slow-strain-rate tensile tests and microstructural analysis by Auger electron spectroscopy were conducted on specimens of high- and commercial-purity (HP and CP) heats of Type 304 stainless steel obtained from neutron absorber tubes and a control blade sheath after irradiation up to 2.5 {times} 10{sup 2l} n{center_dot}cm{sup {minus}2} (E > l MeV) in boiling water reactors (BWRs). The susceptibility of the HP absorber tubes to intergranular stress corrosion cracking (IGSCC) was higher than that of the CP absorber tubes or the CP control blade sheath. IGSCC susceptibilities of the BWR components could not be correlated to segregation impurities on grain boundaries. However for comparable fluence levels, Cr on grain-boundaries.

  20. Impact behavior of 9-Cr and 12-Cr ferritic steels after low-temperature irradiation

    SciTech Connect

    Klueh, R.L.; Vitek, J.M.; Corwin, W.R.; Alexander, D.J.

    1987-01-01

    Miniature Charpy impact specimens of 9Cr-1MoVNb and 12Cr-1MoVW steels and these steels with 1 and 2% Ni were irradiated in the High-Flux Isotope Reactor (HFIR) at 50/sup 0/C to displacement damage levels of up to 9 dpa. Nickel was added to study the effect of transmutation helium. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT). The 9Cr-1MoVNb steels, with and without nickel, showed a larger shift than the 12Cr-1MoVW steels, with and without nickel. The results indicated that helium also increased the DBTT. The same steels were previously irradiated at higher temperatures. From the present and past tests, the effect of irradiation temperature on the DBTT behavior can be evaluated. For the 9Cr-1MoVNb steel, there is a continuous decrease in the magnitude of the DBTT increase up to an irradiation temperature of about 400/sup 0/C, after which the shift drops rapidly to zero at about 450/sup 0/C. The DBTT of the 12Cr-1MoVW steel shows a maximum increase at an irradiation temperature of about 400/sup 0/C and less of an increase at either higher or lower irradiation temperatures.

  1. Influence of laser shock peening on irradiation defects in austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Lu, Qiaofeng; Su, Qing; Wang, Fei; Zhang, Chenfei; Lu, Yongfeng; Nastasi, Michael; Cui, Bai

    2017-06-01

    The laser shock peening process can generate a dislocation network, stacking faults, and deformation twins in the near surface of austenitic stainless steels by the interaction of laser-driven shock waves with metals. In-situ transmission electron microscopy (TEM) irradiation studies suggest that these dislocations and incoherent twin boundaries can serve as effective sinks for the annihilation of irradiation defects. As a result, the irradiation resistance is improved as the density of irradiation defects in laser-peened stainless steels is much lower than that in untreated steels. After heating to 300 °C, a portion of the dislocations and stacking faults are annealed out while the deformation twins remain stable, which still provides improved irradiation resistance. These findings have important implications on the role of laser shock peening on the lifetime extension of austenitic stainless steel components in nuclear reactor environments.

  2. Reduction method of DBTT shift due to irradiation for reduced-activation ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Wakai, E.; Okubo, N.; Ando, M.; Yamamoto, T.; Takada, F.

    2010-03-01

    The method for reducing irradiation-induced DBTT shift of reduced-activation ferritic/martensitic steels was examined. F82H-LN (low nitrogen, 20 ppm), F82H+60 ppm 11B+200 ppmN and F82H+60 ppm 10B+200 ppmN steels tempered at 780 °C for 0.5 h were irradiated at 250 °C to 2 dpa, and the results for Charpy impact tests were analyzed. The upper shelf energy of F82H+ 11B+N steel was hardly changed by the irradiation, and DBTT shift was very small. From our research, DBTT shift due to irradiation can be reduced by the control of tempered conditions before irradiation, and it is found to be furthermore reduced by impurity doping with 60 ppm 11B and 200 ppmN to F82H steel.

  3. Deformation and cracking of irradiated austenitic stainless steels

    SciTech Connect

    Carter, R.D.; Atzmon, M.; Was, G.S.

    1995-12-31

    Samples of proton-irradiated 304L stainless steel were deformed by constant extension rate tensile tests at strain rates of 3 {times} 10{sup {minus}7} s{sup {minus}1} and 3 {times} 10{sup {minus}8} s{sup {minus}1} to strains of up to 10% at 288--350 C in argon. Minor cracking was observed in and around spinel inclusions in the material, however no intergranular cracking of the type observed in water environments was found. Thus intergranular cracking cannot occur by a radiation hardening mechanism alone. The microstructures that resulted from irradiation and deformation were characterized using electron microscopy. Surface slip band formation is observed on one or two {l_brace}111{r_brace} slip systems in each grain. The slip bands correspond to dislocation channels in the material as identified by transmission electron microscopy. The channels form by activation of grain-boundary dislocation sources, with the emitted dislocations sweeping through the grain interior to the opposing rain boundaries. During this process, the dislocations remove the radiation-produced defects. Slip band and dislocation channel densities increase with increasing strain in the samples. These results are used to interpret stress corrosion cracking behavior in this material.

  4. 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.

  5. 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.

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

    NASA Astrophysics Data System (ADS)

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

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

  7. 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.

  8. Irradiation effects on magnetic properties in neutron and proton irradiated reactor pressure vessel steel

    SciTech Connect

    Park, D.G.; Hong, J.H.; Kim, I.S.; Kim, H.C.

    1999-09-01

    The effects of neutron and proton dose on the magnetic properties of a reactor pressure vessel (RPV) steel were investigated. The coercivity and maximum induction increased in two stages with respect to neutron dose, being nearly constant up to a dose of 1.5 x 10{sup {minus}7} dpa, followed by a rapid increase up to a dose of 1.5 x 10{sup {minus}5} dpa. The coercivity and maximum induction in the proton irradiated specimens also showed a two stage variation with respect to proton dose, namely a rapid increase up to a dose of 0.2 x 10{sup {minus}2} dpa, then a decrease up to 1.2 x 10{sup {minus}2} dpa. The Barkhausen noise (BN) amplitude in neutron irradiated specimens also varied in two stages in a reverse manner, the transition at the same dose of 1.5 x 10{sup {minus}7} dpa. The BN amplitude in proton irradiated specimens decreased by 60% up to 0.2 x 10{sup {minus}2} dpa followed by an increase up to 1.2 x 10{sup {minus}2} dpa. The results were in good accord with the one dimensional domain wall model considering the density of defects and wall energy.

  9. Mechanical properties and microstructure of advanced ferritic-martensitic steels used under high dose neutron irradiation

    NASA Astrophysics Data System (ADS)

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

    Some results of the study of mechanical properties and structure of ferritic-martensitic chromium steels with 13% and 9% chromium, irradiated in the BOR-60 reactor up to different damage doses are presented in this report. Results concerning the behaviour of commercial steels, containing to molybdenum, vanadium and niobium, and developed for the use in fusion reactors, are compared to low-activation steels in which W and Ta replaced Mo and Nb. It is shown that after irradiation to the dose of ˜10 dpa at 400°C 0.1C-9Cr-1W, V, Ta steels are prone to lower embrittlement as deduced from fracture surface observations of tensile specimens. Peculiarities of fine structure and fracture mode, composition and precipitation reactions in steels during irradiation are discussed.

  10. Impact behavior of reduced-activation steels irradiated to 24 dpa*1

    NASA Astrophysics Data System (ADS)

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

    1996-10-01

    Charpy impact tests were previously conducted on eight chromium-tungsten steels after irradiation at 365°C to 6-8 and 15-17 dpa in the Fast Flux Test Facility. These same steels, which range in concentration from 2.25 to 12 wt% (all steels contained 0.1%C), have now been irradiated to 20-24 dpa under the same conditions. Post-irradiation Charpy impact tests after 20-24 dpa showed that the loss of impact toughness, as measured by an increase in the ductile—brittle transition temperature (DBTT) and a decrease in the upper-shelf energy, remained relatively unchanged from the values after 15-17 dpa. As before, the most irradiation-resistant steels were two 9% Cr steels: the DBTT of a 9Cr2W0.25V steel increased 59°C, and for the same composition with an addition of 0.07% Ta, the DBTT increased only 21°C. The other steels developed shifts in DBTT of 100 to 300°C. A 2.25% Cr steel with 2% W and 0.25% V was less severely affected by irradiation than 2.25% Cr steels with 0.25% V and no tungsten, 2% W and no vanadium, and with 1% W and 0.25% V. Steels with 5 and 12% Cr, 2% W, and 0.25% V had properties between those of the 2.25Cr and 9Cr steels.

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

    SciTech Connect

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

    1997-06-01

    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 will be 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. In addition to irradiation hardening, neutrons from the fusion reaction will produce large amounts of helium in the steels used to construct fusion power plant components. Tests to simulate the fusion environment indicate that helium can also affect the toughness. Steels are being developed for fusion applications that have a low DBTT prior to irradiation and then show only a small shift after irradiation. A martensitic 9Cr-2WVTa (nominally Fe-9Cr-2W-0.25V-0.07Ta-0.1C) steel had a much lower DBTT than the conventional 9Cr-1MoVNb steel prior to neutron irradiation and showed a much smaller increase in DBTT after irradiation. 27 refs., 5 figs., 1 tab.

  12. Mechanical properties and TEM examination of RAFM steels irradiated up to 70 dpa in BOR-60

    NASA Astrophysics Data System (ADS)

    Gaganidze, E.; Petersen, C.; Materna-Morris, E.; Dethloff, C.; Weiß, O. J.; Aktaa, J.; Povstyanko, A.; Fedoseev, A.; Makarov, O.; Prokhorov, V.

    2011-10-01

    Mechanical properties of Reduced Activation Ferritic/Martensitic (RAFM) steels were studied after irradiation in BOR-60 reactor to a neutron displacement damage of 70 dpa at 330-340 °C. Yield stress and Ductile-to-Brittle-Transition-Temperature of EUROFER97 indicate saturation of hardening and embrittlement. The phenomenological models for description of microstructure evolution and resulting irradiation hardening and embrittlement are discussed. The evolution of yield stress with dose is qualitatively understood within a Whapham and Makin model. Dislocation loops examined in TEM are considered a main source for low-temperature irradiation hardening. The analysis of the fatigue data in terms of the inelastic strain reveals comparable fatigue behaviour both for unirradiated and irradiated conditions, which can be described by a common Manson-Coffin relation. The study of helium effects in B-doped model steels indicated progressive material embrittlement with helium content. Post-irradiation annealing of RAFM steels yielded substantial recovery of mechanical properties.

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

    SciTech Connect

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

    1984-01-01

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

  14. Effect of Initial Heat Treatment on DBTT of F82H Steel Irradiated by Neutrons

    SciTech Connect

    Wakai, E.; Ando, M.; Matsukawa, S.; Taguchi, T.; Yamamoto, T.; Tomita, H.; Takada, F.

    2005-05-15

    The dependence of ductile-brittle transition temperature (DBTT) on tempering time and temperature was examined for a martensitic steel F82H irradiated at 150 and 250 deg. C to a neutron dose of 1.9 dpa in the JMTR. The heat treatment was performed at 750 and 780 deg. C for 0.5 h after the normalizing at 1040 deg. C for 0.5 h. The tempering time at 750 deg. C was varied from 0.5 to 10 h. 1/3CVN specimens were used in this study, and the absorbed energies in the impact tests were measured as a function of temperature. DBTT of F82H steels irradiated at 250 deg. C to 1.9 dpa was ranged from -23 to 25 deg. C, and DBTT of F82H steels irradiated at 150 deg. C to 1.9 dpa was ranged from 0 to 15 deg. C. DBTT of F82H steels irradiated at 250 deg. C depended strongly on temperature and time of tempering, and it tended to decrease with increasing yield stress. The effect of tempering conditions on DBTT was smaller in the specimens irradiated at 150 deg. C. DBTT due to irradiation in the F82H steels irradiated at 250 deg. C tended to decrease with increasing time and temperature of tempering.

  15. 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 8 reduced-activation Cr-W steels after irradiation to 15-17 and 26-29 dpa, and Charpy impact tests were conducted on steels irradiated to 26-29 dpa. Irradiation was in Fast Flux Test Facility at 365 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 20,000 h at 365 C. Thermal aging had little effect on tensile properties or ductile-brittle transition temperature (DBTT), but several steels showed a slight increase in upper-shelf energy (USE). After 7 dpa, strength increased (hardened) and then remained relatively unchanged through 26-29 dpa (ie, strength saturated with fluence). Post-irradiation Charpy impact tests after 26-29 dpa showed that the loss of impact toughness (increased DBTT, decreased 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 had the most irradiation resistance.

  16. 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.

  17. Fracture toughness of irradiated wrought and cast austenitic stainless steels in BWR environment.

    SciTech Connect

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

    2007-01-01

    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. 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). We look at the results of a study of simulated light-water reactor coolants, material chemistry, and irradiation damage and their effects on the susceptibility to stress-corrosion cracking of various commercially available and laboratory-melted stainless steels.

  18. 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.

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

    SciTech Connect

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

    2008-01-01

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

  20. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2006-03-01

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

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

    SciTech Connect

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

    2006-03-06

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

  3. Ar ion irradiation hardening of high-Cr ferritic/martensitic steels at 700 °C

    NASA Astrophysics Data System (ADS)

    Shen, Yinzhong; Zhu, Jun; Huang, Xi

    2016-03-01

    High-Cr ferritic/martensitic (FM) steels are being considered for applications as fuel cladding or core structures for Generation-IV reactors. Because high temperatures approaching 923-973 K (650-700 °C) are envisioned in the designs of Generation IV reactors, irradiation response of high-Cr FM steels at the high temperatures requires investigations. Response of two high-Cr FM steels P92 and 11Cr to irradiation at 973 K (700 °C) was investigated through Ar ion irradiation in combination with damage simulations, nanoindentation measurements and microstructure analyses. Irradiation hardening occurred in both steels after Ar ion irradiation at 973 K (700 °C) to 10 dpa, providing the first evidence that irradiation hardening can occur at a high irradiation temperature of 973 K (700 °C) in high-Cr FM steels. Argon bubbles with a very high number density and an average diameter of about 2.6-3 nm formed in the two steels after the irradiation. The irradiation hardening occurred in the two steels is attributed to the formation of these high-number-density fine argon bubbles produced by the irradiation homogeneously distributed in the matrix. Difference in the magnitude of irradiation hardening between the two steels was also discussed.

  4. 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.

  5. Temperature and strain-rate effects on deformation mechanisms in irradiated stainless steel

    SciTech Connect

    Brimhall, J.L.; Cole, J.I.; Vetrano, J.S.; Bruemmer, S.M.

    1994-11-01

    Analysis of the deformation microstructures in ion-irradiated stainless steel shows twinning to be the predominant deformation mode at room temperature. Dislocation channelling also occurs under slow strain rate conditions. Stresses required for twinning were calculated by the model of Venables and are compatible with observed yield stresses in neutron-irradiated material if loops are the principal twin source. Computation of the expected radiation hardening from the defect structure, based on a simple model, is consistent with yield strengths measured on neutron-irradiated steels. Lower yield stresses and greater thermal energy at 288 C lessen the probability of twinning and dislocation channeling becomes the primary deformation mode at the higher temperature. However, preliminary early results show that some twinning does occur in the irradiated stainless steel even at the higher temperature when higher strain rates are used.

  6. 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.

  7. 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

  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. 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).

  10. Neutron-irradiated model alloys and pressure-vessel steels studied using positron spectroscopy

    NASA Astrophysics Data System (ADS)

    Cumblidge, Stephen Eric

    We have used positron-annihilation-lifetime spectroscopies to examine microstructural evolution of pressure vessel steels and model alloys that have systematically varied amounts of copper, nickel, and phosphorus during neutron irradiation and post-irradiation annealing. The objective of this work was to characterize the neutron-irradiation induced microstructural features that cause the embrittlement of nuclear reactor pressure-vessel steel. We used positron annihilation lifetime spectroscopy and Doppler-broadening spectroscopy to examine the model alloys and pressure-vessel steels before and after irradiation and after post-irradiation annealing. We followed the changes in the mechanical properties of the materials using Rockwell 15N hardness measurements. The results show that in both the model alloys and pressure-vessel steels neutron irradiation causes the formation of vacancy-type defect clusters and a fine distribution of copper- and nickel-enriched metallic precipitates. The vacancy clusters are small in size and were present in all samples, and disappear upon annealing at 450°C. The metallic precipitates are present only in the model alloy samples with either high Cu or a combination of medium Cu and high Ni, and they remain in the microstructure after annealing up to 550°C, starting to anneal possibly at 600°C. The neutron-irradiated pressure vessel steels behave similarly to the high Cu samples, indicating that neutron irradiation induced precipitation occurs in these alloys as well. This work provides independent evidence for the irradiation-induced metallic precipitates seen by other techniques, gives evidence for the exact nature of the matrix damage, and is significant to understanding the in-service degradation of pressure vessel materials.

  11. Embrittlement of CrMo steels after low fluence irradiation in HFIR

    NASA Astrophysics Data System (ADS)

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

    1995-02-01

    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°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 toughness. Displacement damage was produced by fast neutrons, and helium was formed by the reaction of 58Ni with thermal neutrons in the mixed-neutron spectrum of HFIR. 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°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 behavior 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.

  12. 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.

  13. 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).

  14. 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.

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

    SciTech Connect

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

    1986-01-01

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

  16. Effects of hydrogen on mechanical properties in irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Morisawa, J.; Kodama, M.; Nishimura, S.; Asano, K.; Nakata, K.; Shima, S.

    1994-09-01

    To investigate the hydrogen effect on mechanical properties of solution annealed Type 304 stainless steel, tensile tests of neutron irradiated materials were conducted after a hydrogen charging and discharging process (hydrogen treatment). Elongation was less with increasing neutron fluence after hydrogen treatment than that of as-irradiated specimens. Intergranular cracking occurred by the hydrogen treatment in heavier irradiated specimens, in which the Cr depleted zone along grain boundary was observed. Embrittlement and intergranular cracking after the hydrogen treatment were estimated to be attributed to the Cr depleted zone at the grain boundary due to neutron irradiation.

  17. 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.

  18. 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.

  19. 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).

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

    SciTech Connect

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

    2012-12-08

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

  1. 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.

  2. Microstructure property analysis of HFIR-irradiated reduced-activation ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Tanigawa, H.; Hashimoto, N.; Sakasegawa, H.; Klueh, R. L.; Sokolov, M. A.; Shiba, K.; Jitsukawa, S.; Kohyama, A.

    2004-08-01

    The effects of irradiation on the Charpy impact properties of reduced-activation ferritic/martensitic steels were investigated on a microstructural basis. It was previously reported that the ductile-brittle transition temperature (DBTT) of F82H-IEA and its heat treatment variant increased by about 130 K after irradiation at 573 K up to 5 dpa. Moreover, the shifts in ORNL9Cr-2WVTa and JLF-1 steels were much smaller, and the differences could not be interpreted as an effect of irradiation hardening. The precipitation behavior of the irradiated steels was examined by weight analysis and X-ray diffraction analysis on extraction residues, and SEM/EDS analysis was performed on extraction replica samples and fracture surfaces. These analyses suggested that the difference in the extent of DBTT shift could be explained by (1) smaller irradiation hardening at low test temperatures caused by irradiation-induced lath structure recovery (in JLF-1), and (2) the fracture stress increase caused by the irradiation-induced over-solution of Ta (in ORNL9Cr-2WVTa).

  3. 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.

  4. 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.

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

    NASA Astrophysics Data System (ADS)

    Dai, Yong; Marmy, Pierre

    2005-08-01

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

  6. 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.

  7. Thermally activated deformation of irradiated reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Böhmert, J.; Müller, G.

    2002-03-01

    Temperature and strain rate change tensile tests were performed on two VVER 1000-type reactor pressure vessel welds with different contents of nickel in unirradiated and irradiated conditions in order to determine the activation parameters of the contribution of the thermally activated deformation. There are no differences of the activation parameters in the unirradiated and the irradiated conditions as well as for the two different materials. This shows that irradiation hardening preferentially results from a friction hardening mechanism by long-range obstacles.

  8. Helium effects on the mechanical properties of neutron-irradiated Cr-Mo ferritic steels

    SciTech Connect

    Klueh, R.L.

    1990-01-01

    In the first wall of a fusion rector, large amounts of transmutation helium will be produced simultaneously with the displacement damage caused by high-energy neutrons from the fusion reaction. One method used to simulate irradiation effects for ferritic steels is to add nickel to the steels and irradiate them in a mixed-spectrum reactor. Fast neutrons in the spectrum produce displacement damage, while transmutation 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 tensile properties and toughness. Results from these studies are summarized.

  9. 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.

  10. Microstructure evolution and degradation mechanisms of reactor internal steel irradiated with heavy ions

    NASA Astrophysics Data System (ADS)

    Borodin, O. V.; Bryk, V. V.; Kalchenko, A. S.; Parkhomenko, A. A.; Shilyaev, B. A.; Tolstolutskaya, G. D.; Voyevodin, V. N.

    2009-03-01

    Structure evolution and degradation mechanisms during irradiation of 18Cr-10Ni-Ti steel (material of VVER-1000 reactor internals are investigated). Using accelerator irradiations with Cr3+ and Ar+ ions allowed studying effects of dose rate, different initial structure state and implanted ions on features of structure evolution and main mechanisms of degradation including low temperature swelling and embrittlement of the 18Cr-10Ni-Ti steel. It is shown that differences in dose rate at most irradiation temperatures mainly exert their influence on the duration of the swelling transient regime. Calculations of possible transmutation products during irradiation of this steel in a VVER-1000 spectrum were performed. It is shown that gaseous atoms (He and H), which are generated simultaneously with radiation defects, stabilize the elements of radiation microstructure and influence the swelling. The nature of deformation under different temperatures of irradiation and of mechanical testing is investigated. It is shown that the temperature sensitivity of swelling behaviour in the investigated steel, with different initial structures can be connected with the dynamic behaviour of point defect sinks.

  11. 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.

  12. Positron annihilation study of proton-irradiated reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Liu, Xiangbing; Wang, Rongshan; Ren, Ai; Huang, Ping; Wu, Yichu; Jiang, Jing; Zhang, Chonghong; Wang, Xitao

    2012-10-01

    The microstructures, irradiation-induced defects and changes of mechanical property of Chinese domestic A508-3 steels after proton irradiation were investigated by TEM, positron lifetime, slow positron beam Doppler broadening spectroscopy and hardness measurements. The defects were induced by 240 keV proton irradiation with fluences of 1.25×1017 ions cm-2 (0.26 dpa), 2.5×1017 ions cm-2 (0.5 dpa), and 5.0×1017 ions cm-2 (1.0 dpa). The TEM observation revealed that the as-received steel had typical bainitic-ferritic microstructures. It was also observed that Doppler broadening S-parameter and average lifetime increased with dose level owing to the formation of defects and voids induced by proton irradiation. The correlation between positron parameters and hardness was found.

  13. Effect of irradiation on the steels 316L/LN type to 12 dpa at 400 °C

    NASA Astrophysics Data System (ADS)

    Bulanova, T.; Fedoseev, A.; Kalinin, G.; Rodchenkov, B.; Shamardin, V.

    2004-08-01

    The 316L type stainless steel is widely used as a structural material for the fission reactors internal structures (core, core supports, etc.) and for experimental irradiation facilities. The 316L(N)-IG type steel is proposed as a main structural material for the ITER reactor (first wall, blanket, vacuum vessel, cooling pipe lines). It is obvious that different steel grades should exhibit different reaction to neutron irradiation. The main objective of this work was to study of irradiation behaviour of three different commercial steels: AISI 316LN, AISI 316L (US grades) and 02X17H14M2 (Russian steel grade that is similar to 316L). Irradiation effect on the three commercial steels of 316L family to ˜12 dpa at the temperature ˜370-400 °C on the tensile properties, microstructure, swelling and susceptibility to SCC are described in the paper.

  14. Neutron Irradiation Effects on Fatigue Crack Propagation in Type 316 Stainless Steels at 649 C.

    DTIC Science & Technology

    1980-08-01

    maintaining the maximum tensile load constant for selected time periods duriig each cycle. Induction heating was employed to achieve a test temperature of...7 AD-A OB 052 NAVAL RESEAR ICH LAS WASHINGTON DC F/6 11/BNEUTRON IRRADIATION EFFECTS ON FATIGUE CRACK PROPAGATION IN TYP-EC(UlU UNL AUG 80 0 .J...and Identify by block number) Radiation Microstruture B Irradiation Fatigue Stainless steels Crack propagation Radiation effects High temperature V

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

    SciTech Connect

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

    1980-01-01

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

  16. 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.

  17. 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.

  18. 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.

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

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale J.; Was, Gary S.

    2014-01-01

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

  20. 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.

  1. 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.

  2. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

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

    DOE PAGES

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; ...

    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

  5. Impact of the nanostructuration on the corrosion resistance and hardness of irradiated 316 austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Hug, E.; Prasath Babu, R.; Monnet, I.; Etienne, A.; Moisy, F.; Pralong, V.; Enikeev, N.; Abramova, M.; Sauvage, X.; Radiguet, B.

    2017-01-01

    The influence of grain size and irradiation defects on the mechanical behavior and the corrosion resistance of a 316 stainless steel have been investigated. Nanostructured samples were obtained by severe plastic deformation using high pressure torsion. Both coarse grain and nanostructured samples were irradiated with 10 MeV 56Fe5+ ions. Microstructures were characterized using transmission electron microscopy and atom probe tomography. Surface mechanical properties were evaluated thanks to hardness measurements and the corrosion resistance was studied in chloride environment. Nanostructuration by high pressure torsion followed by annealing leads to enrichment in chromium at grain boundaries. However, irradiation of nanostructured samples implies a chromium depletion of the same order than depicted in coarse grain specimens but without metallurgical damage like segregated dislocation loops or clusters. Potentiodynamic polarization tests highlight a definitive deterioration of the corrosion resistance of coarse grain steel with irradiation. Downsizing the grain to a few hundred of nanometers enhances the corrosion resistance of irradiated samples, despite the fact that the hardness of nanocrystalline austenitic steel is only weakly affected by irradiation. These new experimental results are discussed in the basis of couplings between mechanical and electrical properties of the passivated layer thanks to impedance spectroscopy measurements, hardness properties of the surfaces and local microstructure evolutions.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Low-temperature irradiation effects on tensile and Charpy properties of low-activation ferritic steels

    NASA Astrophysics Data System (ADS)

    Shiba, Kiyoyuki; Hishinuma, Akimichi

    2000-12-01

    Tensile and Charpy properties of low-activation ferritic steel, F82H irradiated up to 0.8 dpa at low temperature below 300°C were investigated. The helium effect on these properties was also investigated using the boron isotope doping method. Neutron irradiation increased yield stress accompanied with ductility loss, and it also shifted the ductile-to-brittle transition temperature (DBTT) from -50°C to 0°C. Boron-doped F82H showed larger degradation in DBTT and ductility than boron-free F82H, while they had the same yield stress before and after irradiation.

  11. 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.

  12. 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.

  13. Effect of nickel content on the neutron irradiation embrittlement of Ni-Mo-Cr steels

    NASA Astrophysics Data System (ADS)

    Lee, Chang-Hoon; Kasada, R.; Kimura, A.; Lee, Bong-Sang; Suh, Dong-Woo; Lee, Hu-Chul

    2013-11-01

    The influence of nickel on the neutron irradiation embrittlement of Ni-Mo-Cr reactor pressure vessel (RPV) steels was investigated using alloys containing nickel in the range of 0.9-3.5 wt%. In all investigated alloys, the neutron irradiation with two dose conditions of 4.5 × 1019 neutron/cm2 at 290 °C and 9.0 × 1019 neutron/cm2 at 290 °C, respectively, increased the hardness and ductile-to-brittle transition temperature (DBTT). However, the increases of the hardness and DBTT resulting from the neutron irradiation were primarily affected by the irradiation dose that is closely related to the generation of irradiation defects, but not by the nickel content. In addition, a linear relationship between the changes in the hardness and DBTT subjected to the irradiation was confirmed. These results demonstrate that increasing the nickel content up to 3.5 wt% does not have a harmful effect on the irradiation embrittlement of Ni-Mo-Cr reactor pressure vessel (RPV) steels.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Exploratory Study of Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    SciTech Connect

    Chernobaeva, A.A., Kryukov, A.M., Nikolaev, Y.A., Korolev, Y.N. , Sokolov, M.A., Nanstad, R.K.

    1997-12-31

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVS) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The working group agreed that each side would irradiate, anneal, reirradiate (if feasible), and test two materials of the other; so far, only charpy impact and tensile specimens have been included. Oak Ridge National Laboratory (ornl) conducted such a program (irradiation and annealing) with two weld metals representative of VVER-440 AND VVER-1000 RPVS, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation,annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) program plate 02 and Heavy-Section Steel Irradiation (HSSI) program weld 73w. The results for each material from each laboratory are compared with those from the other laboratory. the ORNL experiments with the VVER welds included irradiation to about 1 x 10 (exp 19) N/SQ CM ({gt}1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 X 10 (exp 19) N/SQ CM ({gt}1 MeV).

  19. Status Summary of FY16 Atom Probe Tomography Studies on UCSB ATR-2 Irradiated RPV Steels

    SciTech Connect

    Wells, Peter; Odette, G. Robert

    2016-05-01

    The University of California Santa Barbara-2 RPV Steel Irradiation experiment was awarded in 2010 by the Nuclear Science User Facility (formerly ATR NSUF) through a competitive peer review proposal process. The experiment involved irradiation of nearly 1300 samples distributed over 13 capsules. The major objective of this experiment was to better understand embrittlement behavior of reactor pressure steels at doses beyond which available data exists yet may be achieved if reactor operating licenses are extended beyond 60 years. The experiment was instrumented during irradiation and active temperature control was used to maintain the temperature at the design temperature. Six samples were selected from a large matrix of materials to perform atom probe tomography (APT) to look at formation of high dose phases. The nature and formation behavior of these phases is discussed.

  20. Irradiation-induced embrittlement of a 2.25Cr1Mo steel

    NASA Astrophysics Data System (ADS)

    Song, S.-H.; Faulkner, R. G.; Flewitt, P. E. J.; Smith, R. F.; Marmy, P.; Victoria, M.

    2000-07-01

    Irradiation-induced embrittlement of a 2.25Cr1Mo is investigated by means of small punch testing and scanning electron microscopy (SEM). The ductile-brittle transition temperature (DBTT) determined by the small punch test is much lower than that determined by the standard Charpy test. There are some irradiation-induced embrittlement effects after the steel is irradiated at about 270°C for 46 days with a neutron dose rate of 1.05×10 -8 dpa s -1 and at about 400°C for 86 days with a neutron dose rate of 1.75×10 -8 dpa s -1. In addition, there is some temper embrittlement after the steel is aged at about 400°C for 86 days.

  1. Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    SciTech Connect

    Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

    1998-06-16

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10{sup 19} n/cm{sup 2} (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10{sup 19} n/cm{sup 2} (>l MeV). In both cases, irradiations were conducted at {approximately}290 C and annealing treatments were conducted

  2. Hardening of ODS ferritic steels under irradiation with high-energy heavy ions

    NASA Astrophysics Data System (ADS)

    Ding, Z. N.; Zhang, C. H.; Yang, Y. T.; Song, Y.; Kimura, A.; Jang, J.

    2017-09-01

    Influence of the nanoscale oxide particles on mechanical properties and irradiation resistance of oxide-dispersion-strengthened (ODS) ferritic steels is of critical importance for the use of the material in fuel cladding or blanket components in advanced nuclear reactors. In the present work, impact of structures of oxide dispersoids on the irradiation hardening of ODS ferritic steels was studied. Specimens of three high-Cr ODS ferritic steels containing oxide dispersoids with different number density and average size were irradiated with high-energy Ni ions at about -50 °C. The energy of the incident Ni ions was varied from 12.73 MeV to 357.86 MeV by using an energy degrader at the terminal so that a plateau of atomic displacement damage (∼0.8 dpa) was produced from the near surface to a depth of 24 μm in the specimens. A nanoindentor (in constant stiffness mode with a diamond Berkovich indenter) and a Vickers micro-hardness tester were used to measure the hardeness of the specimens. The Nix-Gao model taking account of the indentation size effect (ISE) was used to fit the hardness data. It is observed that the soft substrate effect (SSE) can be diminished substantially in the irradiated specimens due to the thick damaged regions produced by the Ni ions. A linear correlation between the nano-hardeness and the micro-hardness was found. It is observed that a higher number density of oxide dispersoids with a smaller average diameter corresponds to an increased resistance to irradiation hardening, which can be ascribed to the increased sink strength of oxides/matrix interfaces to point defects. The rate equation approach and the conventional hardening model were used to analyze the influence of defect clusters on irradiation hardening in ODS ferritic steels. The numerical estimates show that the hardening caused by the interstitial type dislocation loops follows a similar trend with the experiment data.

  3. 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

  4. 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.

  5. 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.

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

    SciTech Connect

    Corwin, W.R.

    1995-02-01

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

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

    SciTech Connect

    Rosseel, T.M.

    1998-02-01

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

  8. Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation

    NASA Astrophysics Data System (ADS)

    Lindgren, Kristina; Boåsen, Magnus; Stiller, Krystyna; Efsing, Pål; Thuvander, Mattias

    2017-05-01

    Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation results in nanometre cluster formation, which in turn affects the mechanical properties of the material, e.g. the ductile-to-brittle transition temperature is shifted to higher levels. In this study, cluster formation is characterised in high Ni (1.58%) low Cu (0.04%) steel welds identical to Ringhals R4 welds, using atom probe tomography in both surveillance material and in material irradiated at accelerated dose rates. Clusters containing mainly Ni and Mn, but also some Si and Cu were observed in all of the irradiated materials. Their evolution did not change drastically during irradiation; the clusters grew and new clusters were nucleated. Hence, both the cluster number density and the average size increased with irradiation time. Some flux effects were observed when comparing the high flux material and the surveillance material. The surveillance material has a lower cluster number density, but larger clusters. The resulting impact on the mechanical properties of these two effects cancel out, resulting in a measured hardness that seems to be on the same trend as the high flux material. The dispersed barrier hardening model with an obstacle strength factor of 0.15 was found to reproduce the increase in hardness. In the investigated high flux materials, the clusters' Cu content was higher.

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

    SciTech Connect

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

    1992-12-31

    Plates of 9Cr-1MoVNb and 12Cr-1 MoVW steels were normalized and then tempered at two different tempering conditions. One-third-size Charpy specimens from each steel were irradiated to 7.4-8{times}10{sup 26} n/m{sup 2} (about {approximately}35 dpa) at 420{degrees}C in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility. Specimens were also thermally aged to 20,000 h at 400{degrees}C to compare the effect of aging and irradiation. Previous work on the steels irradiated to 4-5 dpa at 365{degrees}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.

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

    SciTech Connect

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

    1992-12-31

    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 420C in the Materials Open Test Assembly of the Fast Flux Test Facility. Specimens were also thermally aged to 20000 h at 400C to determine the effect of aging during irradiation. Previous work on these steels irradiated to 4--5 dpa at 365C 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.

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

    SciTech Connect

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

    1992-01-01

    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 420C in the Materials Open Test Assembly of the Fast Flux Test Facility. Specimens were also thermally aged to 20000 h at 400C to determine the effect of aging during irradiation. Previous work on these steels irradiated to 4--5 dpa at 365C 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.

  12. 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.

  13. 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.

  14. Microstructural characterization of irradiated Fe-Cu-Ni-P model steels

    SciTech Connect

    Miller, M.K.; Hoelzer, D.T.; Ebrahimi, F.; Hawthorne, J.R.; Burke, M.G.

    1987-01-01

    The microstructure of Fe-Cu-Ni-P model pressure vessel steels after neutron irradiation and thermal aging has been characterized by atom probe field-ion microscopy and augmented by transmission electron microscopy. High densities of small, roughly spherical or disc shaped copper clusters/precipitates were observed in the neutron irradiated alloys that contained copper. Small spherical phosphorus clusters were observed in the irradiated copper-free alloys, and copper phosphides were observed in a high phosphorus Fe-Cu-Ni-P alloy. None of these clusters/precipitates were observed in the thermally aged materials. The increases in the tensile and yield strengths that were observed after neutron irradiation resulted from these clusters and other lattice defects. 14 refs., 8 figs., 2 tabs.

  15. 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.

  16. 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.

  17. 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.

  18. Irradiation embrittlement of reactor pressure vessel steel at very high neutron fluence

    NASA Astrophysics Data System (ADS)

    Kryukov, A.; Debarberis, L.; von Estorff, U.; Gillemot, F.; Oszvald, F.

    2012-03-01

    For the prediction of radiation embrittlement of RPV materials beyond the NPP design time the analysis of research data and extended surveillance data up to a fluence ˜23 × 1020 cm-2 (E > 0.5 MeV) has been carried out. The experimental data used for the analysis are extracted from the International Database of RPV materials. Key irradiation embrittlement mechanisms, direct matrix damage, precipitation and element segregation have been considered. The essential part of the analysis concerns the assessment of irradiation embrittlement of WWER-440 steel irradiated with very high neutron fluence. The analysis of several surveillance sets irradiated at a fluence up to 23 × 1020 cm-2 (E > 0.5 MeV) has been performed. The effect of the main influencing chemical elements phosphorus and copper has been verified up to a fluence of 4.6 × 1020 cm-2 (E > 0.5 MeV). The data are indicating good radiation stability, in terms of the Charpy transition temperature shift and yield strength increase for steels with relatively low concentrations of copper and phosphorus. The linear dependence between ΔTk and ΔRp0.2 can be an evidence of strengthening mechanisms of irradiation embrittlement and absence of non-hardening embrittlement even at very high neutron fluence.

  19. The radiation swelling effect on fracture properties and fracture mechanisms of irradiated austenitic steels. Part II. Fatigue crack growth rate

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

    The experimental data on the fatigue crack growth rate (FCGR) have been obtained for austenitic steel of 18Cr-10Ni-Ti grade (Russian analog of AISI 321 steel) irradiated up to neutron dose of 150 dpa with various radiation swelling. The performed study of the fracture mechanisms for cracked specimens under cyclic loading has explained why radiation swelling affects weakly FCGR unlike its effect on fracture toughness. Mechanical modeling of fatigue crack growth has been carried out and the dependencies for prediction of FCGR in irradiated austenitic steel with and with no swelling are proposed and verified with the obtained experimental results. As input data for these dependencies, FCGR for unirradiated steel and the tensile mechanical properties for unirradiated and irradiated steels are used.

  20. Status of Irradiation Tests of Dilute Uranium Alloys in NaK-Containing Stainless Steel Capsules

    SciTech Connect

    McDonell, W.R.

    2001-03-26

    To extend experience with uranium metal fuels to the high exposures required for power reactor operation, the Savannah River Laboratory has conducted over several years a series of irradiation tests of small uranium specimens of various alloy compositions in NaK-containing stainless steel capsules. These tests were designed specifically to establish the limits on exposure that could be reached during irradiation of the alloys at various temperatures without swelling and to determine the metallurgical factors that promoted the stability of the alloys. This paper discusses those test results.

  1. 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.

  2. Swelling behavior of 20% CW 316 stainless steel cladding irradiated with and without adjacent fuel

    SciTech Connect

    Makenas, B.J.; Bates, J.F.; Jost, J.W.

    1982-01-01

    Swelling behavior has been evaluated for irradiated 20% CW 316 stainless steel used as cladding material for mixed-oxide fuel pins in EBR-II. This behavior has been compared statistically with the behavior of a large number of specimens which were irradiated without adjacent fuel in the same reactor. In spite of the chemical environment and stresses experienced by fueled cladding, the fueled and nonfueled cladding appear to behave in a similar manner although some divergence was noted for one of the cases studied.

  3. Swelling behavior of 20% CW 316 Stainless Steel cladding irradiated with and without adjacent fuel. [LMFBR

    SciTech Connect

    Makenas, B.J.; Bates, J.F.; Jost, J.W.

    1982-06-01

    Swelling behavior has been evaluated for irradiated 20% CW 316 Stainless Steel used as cladding material for mixed-oxide fuel pins in EBR-II. This behavior has been compared statistically with the behavior of a large number of specimens which were irradiated without adjacent fuel in the same reactor. In spite of the chemical environment and stresses experienced by fueled cladding, the fueled and nonfueled cladding appear to behave in a similar manner although some divergence was noted for one of the cases studied.

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

    NASA Astrophysics Data System (ADS)

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

    2002-12-01

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

  5. Magnetic hysteresis properties of neutron-irradiated VVER440-type nuclear reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Kobayashi, S.; Gillemot, F.; Horváth, Á.; Székely, R.; Horváth, M.

    2012-11-01

    The development of non-destructive evaluation methods for irradiation embrittlement in nuclear reactor pressure vessel steels has a key role for safe and long-term operation of nuclear power plants. In this study, we have investigated the effect of neutron irradiation on base and weld metals of Russian VVER440-type reactor pressure vessel steels by measurements of magnetic minor hysteresis loops. A minor-loop coefficient, which is obtained from a scaling power-law relation of minor-loop parameters and is a sensitive indicator of internal stress, is found to change with neutron fluence for both metals. While the coefficient for base metal exhibits a local maximum at low fluence and a subsequent slow decrease, that for weld metal monotonically decreases with fluence. The observed results are explained by competing mechanisms of nanoscale defect formation and recovery, among which the latter process plays a dominant role for magnetic property changes in weld metal due to its ferritic microstructure.

  6. Irradiation-induced impurity segregation and ductile-to-brittle transition temperature shift in high chromium ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Lu, Z.; Faulkner, R. G.; Flewitt, P. E. J.

    2007-08-01

    A model is presented to predict irradiation-induced impurity segregation and its contribution to the ductile-to-brittle transition temperature (DBTT) shift in high chromium ferritic steels. The hardening contribution (dislocation loops, voids and precipitates) is also considered in this study. The predicted results are compared with the experimental DBTT shifts data for irradiated 9Cr1MoVNb and 12Cr1MoVW steels with different grain sizes.

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

    SciTech Connect

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

    1999-10-01

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

  8. 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.

  9. 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.

  10. Neutron Irradiation Effects on the Mechanical Properties of HY-80 Steel

    DTIC Science & Technology

    1986-12-01

    Some Pressure-Vessel Steels," ASTM STP 2761, American Society for Testing - and Materials, 1960, p. 102. E18 Hasegawa, M., "Irradiation Tests of...Materials ( ASTM ), and then included in ASTM 4. STP 457, C183. Hasegawa not only tested U. S. Navy HY-80 I steel, but he also tested several Japanese steels...0.140.860.0180.010 - 1.54 0.614.28 SIC3-ASE B.E.A.F. - 76 0314 0. 18 0.920 0.011 0.015 0.065 1’.61 0.50S3.01 ( ASTM ref) SIC3-ASA B.E.A.F. 20,000 M0 0.17 0. 290. SS0

  11. 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

  12. Mechanical and Raman spectroscopic studies of multi-ion-beam irradiated 12,18Cr-oxide dispersion strengthened steels

    NASA Astrophysics Data System (ADS)

    Zhang, Yanwen; Qian, Xin; Wang, Xu; Liu, Shiyi; Wang, Cheng; Li, Ting; Zhao, Ziqiang; Lu, Daogang

    2013-02-01

    12,18Cr-oxide dispersion strengthened (ODS) steels were irradiated at room temperature by single beam (2.2 MeV He+ or 3.0 MeV H+ or 21.0 MeV Si4+), dual-ion-beam (2.2 MeV He+ and 3.0 MeV H+) and triple-ion-beam (21.0 MeV Si4+, 2.2 MeV He+ and 3.0 MeV H+). Five combination of H, He, He+H, Si, Si+He+H irradiation were used. The mechanical properties of 12,18Cr-ODS steels exhibited that triple-ion-beam irradiation could strengthen irradiation swelling and hardening effect. Carbon segregation and several new carbon peaks appeared in the Raman spectrum of irradiated 12Cr-ODS steel. A clear correlation was established between the carbon distribution and the damage distribution.

  13. Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

    SciTech Connect

    Marquis, Emmanuelle; Wirth, Brian; Was, Gary

    2016-03-28

    Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and

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

    NASA Astrophysics Data System (ADS)

    Ha, Yoosung; Kimura, Akihiko

    2015-12-01

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

  15. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-05-01

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

  17. 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.

  18. 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.

  19. 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.

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

    NASA Astrophysics Data System (ADS)

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

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

  1. Spherical nanoindentation of proton irradiated 304 stainless steel: A comparison of small scale mechanical test techniques for measuring irradiation hardening

    NASA Astrophysics Data System (ADS)

    Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley; Vo, Hi T.; Maloy, Stuart A.; Hosemann, Peter; Mara, Nathan A.

    2017-09-01

    Experimentally quantifying the mechanical effects of radiation damage in reactor materials is necessary for the development and qualification of new materials for improved performance and safety. This can be achieved in a high-throughput fashion through a combination of ion beam irradiation and small scale mechanical testing in contrast to the high cost and laborious nature of bulk testing of reactor irradiated samples. The current work focuses on using spherical nanoindentation stress-strain curves on unirradiated and proton irradiated (10 dpa at 360 °C) 304 stainless steel to quantify the mechanical effects of radiation damage. Spherical nanoindentation stress-strain measurements show a radiation-induced increase in indentation yield strength from 1.36 GPa to 2.72 GPa and a radiation-induced increase in indentation work hardening rate of 10 GPa-30 GPa. These measurements are critically compared against Berkovich nanohardness, micropillar compression, and micro-tension measurements on the same material and similar grain orientations. The ratio of irradiated to unirradiated yield strength increases by a similar factor of 2 when measured via spherical nanoindentation or Berkovich nanohardness testing. A comparison of spherical indentation stress-strain curves to uniaxial (micropillar and micro-tension) stress-strain curves was achieved using a simple scaling relationship which shows good agreement for the unirradiated condition and poor agreement in post-yield behavior for the irradiated condition. The disagreement between spherical nanoindentation and uniaxial stress-strain curves is likely due to the plastic instability that occurs during uniaxial tests but is absent during spherical nanoindentation tests.

  2. Spherical nanoindentation of proton irradiated 304 stainless steel: A comparison of small scale mechanical test techniques for measuring irradiation hardening

    DOE PAGES

    Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley; ...

    2017-06-27

    Experimentally quantifying the mechanical effects of radiation damage in reactor materials is necessary for the development and qualification of new materials for improved performance and safety. This can be achieved in a high-throughput fashion through a combination of ion beam irradiation and small scale mechanical testing in contrast to the high cost and laborious nature of bulk testing of reactor irradiated samples. The current paper focuses on using spherical nanoindentation stress-strain curves on unirradiated and proton irradiated (10 dpa at 360 °C) 304 stainless steel to quantify the mechanical effects of radiation damage. Spherical nanoindentation stress-strain measurements show a radiation-inducedmore » increase in indentation yield strength from 1.36 GPa to 2.72 GPa and a radiation-induced increase in indentation work hardening rate of 10 GPa–30 GPa. These measurements are critically compared against Berkovich nanohardness, micropillar compression, and micro-tension measurements on the same material and similar grain orientations. The ratio of irradiated to unirradiated yield strength increases by a similar factor of 2 when measured via spherical nanoindentation or Berkovich nanohardness testing. A comparison of spherical indentation stress-strain curves to uniaxial (micropillar and micro-tension) stress-strain curves was achieved using a simple scaling relationship which shows good agreement for the unirradiated condition and poor agreement in post-yield behavior for the irradiated condition. Finally, the disagreement between spherical nanoindentation and uniaxial stress-strain curves is likely due to the plastic instability that occurs during uniaxial tests but is absent during spherical nanoindentation tests.« less

  3. 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.

  4. IASCC susceptibility of irradiated austenitic stainless steel under very low dissolved oxygen

    SciTech Connect

    Kodama, Mitsuhiro; Katsura, Ryoei; Morisawa, Junichiro; Nishimura, Seiji; Suzuki, Shunichi; Takamori, Kenro; Shima, Seishi; Kato, Takahiko

    1995-12-31

    Slow strain rate tests of Type 304 stainless steel (SS) irradiated to 1.3 {times} 10{sup 26} n/m{sup 2} (E>1MeV) were conducted in high-temperature water and argon gas environment to discuss irradiation-assisted stress corrosion cracking (IASCC) mechanism with respect to the dissolved oxygen (DO) effect. IASCC susceptibility of Type 304 SS decreased with decreasing DO. However, IASCC was not mitigated completely in the hydrogen injected water. And IG fracture was not observed in the case of argon gas environment. These results indicated that the high-temperature aqueous environment was an indispensable condition for the occurrence of IASCC. Moreover, lowering DO(<1ppb) did not necessarily eliminate IASCC susceptibility when austenitic stainless steel was irradiated to high neutron fluence. By considering H{sub 2}O{sub 2} formed by {gamma}-ray irradiation, IASCC at very low DO could not be explained by an active path corrosion model. At high DO, IASCG would be affected by the active path corrosion of radiation-induced chromium depletion. However, at very low DO, the possibility that IASCC would be affected by other mechanisms such as hydrogen embrittlement was suggested.

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

    NASA Astrophysics Data System (ADS)

    Tan, L.; Busby, J. T.

    2013-11-01

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

  6. 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.

  7. 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.

  8. The role of irradiated microstructure in the localized deformation of austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Jiao, Z.; Was, G. S.

    2010-12-01

    Localized deformation has emerged as a potential factor in irradiation-assisted stress corrosion cracking of austenitic stainless steels in LWR environments and the irradiated microstructure may be a critical factor in controlling the degree of localized deformation. Seven austenitic alloys with various compositions were irradiated using 2-3 MeV protons to doses of 1 and 5 dpa at 360 °C. The irradiated microstructure consisting of dislocation loops and voids was characterized using transmission electron microscopy. The degree of localized deformation was characterized using atomic force microscopy on the deformed samples after conducting constant extension rate tension tests to 1% and 3% strain in argon. Localized deformation was found to be dependent on the irradiated microstructure and to correlate with hardening originating from dislocation loops. Dislocation loops enhance the formation of dislocation channels and localize deformation into existing channels. On the contrast, voids mitigate the degree of localized deformation. The degree of localized deformation decreases with SFE with the exception of alloy B. Localized deformation was found to have similar dependence on SFE as loop density suggesting that SFE affects localized deformation by altering irradiated microstructure.

  9. Ion-irradiation-induced microstructural modifications in ferritic/martensitic steel T91

    NASA Astrophysics Data System (ADS)

    Liu, Xiang; Miao, Yinbin; Li, Meimei; Kirk, Marquis A.; Maloy, Stuart A.; Stubbins, James F.

    2017-07-01

    In this paper, in situ transmission electron microscopy investigations were carried out to study the microstructural evolution of ferritic/martensitic steel T91 under 1 MeV Krypton ion irradiation up to 4.2 × 1015 ions/cm2 at 573 K, 673 K, and 773 K. At 573 K, grown-in defects are strongly modified by black-dot loops, and dislocation networks together with black-dot loops were observed after irradiation. At 673 K and 773 K, grown-in defects are only partially modified by dislocation loops; isolated loops and dislocation segments were commonly found after irradiation. Post irradiation examination indicates that at 4.2 × 1015 ions/cm2, about 51% of the loops were a0 / 2 < 111 > type for the 673 K irradiation, and the dominant loop type was a0 < 100 > for the 773 K irradiation. Finally, a dispersed barrier hardening model was employed to estimate the change in yield strength, and the calculated ion data were found to follow the similar trend as the existing neutron data with an offset of 100-150 MPa.

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

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

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

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

    SciTech Connect

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

    2003-09-15

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

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

    SciTech Connect

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

    1983-01-01

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

  13. Microstructural characteristics and embrittlement phenomena in neutron irradiated 309L stainless steel RPV clad

    NASA Astrophysics Data System (ADS)

    Lee, J. S.; Kim, I. S.; Kasada, R.; Kimura, A.

    2004-03-01

    The effects of neutron irradiation on the microstructural features and mechanical properties of 309L stainless steel RPV clad were investigated using TEM, SEM, small tensile, microhardness and small punch (SP) tests. The neutron irradiations were performed at 290 °C up to the fluences of 5.1 × 10 18 and 1.02 × 10 19 n/cm 2 (>1 MeV) in Japan Materials Testing Reactor (JMTR). The microstructure of the clad before and after irradiation was composed of main part of fcc austenite, a few percent of bcc δ-ferrite and small amount of brittle σ phase. After irradiation, not only the yield stress and microhardness, but SP ductile to brittle transition temperature (SP-DBTT) were increased. However, the increase in SP-DBTT is almost saturated, independent of the neutron fluence. Based on the TEM observation, the origin of irradiation hardening was accounted for by the irradiation-produced defect clusters of invisible fine size (<1-2 nm), and the shift of SP-DBTT was primary due to the higher hardening and the preferential failure of δ-ferrite. The embrittlement of the clad was strongly affected by the initial microstructural factors, such as the amount of brittle σ phase, which caused a cracking even in an early stage of deformation.

  14. Mechanical characteristics and swelling of austenitic Fe-Cr-Mn steels irradiated in the SM-2 and BOR-60 reactors

    NASA Astrophysics Data System (ADS)

    Shamardin, V. K.; Bulanova, T. M.; Neustroev, V. S.; Ivanov, L. I.; Djomina, E. V.; Platov, Yu. M.

    1991-03-01

    Three types of austenitic Fe-Cr-Mn stainless steels were irradiated simultaneously with Fe-Cr-Ni austenitic steel at temperatures from 400 to 800°C in the mixed spectrum of the high flux SM-2 reactor to 10 dpa and 700 appm of He and in the BOR-60 reactor to 60 dpa without He generation. The paper presents the swelling and mechanical properties of steels irradiated in the BOR-60 and SM-2 as a function of the concentration of transmuted He and the value of atomic displacement.

  15. Irradiation accelerated corrosion of 316L stainless steel in simulated primary water

    NASA Astrophysics Data System (ADS)

    Raiman, Stephen S.

    The objective of this work is to understand the effects of irradiation on the corrosion of 316L stainless steel in simulated primary water. 316L stainless steel samples were irradiated with a proton beam while simultaneously exposed to simulated PWR primary water to study the effects of radiation on corrosion. A 3.2 MeV proton beam was transmitted through a 37 microm thick sample that served as a "window" into a corrosion cell containing flowing 320° C water with 3 wppm H2. This design permitted radiolysis and displacement damage to occur on the sample surface in contact with the simulated primary water environment. Samples were irradiated for 4, 12, 24, and 72 hrs at dose rates between 400 and 4000 kGy/s, corresponding to damage rates of 7x10-7 to 7x10-6 dpa/s respectively. The structure and composition of the oxide films were characterized using Raman spectroscopy, STEM, and SEM. Sample areas exposed to direct proton irradiation had inner oxide films that were thinner, more porous, and were deficient in chromium when compared to unirradiated oxides. Outer oxides on irradiated samples exhibited a smaller particle size, and had a significant amount of hematite, which was not found on unirradiated samples. The presence of hematite on irradiated samples indicates an increase in electrochemical potential due to irradiation. Dissolution of chromium-rich spinels due to the elevated potential is identified as a likely mechanism behind the loss of inner oxide chromium. It is suggested that the loss of inner-oxide chromium leads to a less protective inner oxide, and a higher rate of oxide dissolution. Sample areas that were not irradiated, but were exposed to the flow of radiolyzed water, exhibited most of the same phenomena found on irradiated areas including loss of Cr and thinner more porous oxides, indicating that water radiolysis is the primary mechanism. When a sample with a pre-formed oxide was irradiated in the same conditions, the region exposed to radiolyzed

  16. 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.

  17. 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.

  18. Irradiation effects on precipitation and its impact on the mechanical properties of reduced-activation ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Tanigawa, H.; Sakasegawa, H.; Hashimoto, N.; Klueh, R. L.; Ando, M.; Sokolov, M. A.

    2007-08-01

    It was previously reported that reduced-activation ferritic/martensitic steels (RAFs) showed a variety of changes in ductile-brittle transition temperature (DBTT) and yield stress after irradiation at 573 K up to 5 dpa. The precipitation behavior of the irradiated steels was examined and the presence of irradiation induced precipitation which works as if it was forced to reach the thermal equilibrium state at irradiation temperature 573 K. In this study, transmission electron microscopy was performed on extraction replica specimens to analyze the size distribution of precipitates. It turned out that the hardening level multiplied by the square root of the average block size showed a linear dependence on the extracted precipitate weight. This dependence suggests that the difference in irradiation hardening between RAFs was caused by different precipitation behavior on block, packet and prior austenitic grain boundaries during irradiation. The simple Hall-Petch law could be applicable for interpreting this dependence.

  19. Investigation of the development of irradiation-induced precipitates in VVER-440 type reactor pressure vessel steels and weld metals after irradiation and annealing

    SciTech Connect

    Grosse, M.; Nitzsche, P.; Boehmert, J.; Brauer, G.

    1999-10-01

    The development of irradiation-induced precipitates in VVER-440 type reactor pressure vessel steels 15Kh2MFA and weld metals SV-10KhMFT during irradiation and post-irradiation annealing is studied by small angle neutron and X-ray scattering. The kinetic conditions for the precipitation of particles, which already exist in the unirradiated state, seem to be improved at temperatures of about 270 C due to the irradiation. The size distribution of the irradiation-induced precipitates depends on the copper content and differs between weld and base metal. A strong correlation between the formation of irradiation-induced precipitates and the irradiation hardening is found. The hardness nearly linearly depends on the number of these precipitates.

  20. 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.

  1. Analytical Transmission Electron Microscopy Characterization of Stress Corrosion Cracks in an Irradiated Type 316 Stainless Steel Core Component

    SciTech Connect

    Thomas, Larry E.; Bruemmer, Stephen M.

    2002-05-31

    Irradiation-assisted stress-corrosion cracking (IASCC) of a cold-worked type 316 stainless steel baffle/former bolt from a pressurized-water reactor (PWR) was investigated by analytical transmission electron microscopy (ATEM). Nanometer-resolution methods for feature-specific analysis were used to characterize irradiation and corrosion-affected microstructures of the crack tip. The work is part of an international cooperative program to characterize light-water-reactor core components that experience IASCC. This is the first detailed ATEM examination of in-service cracks in neutron-irradiated austenitic stainless steel.

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

    SciTech Connect

    Gelles, D.S.; Shibayama, T.

    1998-09-01

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

  3. Accelerated corrosion and oxide dissolution in 316L stainless steel irradiated in situ in high temperature water

    NASA Astrophysics Data System (ADS)

    Raiman, Stephen S.; Was, Gary S.

    2017-09-01

    316L stainless steel samples were irradiated with a proton beam while simultaneously exposed to high temperature water with added hydrogen (320 °C, 3 wppm H2, neutral pH) to study the effect of radiation on stainless steel corrosion. Irradiated samples had thinner and more porous inner oxides with a lower chromium content when compared to unirradiated samples. Observations suggest that depletion of chromium from the inner oxide can be attributed to the dissolution of chromium-rich spinel oxides in irradiated water, leading to an accelerated rate of inner oxide dissolution. Sample areas which were not irradiated, but were exposed to the flow of irradiated water were also found to be porous and deficient in chromium, indicating that these phenomena can be attributed primarily to water radiolysis. A new empirical equation for oxide growth and dissolution is used to describe the observed changes in oxide thickness under irradiation. An experiment in which a stainless steel sample was exposed to high temperature water (320 °C, 3 wppm H2, neutral pH) without irradiation, and then exposed for a second time with irradiation was conducted to observe the effect of irradiation on a pre-formed protective film. After the irradiated exposure, the sample exhibited chromium loss in regions which were directly irradiated, but not on regions exposed only to irradiated water, suggesting that a pre-formed protective oxide may be effective in preventing chromium loss due to irradiated water. Additionally, this observation suggests that enhanced kinetics under irradiation may have accelerated dissolution of chromium from the inner oxide.

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

    SciTech Connect

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

    2013-01-01

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

  5. Research on the energy coupling coefficient of 45# steel and 304 stainless steel under 3.8μm CW laser irradiation

    NASA Astrophysics Data System (ADS)

    Zhang, Xiangyu; Chen, Minsun; Zhao, Guomin; Zhang, Tianyu

    2016-11-01

    Experimental research on the energy coupling characteristic of 45# steel and 304 stainless steel under mid-infrared CW laser irradiation is carried out. Based on the classical electromagnetic theory, the theoretical formula of the energy coupling coefficient is derived under ideal condition. In order to obtain the energy coupling coefficient, an experimental system for reflectance measurement is set up by an integrating sphere. The curves about energy coupling coefficient and the temperature variation are measured respectively. The mechanism about the variation of energy coupling coefficient of sample under mid-infrared CW laser irradiation is also discussed. The experimental results show that the energy coupling coefficient of sample increased with temperature rising, but the curve in the heating stage is not consistent with the curve in the cooling stage, which means the change of the energy coupling coefficient is not a reversible process. Combined with the experimental phenomena and the energy dispersive spectrometry, the qualitative analysis about the differences between the 45# steel and 304 stainless steel is presented after irradiation. It indicates that the oxidation reaction has a significant effect on the laser interaction with sample. Accordingly, the variation of coupling coefficient of 304 stainless steel is not as obvious as that of 45# steel.

  6. 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

  7. 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.

  8. 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.

  9. 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.

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

    NASA Astrophysics Data System (ADS)

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

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

  11. 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.

  12. 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.

  13. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    NASA Astrophysics Data System (ADS)

    Chiapetto, M.; Messina, L.; Becquart, C. S.; Olsson, P.; Malerba, L.

    2017-02-01

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a "grey-alloy" approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  14. Nondestructive evaluation of neutron irradiation embrittlement for reactor vessel steel by magnetomechanical acoustic emission technique

    SciTech Connect

    Maeda, Noriyoshi; Yamaguchi, Atsunori; Saito, Kiyoshi; Hirasawa, Taiji; Komura, Ichiroh; Chujou, Noriyuki

    1999-10-01

    A modified magnetomechanical acoustic emission (MAE) technique denoted Pulse MAE, in which the magnetizing current has a rectangular wave form, was developed as an NDE technique. Its applicability to the radiation damage for reactor pressure vessel steel was evaluated. The reactor pressure vessel steel A533B base metal and weld metal were irradiated to the two fluence levels: 5 {times} 10{sup 22} and 3 {times} 10{sup 23} n/m{sup 2} at 288 C. One side of the specimen was electropolished after irradiation. Pulse MAE signals were measured with a 350 kHz resonance frequency AE sensor at the moment when the magnetizing voltage is applied from zero to the set-up value abruptly. The AE signals were analyzed and the peak voltage Vp was determined for the measuring parameter. The peak voltage Vp showed the tendency to increase monotonically with increasing neutron fluence. The relationship between the Vp and mechanical properties such as yield stress, tensile strength and Charpy transition temperature were also obtained. The Pulse MAE technique proved to have the possibility to detect and evaluate the neutron irradiation embrittlement. The potential of the Pulse MAE as an effective NDE technique and applicability to the actual components are discussed.

  15. 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.

  16. 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

  17. Effect of carbon and nitrogen on grain boundary segregation in irradiated stainless steels

    NASA Astrophysics Data System (ADS)

    Kano, F.; Fukuya, K.; Hamada, S.; Miwa, Y.

    1998-10-01

    SUS304 stainless steels with carbon contents of 0.052%, 0.019% and 0.004% and SUS316L stainless steels with nitrogen contents of 0.095%, 0.032% and 0.003% were irradiated with 12 MeV Ni ions at 573 K to a dose of 1 dpa at 1 μm depth. Microstructure and grain boundary chemical composition were investigated using a transmission electron microscope with a field-emission-gun (FE-TEM) at the probe size of 0.5 nm. The number density of dislocation loop was higher as the carbon content was higher and was almost independent of nitrogen content. With increasing carbon and nitrogen content, the degree of Cr depletion and Si/Ni segregation was decreased. Both carbon and nitrogen suppressed the Cr depletion and Si/Ni segregation. The suppression effect of carbon was larger than that of nitrogen.

  18. 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.

  19. Formation of high spatial frequency ripples in stainless steel irradiated by femtosecond laser pulses in water

    NASA Astrophysics Data System (ADS)

    Huo, Yanyan; Jia, Tianqing; Feng, Donghai; Zhang, Shian; Liu, Jukun; Pan, Jia; Zhou, Kan; Sun, Zhenrong

    2013-05-01

    We report the formation of high spatial frequency ripples (HSFRs) in stainless steel irradiated by 50 fs, 800 nm, 1 kHz femtosecond laser pulses in water. The period of the HSFRs, Λ, is less than 0.2λ, where λ is the laser wavelength. We further conduct theoretical calculations to study the ultrafast dynamics, and find that double splitting of the low spatial frequency ripples (LSFRs, Λ > 0.45λ) plays a decisive role in the formation of HSFRs. Similar experiments are conducted in copper, however, no splitting of LSFRs is observed. The different experimental results on stainless steel and copper conducted in water and in air are also discussed.

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

    NASA Astrophysics Data System (ADS)

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

    2011-02-01

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

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

    SciTech Connect

    Chopra, O. K.; Roa, A. S.; Environmental Science Division; U.S. NRC

    2010-12-15

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

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

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

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

  3. 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.

  4. Dynamic fracture toughness of irradiated A533 Grade B Class 1 pressure vessel steel

    SciTech Connect

    Murty, K.L.; Bamford, W.H.; Shogun, R.P.

    1984-03-01

    The effect of neutron radiation on the fracture characteristics of an A533 Grade B Class 1 pressure vessel steel was investigated using standard and instrumented precracked Charpy impact tests. Use of the instrumented impact test with precracked specimens has allowed fracture toughness values to be determined from the Charpy test. Neutron exposure resulted in minute decreases in the upper-shelf Charpy energy and fracture toughness, and an increase in the ductile brittle transition temperature (DBTT). The Charpy transition temperature shifted about 29 K while the fracture toughness shift was about 20 K. The temperature variation of the dynamic yield strength exhibited dips at DBTT for both unirradiated archive and irradiated materials.

  5. 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.

  6. Change in the properties of FeCrNi and FeCrMn austenitic steels under mixed and fast neutron irradiation

    NASA Astrophysics Data System (ADS)

    Shamardin, V. K.; Bulanova, T. M.; Golovanov, V. N.; Neustroyev, V. S.; Povstyanko, A. V.; Ostrovsky, Z. E.

    1996-10-01

    Detailed investigations are performed on mechanical properties, swelling and structure of different types of FeCrNi and FeCrMn austenitic stainless steels irradiated in the SM-2 high-flux research reactor and BOR-60 fast reactor. Steel irradiation temperatures are ranging from 100 up to 800°C and the maximum achieved level of damage doses is 60 dpa for FeCrMn steel (with 4-5% of Ni) and 30 dpa for steels of the C12Cr20MnWT type. Presented are dose dependencies of swelling and mechanical properties of FeCrNi and FeCrMn steels. It is shown that at temperatures below 530°C the investigated FeCrMn steel systems are less susceptible to swelling as compared to FeCrNi ones. FeCrMn steels showed a lower value of irradiation embrittlement after irradiation in the mixed spectrum at temperatures from 100 up to 400°C and much higher embrittlement after irradiation from 350 up to 400°C in the fast spectrum in comparison with FeCrNi steels. Higher hardening rate of FeCrMn steels after their irradiation in BOR-60 is attributed to the presence of dislocation loops and defects of high density in the structure. The structural change features in FeCrMn steels under irradiation are considered taking into account austenite stabilization in the initial state.

  7. 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.

  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. Magnetic properties of a highly neutron-irradiated nuclear reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Kobayashi, S.; Gillemot, F.; Horváth, Á.; Székely, R.

    2012-02-01

    We report results of minor B- H loop measurements on a highly neutron-irradiated A533B-type reactor pressure vessel steel. A minor-loop coefficient, which is a sensitive indicator of internal stress, changes with neutron fluence, but depends on relative orientation to the rolling direction in the low fluence regime. At a higher fluence of ˜10 × 10 23 m -2, on the other hand, an anomalous increase of the coefficient was detected irrespective of the orientation. The results were interpreted as due to competing irradiation mechanisms of the formation of Cu-rich precipitates, recovery process, and the formation of late-blooming Mn-Ni-Si-rich clusters.

  10. 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.

  11. 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.

  12. Ion irradiation effects on a martensitic stainless steel designed for reduced long-life radioactivity

    SciTech Connect

    Griffin, R.D. . Fusion Technology Inst.); Zinkle, S.J. ); Dodd, R.A.; Kulcinski, G.L. ); Gelles, D.S. )

    1990-04-01

    Alloys with reduced long-life radioactivity (low activation alloys) are being developed to increase the acceptability of fusion power. The phase stability and swelling resistance of a 12Cr-6.5Mn-1W-0.3V-0.1C martensitic steel were evaluated by transmission electron microscopy following 3.8 MeV Fe{sup ++} ion irradiation with and without He coimplantation. Ion irradiations were performed at 450{degree}C, 550{degree}C, and 650{degree}C to approximately 10, 20, and 40 dpa. At 550{degree}C, approximately 20 appm He/dpa was coimplanted with the 3.8 MeV Fe{sup ++} ions. The specimens were examined at a depth approximately halfway between the surface and the mean ion range in order to minimize the influence of the surface and of injected ions. At all temperatures, M{sub 23}C{sub 6}, also present in the unirradiated structure, was the only precipitate present. A nonuniform distribution of loops also formed at all temperatures. After the 450{degree}C and 650{degree}C irradiations, no voids were present. At 550{degree}C, the helium did not appear to have much effect. Very few faceted voids formed. At 20 and 40 dpa some bubbles were found but their density was very low. At 650{degree}C, a structure similar to a heavily over-tempered steel was produced by the irradiation. At 550{degree}C recovery was seen to a lesser extent. Little to no recovery was seen at 450{degree}C.

  13. Irradiation and annealing behavior of 15Kh2MFA reactor pressure vessel steel

    SciTech Connect

    Popp, K.; Bergmann, U.; Bergner, F.; Hampe, E.; Leonhardt, W.D.; Schuetzler, H.; Viehrig, H.

    1993-12-01

    This work deals with the mechanical properties of reactor pressure vessel (RPV) steels used in the pressurized water reactors (PWR) of former Soviet type WWER-440. The materials under investigation were a forging (base metal 15Kh2MFA) and the corresponding weld. Charpy 5-notch specimens and tensile test specimens were irradiated in the PWR WWER-2 Rheinsberg at about 270 C up to the two neutron fluence levels of 4 {times} 10{sup 18} and 5 {times} 10{sup 19} n/cm{sup 2} (E > 1 MeV). Post irradiation annealing heat treatments were performed, among others a 475 C/152 h treatment of technical interest. A set of experimental data is given regarding the influence of sampling depth (through-thickness position within the forging), neutron irradiation, and annealing on the properties derived from instrumented Charpy impact testing, tensile and hardness tests. The ferrite content varies through the thickness of the forging. The variation of the mechanical properties can be explained qualitatively with the varying ferrite content. The surface layer of the forging is more sensitive to neutron irradiation than material from the 1/4-T position. To evaluate the effect of annealing heat treatment, the kinetics of the recovery process for the hardness has been investigated. The recovery coefficients for different mechanical properties and parameters have been compared. The annealing behavior is too complex to predict the effect of a large-scale annealing of an RPV on the basis of single hardness measurements.

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

    DOE PAGES

    Chen, Wei-Ying; Li, Meimei; Kirk, Marquis A.; ...

    2015-08-21

    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 linemore » 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. Finally, we attributed this difference 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.« less

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

    SciTech Connect

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

    2015-08-21

    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. Finally, we attributed this difference 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.

  16. 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.

  17. 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.

  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. Prediction of yield stress and Charpy transition temperature in highly neutron irradiated ferritic steels

    NASA Astrophysics Data System (ADS)

    Windsor, Colin; Cottrell, Geoff; Kemp, Richard

    2010-07-01

    Recent predictions have been made of metallurgical properties of low-activation ferritic/martensitic steels alloys at the high irradiation levels (displacements per atom or dpa) needed for a fusion power plant as based on measurements at low irradiation levels where more data are available. These predictions have been published for the yield stress and for the Charpy ductile to brittle transition temperature shift. The neural network model predictions use training data up to a certain dpa level to predict metallurgical properties above this level. This 'extrapolation' mode of neural networks is explored in some detail. Our studies revealed an increasing accuracy of predictions as the test dpa level is increased for both yield stress and Charpy shift predictions. This result suggests that a model exists for these metallurgical properties as a function of dpa level which becomes more accurate as the available irradiation range in the training data is increased. The explanation suggested is that the metallurgical annealing, which occurs as the irradiation level is increased, simplifies the microstructure and makes prediction more reliable.

  1. High Strain Fatigue Properties of the F82H Ferritic-Martensitic Steel under Proton Irradiation.

    SciTech Connect

    Marmy, P; Oliver, Brian M. )

    2003-05-15

    During the up and down cycles of a fusion reactor, the first wall is exposed concomitantly to a flux of energetic neutrons that generates radiation defects and to a neutron thermal flux that induces thermal stresses. The resulting strains may exceed the elastic limit and induce a plastic deformation in the material. A similar situation occurs in the window of a spallation liquid source target and results in the same type of damage. This particular loading has been simulated in F82H martensitic ferritic steel, using a device allowing a fatigue test to be carried out during irradiation with 590 MeV protons. All fatigue tests were carried out at 300?C, in a strain controlled test at strain levels around 0.8%. Two different signals have been used: a fully symmetrical triangle wave signal (R=-1) and a triangle ramp with 2 min tension holds. The fatigue was investigated under three different conditions: unirradiated , irradiated and post irradiation tested, and finally in beam tested. The main result is that the in beam tested specimens have the lowest life as compared to the post irradiation tested specimens and unirradiated specimens. Hydrogen is suspected to be the main contributor to the observed embrittlement.

  2. High strain fatigue properties of F82H ferritic martensitic steel under proton irradiation

    NASA Astrophysics Data System (ADS)

    Marmy, P.; Oliver, B. M.

    2003-05-01

    During the up and down cycles of a fusion reactor, the first wall is exposed concomitantly to a flux of energetic neutrons that generates radiation defects and to a thermal flux that induces thermal stresses. The resulting strains may exceed the elastic limit and induce plastic deformation in the material. A similar situation occurs in the window of a spallation liquid source target and results in the same type of damage. This particular loading has been simulated in F82H ferritic-martensitic steel, using a device allowing a fatigue test to be carried out during irradiation with 590 MeV protons. All fatigue tests were carried out in a strain controlled test at strain levels around 0.8% and at 300 °C. Two different signals have been used: a fully symmetrical triangle wave signal ( R=-1) and a triangle ramp with 2 min tension holds. The fatigue was investigated under three different conditions: unirradiated, irradiated and post-irradiation tested, and finally in-beam tested. The main result is that the in-beam tested specimens have the lowest life as compared to the post-irradiation tested specimen and unirradiated specimen. Hydrogen is suspected to be the main contributor to the observed embrittlement.

  3. On the character of nanoscale features in reactor pressure vessel steels under neutron irradiation

    NASA Astrophysics Data System (ADS)

    Wirth, Brian David

    Nanostructural features that form in reactor pressure vessel steels under neutron irradiation at around 290°C are responsible for significant hardening and embrittlement. It is well established that the nanostructural features can be separated into well formed precipitates and matrix features comprised of point defect clusters complexed with solutes, which may also include regions of solute enrichment that are not well formed precipitates. However, a more detailed atomicscale understanding of these features is needed to better interpret experimental measurements and provide a physical basis for predictive embrittlement models. The overall objective of this work is to provide atomic-level insight into the character of the nanostructural features and the physical processes involved in their formation. One focus of this work has been on modeling cascade aging; defined as the evolution of self-interstitial and vacancy defects spanning from their spatially correlated birth in displacement cascades over picoseconds to times on the order of >10 5 seconds, when defect populations have built up to steady-state values and no longer have a geometric correlation. During cascade aging, the self-interstitial and vacancy fluxes are responsible for radiation enhanced diffusion, resulting in wellformed precipitates, and are a direct source of matrix defect features. Many-bodied molecular-statics energy relaxation methods have been used to investigate the structure and energetics of self-interstitial and vacancy clusters. The characterization reveals that self-interstitial clusters form as highly kinked, prismatic, perfect proto dislocation loops and vacancy clusters form as faceted three-dimensional clusters. Molecular dynamics simulations of self-interstitial cluster migration reveal that they undergo easy one-dimensional glide, probably due to the presence and easy motion of intrinsic kinks. Our study of the structural characteristics and mobility of the self

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

    SciTech Connect

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

    2001-09-01

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

  5. Self-ordered defect structures in two model F/M steels under in situ ion irradiation

    NASA Astrophysics Data System (ADS)

    Kaoumi, D.; Adamson, J.

    2014-05-01

    Two model F/M steels, 9Cr-model and 12Cr-model, were irradiated with 1 MeV Kr ions in situ in a TEM at temperatures between 20 K and 573 K to doses as high as 15 dpa. During the early stages of irradiation of the two F/M steels, defect clusters were rather uniformly distributed within grains, and a saturation density was quickly reached. However, at higher doses, self-ordering alignments of defect clusters were found in some grains. The regularly ordered arrays of small loops were observed in the two F/M steels along <1 1 0> directions with spacing about 30-50 nm. Once the aligned structure was created, it was stable under further irradiation. The possible mechanisms for the “self-organization”/“ordering” of the clusters were investigated. This paper describes the process and its temperature dependence, and the possible mechanisms are discussed.

  6. 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.

  7. Modeling the post-yield flow behavior after neutron and electron irradiation of steels and iron-base alloys.

    SciTech Connect

    Dimelfi, R. J.

    1999-01-13

    Irradiation hardening is an issue of practical importance as it relates to the remanent life and the nature of failure of reactor components exposed to displacement-producing radiation. For example, irradiation-induced yield strength increases in pressure vessel steels are directly related to increases in the ductile-to-brittle-transition-temperature of these materials. Other issues associated with hardening, such as reductions in ductility, toughness and fatigue life of structural steels are also of concern. Understanding these phenomena requires studies of fundamental microstructural mechanisms of hardening. Because of the limited supply of neutron-irradiated surveillance material, difficulties posed by the radioactivity of neutron-exposed samples and the uncertainty of irradiation conditions in this case, fundamental studies are often conducted using well-controlled experiments involving irradiation by electrons instead of neutrons. Also, in such studies, simple model alloys are used in place of steels to focus on the influence of specific alloy constituents. It is, therefore, important to understand the relationship between the results of this kind of experiment and the effects of in-reactor neutron exposure in order to use them to make predictions of significance to reactor component life. In this paper, we analyze the tensile behavior of pressure vessel steels (A212B and A350) irradiated by neutrons and electrons. The results show that the post-yield true stress/true strain behavior can provide fingerprints of the different hardening effects that result from irradiation by the two particles, which also reflect the influence of alloy content. Microstructurally-based models for irradiation-induced yield strength increases, combined with a model for strain hardening, are used to make predictions of the different effects of irradiation by the two particles on the entire flow curve that agree well with data.

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

    SciTech Connect

    Corwin, W.R.

    1996-08-01

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

  9. Neutron exposure parameters for the fifth heavy section steel technology irradiation series

    SciTech Connect

    Stallmann, F.W.; Kam, F.B.K.; Baldwin, C.A.

    1985-07-01

    The NRC's Heavy Section Steel Technology (HSST) Program is concerned with the investigation of cracklike flaws in reactor pressure vessel steels. In the fifth irradiation series, capsules containing a variety of metallurgical test specimens were irradiated to fluences in the range of 1 . 10/sup 19/ to 3 . 10/sup 19/ neutrons/cm/sup 2/ (E > 1.0 MeV). In order to correlate radiation embrittlement to damage fluences, accurate determination of the neutron fluence spectra at the critical location of the test specimen is needed. The part of the neutron spectrum which is responsible for the radiation damage is characterized as ''damage exposure parameter.'' Fluences for energies greater than 1.0 MeV (F > 1.0 MeV) is the most widely used parameter; however, current thinking favors displacements per atom (dpa) in iron as better related to the physical mechanism of radiation damage. Fluences for energies greater than 0.1 MeV (F > 0.1 MeV) are also considered since neutrons in the 0.1 and 1.0 MeV range are likely to contribute to the damage. In order not to prejudice future investigations, all three damage parameters F > 1.0, MeV F > 0.1 MeV, and dpa are listed in this report.

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

    NASA Astrophysics Data System (ADS)

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

    2014-08-01

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

  11. 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.

  12. 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.

  13. 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.

  14. Effects of neutron irradiation on hydrogen-induced intergranular fracture in a low activation 9%Cr-2%W steel

    NASA Astrophysics Data System (ADS)

    Kimura, A.; Kayano, H.; Narui, M.

    1991-03-01

    Hydrogen charging changed the fracture mode in tensile tests at room temperature from ductile shear rupture to intergranular cracking, resulting in a considerable reduction of the ductility of a low activation 9%Cr-2%W martensitic steel. The critical hydrogen charging current density required to cause hydrogen-induced intergranular cracking was reduced by neutron irradiation, suggesting that neutron irradiation enhanced hydrogen-induced intergranular cracking. This hydrogen-induced intergranular cracking was not caused by irreversible damage due to hydrogen charging, since it disappeared after aging at room temperature. The recovery rate of the fracture mode from intergranular cracking to ductile rupture during aging at room temperature was reduced by irradiation. A mechanism of irradiation-induced enhancement of hydrogen embrittlement in a low activation 9%Cr-2%W martensitic steel is proposed.

  15. The tensile and fatigue properties of DIN 1.4914 martensitic stainless steel after 590 MeV proton irradiation

    NASA Astrophysics Data System (ADS)

    Marmy, P.; Victoria, M.

    1992-09-01

    Tensile and low cycle fatigue subsize specimens of DIN 1.4914 martensitic steel (MANET) have been irradiated with 590 MeV protons to doses up to 1 dpa and at temperatures between 363 and 703 K. The helium produced by spallation reactions was measured as 130 appm/dpa. A strong radiation hardening is found, which decreases as the irradiation temperature increases. The tensile elongation is reduced after irradiation, but the fracture mode is always ductile and transgranular. The radition hardening produced at low irradiation temperatures is recovered after annealing at higher temperatures. Continous softening is observed during low cycle fatigue testing. The rate of softening of the irradiated material is stonger than that of the unirradiated material and tends to reach the saturation level of the latter. The irradiation badly affects the fatigue life, particularly in the temperature domain of dynamic strain ageing between 553 and 653 K.

  16. Synthesis of atom probe experiments on irradiation-induced solute segregation in French ferritic pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Auger, P.; Pareige, P.; Welzel, S.; Van Duysen, J.-C.

    2000-08-01

    Microstructural changes due to neutron irradiation cause an evolution of the mechanical properties of reactor pressure vessels (RPV) steels. This paper aims at identifying and characterising the microstructural changes which have been found to be responsible in part for the observed embrittlement. This intensive work relies principally on an atom probe (AP) study of a low Cu-level French RPV steel (Chooz A). This material has been irradiated in in-service conditions for 0-16 years in the frame of the surveillance program. Under this aging condition, solute clustering occurs (Cu, Ni, Mn, Si, P, …). In order to identify the role of copper, experiments were also carried out on Fe-Cu model alloys submitted to different types of irradiations (neutron, electron, ion). Cu-cluster nucleation appears to be directly related to the presence of displacement cascades during neutron (ion) irradiation. The operating basic physical process is not clearly identified yet. A recovery of the mechanical properties of the irradiated material can be achieved by annealing treatments (20 h at 450°C in the case of the RPV steel under study, following microhardness measurements). It has been shown that the corresponding microstructural evolution was a rapid dissolution of the high number density of irradiation-induced solute clusters and the precipitation of a very low number density of Cu-rich particles.

  17. 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.

  18. 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

  19. 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

  20. 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.

  1. In-situ observation of an austenitic stainless steel weld joint during helium irradiation

    NASA Astrophysics Data System (ADS)

    Hamada, S.; Hojou, K.; Hishinuma, A.

    1993-10-01

    Microstructural evolution during helium-ion irradiation at 773 K in a weld metal sample (containing 10% δ-ferrite) of Ti-modified austenitic stainless steel was observed in-situ using a transmission electron microscope. Very fine helium bubbles formed at high number density in both δ-ferrite and austenite by a dose of 3 × 10 19ionsm-2. However, the different microstructural evolution developed in the two phases with increasing dose. Fine bubbles in δ-ferrite rapidly grew with increasing dose and coalescenced at doses beyond 9 × 10 19ionsm-2. Tiny bubbles in austenite remained very stable during irradiation to a dose of 2 × 10 20ionsm-2. The number density of bubbles was about one order of magnitude larger in austenite than that in δ-ferrite, and increased with increasing dose. Swelling became much larger in δ-ferrite than in austenite, as a result. This is the inverse phenomenon to the conventional result that swelling is lower in ferrite than in austenite. Sigma phase formed by radiation-enhancement at grain boundaries between δ-ferrite and austenite and at the interfaces within δ-ferrite at a dose 9 × 10 19ionsm-2 and grew with increasing dose. The chemical composition of σ-phase formed during irradiation showed Cr and Mo enrichment, and Fe and Ni depletion compared with σ-phase formed thermally.

  2. Void-precipitate association during neutron irradiation of austenitic stainless steel

    SciTech Connect

    Pedraza, D.F.; Maziasz, P.J.

    1986-01-01

    Microstructural data has recently become available on a single heat of 316 stainless steel irradiated in EBR-II and HFIR, over a wide range of irradiation temperature (55 to 750/sup 0/C), dose (7 to 75 dpa), and helium generation rate (0.5 to 55 at. ppM He/dpa). Extensive information on precipitate compositions and characteristics are included. The data reveal several important relationships between the development of voids and precipitation. Precipitate associated voids dominate the swelling of (DO heat) 316 at 500 to 650 C from 8.4 to 36 dpa in EBR-II. Cold work (CW) or helium preinjection delay void formation in EBR-II. Higher helium generation in HFIR also delays void formation at 500 to 640/sup 0/C in SA 316 and CW DO heat 316. The delay persists in CW 316 at least to 61 dpa in HFIR, but abundant matrix and precipitate-associated voids form in SA after 47 dpa. In another heat of CW 316 (N-lot) irradiated in HFIR matrix and precipitate voids form readily after 22 to 44 dpa at 500 to 600/sup 0/C.

  3. Effects of water chemistry on intergranular cracking of irradiated austenitic stainless steels

    SciTech Connect

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

    1995-12-31

    To determine the effects of water chemistry on the susceptibility to irradiation-assisted stress corrosion cracking (IASCC) in austenitic stainless steels, constant-extension-rate tests were conducted in simulated BWR environments on several heats of high- and commercial-purity (HP and CP) Type 304 SS specimens from BWR components irradiated to fluences up to 2.4 {times} 10{sup 21} n cm{sup {minus}2} (E > 1 MeV). Effects of dissolved oxygen (DO) and electrochemical potential (ECP) in 289 C water were investigated. Dependence of susceptibility to intergranular stress corrosion cracking (IGSCC) on DO was somewhat different for the two materials. Susceptibility of the HP heats, less influenced by DO and ECP, was higher than that of CP material for all DO and fluence levels. Percent IGSCC in the CP material was negligible for DO < 0.01 ppm or ECP <{minus}140 mV SHE. Results of analysis by Auger electron spectroscopy indicated that the HP neutron absorber tubes were characterized by relatively lower concentrations of Cr, Ni, and Li and relatively higher concentrations of F and N on grain boundaries than those of the CP materials. It is suggested that a synergism between irradiation-induced grain-boundary Cr depletion and fabrication-related fluorine contamination plays an important role in the stress corrosion cracking behavior of the HP neutron absorber tubes.

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

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale J.; Was, Gary S.

    2016-12-01

    This study intended to understand how dislocation channeling affects IASCC initiation using a novel four-point bend test. Stainless steels used in this study (irradiated in the BOR-60 reactor) included a commercial purity 304L alloy irradiated to 5.5, 10.2, and 47.5 dpa, and two high purity alloys, Fesbnd 18Crsbnd 12Ni and Fesbnd 18Crsbnd 25Ni, irradiated to ∼10 dpa. IASCC was enhanced by MnS inclusions, which dissolve in the NWC environment and form oxide caps, creating a crevice condition with a high propensity for crack initiation. Stress concentration at the grain boundary intersecting these sites induced crack initiation, resulting from discontinuous dislocation channels (DC). Stress to initiate IASCC decreased with dose due to earlier DC initiation. The HP Fesbnd 18Crsbnd 12Ni alloy had low IASCC susceptibility and the high Ni alloy did not crack. The difference was attributed to the propensity for DCs to transmit across grain boundaries, which controls stress accumulation at DC - grain boundary intersections.

  5. Irradiation assisted stress corrosion cracking of controlled purity 304L stainless steels

    NASA Astrophysics Data System (ADS)

    Cookson, J. M.; Carter, R. D.; Damcott, D. L.; Atzmon, M.; Was, G. S.

    1993-06-01

    The effect of chromium, phosphorus, silicon and sulfur on the stress corrosion cracking of 304L stainless steel in CERT tests in high purity water or argon at 288°C following irradiation with 3.4 MeV protons at 400°C to 1 dpa, has been investigated using ultrahigh purity alloys (UHP) with controlled impurity additions. Grain boundary segregation of phosphorus or silicon due to proton irradiation was quantified using both Auger electron spectroscopy and scanning transmission electron microscopy, and the alloys with impurity element additions were observed to have greater grain boundary chromium depletion and nickel enrichment than the UHP alloy. The UHP alloy suffered severe cracking in CERT tests in water. Less cracking was found after CERT test of irradiated UHP+Por UHP+Si alloys, despite greater chromium depletion. This suggests a mitigating effect of phosphorus and silicon at grain boundaries. No cracking was found in argon tests, eliminating a purely mechanical embrittlement mechanism, but not eliminating a contribution from radiation hardening. Implanted hydrogen was not a factor in the intergranular cracking found.

  6. 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.

  7. 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.

  8. Mechanism of irradiation assisted stress corrosion crack initiation in thermally sensitized 304 stainless steel

    NASA Astrophysics Data System (ADS)

    Onchi, T.; Dohi, K.; Soneda, N.; Navas, Marta; Castaño, M. L.

    2005-04-01

    Thermally sensitized 304 stainless steels, irradiated up to 1.2 × 1021 n/cm2 (E > 1 MeV), were slow-strain-rate-tensile tested in 290 °C water containing 0.2 ppm dissolved oxygen (DO), followed by scanning and transmission electron microscopic examinations, to study mechanism of irradiation-assisted-stress-corrosion-crack (IASCC) initiation. Intergranular (IG) cracking behaviors changed at a border fluence (around 1 × 1020 n/cm2), above which deformation twinning were predominant and deformation localization occurred earlier with increasing fluence. The crack initiation sites tended to link to the deformation bands, indicating that the crack initiation may be brought about by the deformation bands interacted with grain boundaries. Thus the border fluence is equivalent to the IASCC threshold fluence for the sensitized material, although the terminology of IASCC is originally given to the non-sensitized materials without microstructural definition. The IASCC threshold fluence was found to change with irradiation conditions. Changes in IASCC susceptibility and IASCC threshold fluence with fluence and DO were further discussed.

  9. 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.

  10. 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

  11. 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.

  12. Influence of irradiation number of high current pulsed electron beam on the structure and properties of M50 steel

    NASA Astrophysics Data System (ADS)

    Xu, Fangjun; Tang, Guangze; Guo, Guangwei; Ma, Xinxin; Ozur, G. E.

    2010-08-01

    The influence of high current pulsed electron beam (HCPEB) irradiation numbers on the microstructure, wear and corrosion resistance properties of M50 steel was studied. The crystallize phase, surface morphology, hardness, oxidation wear and corrosion resistance of samples were analyzed using XRD, SEM, nanoindenter, wear tester and electrochemical corrosion tests. The results reveal that the hardness and wear resistance of irradiated samples decrease compared with untreated sample because of the increasing of austenite content in the melted layer; while the corrosion resistance of irradiated samples is higher than untreated sample.

  13. The microstructure of the 1.4914 MANET martensitic steel before and after irradiation with 590 MeV protons

    NASA Astrophysics Data System (ADS)

    Gavillet, D.; Marmy, P.; Victoria, M.

    1992-09-01

    Optical and transmission electron microscope observations, together with SEM (scanning electron microscope) and ASTEM (analytical scanning transmission electron microscope) microanalysis have been performed in samples of the DIN 1.4914 martensitic steel (MANET cast), both before and after irradiation with 590 MeV protons to doses up to 1 dpa at temperatures between 363 and 703 K. The chemical composition of the different carbide geometries have been obtained. No substantial modification of the carbide and precipitate structure is observed after either deformation under fatigue or after irradiation to 1 dpa at 703 K. No bubbles have been observed in a specimen irradiated to 0.7 dpa, containing 87 appm He.

  14. Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50-400)°C

    NASA Astrophysics Data System (ADS)

    Kuleshova, E. A.; Gurovich, B. A.; Bukina, Z. V.; Frolov, A. S.; Maltsev, D. A.; Krikun, E. V.; Zhurko, D. A.; Zhuchkov, G. M.

    2017-07-01

    This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50-400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔTK) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects - dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔTK shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔTK shift in the studied range of irradiation temperature and fluence.

  15. 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.

  16. Study of Fe-12Cr-20Mn-W-C austenitic steels irradiated in the SM-2 reactor

    NASA Astrophysics Data System (ADS)

    Shamardin, V. K.; Bulanova, T. M.; Neustroyev, V. S.; Ostrovsky, Z. E.; Kosenkov, V. M.; Ivanov, L. I.; Djomina, E. V.

    1992-09-01

    A comparison has been made between the mechanical properties and swelling of austenitic stainless steels EP-838 (Fe-Cr-Mn) and 316SS (Fe-Cr-Ni) irradiated in the mixed-neutron spectrum of the SM-2 reactor in the temperature range 400-800°C (every 100°C) to 16 dpa dose with 1000 and 3000 appm helium generation correspondingly, determined by nickel content. EP-838 exhibited less susceptibility to void swelling and radiation hardening. Fe-12Cr-20Mn-W-0.1C steel without nickel irradiated at 100°C to 21 dpa exhibited significant radiation hardening accompanied by α-phase formation in the steel structure.

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

    SciTech Connect

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

    1998-02-01

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

  18. In situ micro-tensile testing on proton beam-irradiated stainless steel

    NASA Astrophysics Data System (ADS)

    Vo, H. T.; Reichardt, A.; Frazer, D.; Bailey, N.; Chou, P.; Hosemann, P.

    2017-09-01

    Small-scale mechanical testing techniques are currently being explored and developed for engineering applications. In particular, micro-tensile testing can add tremendous value, since the entire stress-strain curve, including the strain to failure, can be measured directly. In this work, 304 stainless steel specimens irradiated with 2 MeV protons to 10 dpa (full-cascade setting in the Stopping and Range of Ions in Matter, SRIM, software) at 360 °C was evaluated using micro-tensile testing. It was found that even on the micron scale, the measured strain corresponds well with macroscopic expectations. In addition, a new approach to analyzing sudden slip events is presented.

  19. 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.

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

    SciTech Connect

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

    1996-04-01

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

  1. 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.

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

    SciTech Connect

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

    1984-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2013-01-01

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

  4. Irradiation effects on impact behavior of 12Cr-1MoVW and 2 1/4Cr-1Mo steels

    SciTech Connect

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

    1990-01-01

    Charpy impact tests were conducted on 12CR-1MoVW steel after irradiation in the Fast Flux Test Facility (FFTF) and the Oak Ridge Research Reactor (OOR). One-half-size and one-third-size Charpy specimens were irradiated in FFTF at 365{degree}C. After irradiating half-size-specimens to {approximately}10 and 17 dpa, a shift in ductile-brittle-transition temperature (DBTT) of 160{degree}C was observed for both fluences, indicating a saturation in the shift. A shift DBTT of 151{degree}C was observed for the third-size specimens after irradiation to 10 dpa. Third-size specimens of 12Cr--1MoVW steel irradiated {approximately}7 dpa in the ORR at 330 and 400{degree}C developed shifts in DBTT of 200 and 120{degree}C, respectively, somewhat above and below the shifts observed after irradiation at 365{degree}C in FFTF. This correspondence of results in the mixed-spectrum ORR and the fast-spectrum FFTF is in marked contrast to large differences observed between specimens irradiated in the mixed-spectrum High Flux Isotope Reactor and the fast spectrum Experimental Breeder Reactor. The first data on the effect of fast reactor irradiation on the impact behavior 2{1/4} CR--1Mo steel were obtained. Third-size specimens were irradiated in FFTF to {approximately}10 dpa at 365{degree}C. An increase in DBTT of 170{degree}C was observed, similar of the shift observed for 12Cr--1MoVW steel following comparable irradiation. The reduction in the upper-shelf energy for the 2{1/4} Cr--1Mo steel was less than that observed for 12Cr--1MoVW steel. Because of the low DBTT of unirradiated 2{1/4} Cr--1Mo steel, the DBTT after irradiation remained below that for 12Cr--1MoVW steel.

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

    NASA Astrophysics Data System (ADS)

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

    2000-12-01

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

  6. 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.

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

    DOE PAGES

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, 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

  8. Changes in electromagnetic properties of a low-alloy steel caused by neutron irradiation

    SciTech Connect

    Goto, Toru; Kamimura, Takeo; Kumano, Shintaro; Takeuchi, Iwao; Maeda, Noriyoshi; Yamaguchi, Atsunori

    1999-10-01

    In order to develop a method for the nondestructive evaluation of material deterioration in nuclear pressure vessels, changes in the electromagnetic properties of the low-alloy steel A533B, Class 1 and its weld metal caused by neutron irradiation up to {approximately}3 {times} 10{sup 23} n/m{sup 2} of neutron fluence at 561 K were measured. Electrical resistance, coercivity and Barkhausen noise were selected as the electromagnetic properties to measure. It was found that decreases of several percent in the readings of electrical resistance and coercivity, and an increase of several percent in the Barkhausen noise occurred due to neutron irradiation. Good correlations between the changes in the electromagnetic properties and those in the mechanical properties were confirmed. Furthermore, an equation using the results of the three tests was found to estimate well the transition temperature and yield strength. From this, the authors conclude that the electromagnetic tests have potential as methods for nondestructive evaluation of material deterioration in the reactor vessels of nuclear power plants.

  9. 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.; ...

    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

  10. 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.

  11. 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.

  12. 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.

  13. Effects of irradiation fluence and creep on fracture toughness of 347/348 stainless steel

    SciTech Connect

    Haggag, F M; Server, W L; Reuter, W G; Beeston, J M

    1984-01-01

    The postirradiation fracture toughness of Type 347/348 stainless steel was investigated using 5.08-mm thick three-point bend specimens tested at 427/sup 0/C. The J/sub Ic/ values were determined using the single-specimen unloading compliance technique in accordance with ASTM E 813-81. Equivalent values of plane strain fracture toughness, K/sub Ic/, were computed from experimentally determined J/sub Ic/ values for several fluence levels ranging from 2.3 to 4.8 x 10/sup 22/ n/cm/sup 2/ (E > 1.0 MeV) and for irradiation creep of 0.0, 0.6, 1.1, and 1.8%. The test matrix involved four variables: fluence, creep, helium content, and heat-to-heat variation. Results show that an interpolated trend exists, i.e., K/sub Ic/ decreases with increasing combinations of fluence, creep, and helium content. These results also suggest that irradiation creep has less effect on reducing K/sub Ic/ than has been suggested previously.

  14. Effects of solute elements on irradiation hardening and microstructural evolution in low alloy steels

    NASA Astrophysics Data System (ADS)

    Fujii, Katsuhiko; Ohkubo, Tadakatsu; Fukuya, Koji

    2011-10-01

    The effects of the elements Mn, Ni, Si and Cu on irradiation hardening and microstructural evolution in low alloy steels were investigated in ion irradiation experiments using five kinds of alloys prepared by removing Mn, Ni and Si from, and adding 0.05 wt.%Cu to, the base alloy (Fe-1.5Mn-0.5Ni-0.25Si). The alloy without Mn showed less hardening and the alloys without Ni or Si showed more hardening. The addition of Cu had hardly any influence on hardening. These facts indicated that Mn enhanced hardening and that Ni and Si had some synergetic effects. The formation of solute clusters was not confirmed by atom probe (AP) analysis, whereas small dislocation loops were identified by TEM observation. The difference in hardening between the alloys with and without Mn was qualitatively consistent with loop formation. However, microstructural components that were not detected by the AP and TEM were assumed to explain the hardening level quantitatively.

  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.; ...

    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. Investigation of microstructural evolution under neutron irradiation in Eurofer97 steel by means of small-angle neutron scattering

    NASA Astrophysics Data System (ADS)

    Coppola, R.; Lindau, R.; May, R. P.; Möslang, A.; Valli, M.

    2009-04-01

    Small-angle neutron scattering (SANS) has been utilized to investigate in Eurofer97 steel (9Cr, 0.01C, 1W, 0.2V Fe bal wt%) the microstructural effect of neutron irradiation at 300 °C up to a dose level of 8.4 dpa. For each irradiated sample an unirradiated reference was measured to distinguish as accurately as possible the actual effect of the neutron irradiation. The SANS measurements were carried out at the D22 diffractometer at the High-Flux Reactor of the Institut Max von Laue-Paul Langevin, Grenoble, France. Analysing separately the nuclear and magnetic SANS components obtained after subtraction of the reference from the irradiated sample it appears that the microstructural inhomogeneities produced under such irradiation conditions are non-magnetic ones, such as microvoids. Their size distributions are presented and compared with those previously obtained for the same steel irradiated at 2.5 dpa: with increasing the dose, the volume fraction is increased by a factor of 2 roughly, while the average size of these inhomogeneities remains nearly unchanged.

  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. Radiation induced segregation and precipitation behavior in self-ion irradiated Ferritic/Martensitic HT9 steel

    NASA Astrophysics Data System (ADS)

    Zheng, Ce; Auger, Maria A.; Moody, Michael P.; Kaoumi, Djamel

    2017-08-01

    In this study, Ferritic/Martensitic (F/M) HT9 steel was irradiated to 20 displacements per atom (dpa) at 600 nm depth at 420 and 440 °C, and to 1, 10 and 20 dpa at 600 nm depth at 470 °C using 5 MeV Fe++ ions. The characterization was conducted using ChemiSTEM and Atom Probe Tomography (APT), with a focus on radiation induced segregation and precipitation. Ni and/or Si segregation at defect sinks (grain boundaries, dislocation lines, carbide/matrix interfaces) together with Ni, Si, Mn rich G-phase precipitation were observed in self-ion irradiated HT9 except in very low dose case (1 dpa at 470 °C). Some G-phase precipitates were found to nucleate heterogeneously at defect sinks where Ni and/or Si segregated. In contrast to what was previously reported in the literature for neutron irradiated HT9, no Cr-rich α‧ phase, χ-phases, η phase and voids were found in self-ion irradiated HT9. The difference of observed microstructures is probably due to the difference of irradiation dose rate between ion irradiation and neutron irradiation. In addition, the average size and number density of G-phase precipitates were found to be sensitive to both irradiation temperature and dose. With the same irradiation dose, the average size of G-phase increased whereas the number density decreased with increasing irradiation temperature. Within the same irradiation temperature, the average size increased with increasing irradiation dose.

  1. Effect of laser and/or electron beam irradiation on void swelling in SUS316L austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Yang, Subing; Yang, Zhanbing; Wang, Hui; Watanabe, Seiichi; Shibayama, Tamaki

    2017-05-01

    Large amounts of void swelling still limit the application of austenitic stainless steels in nuclear reactors due to radiation-induced lattice point defects. In this study, laser and/or beam irradiation was conducted in a temperature range of 573-773 K to explore the suppression of void swelling. The results show that during sequential laser-electron beam irradiation, the void nucleation is enhanced because of the vacancy clusters and void nuclei formed under pre-laser irradiation, causing greater void swelling than single electron beam irradiation. However, simultaneous laser-electron dual-beam irradiation exhibits an obvious suppression effect on void swelling due to the enhanced recombination between interstitials and vacancies in the temperature range of 573-773 K; especially at 723 K, the swelling under simultaneous dual-beam irradiation is 0.031% which is only 22% of the swelling under electron beam irradiation (0.137%). These results provide new insight into the suppression of void swelling during irradiation.

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

    NASA Astrophysics Data System (ADS)

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

    2014-06-01

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

  3. 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.

  4. Effect of Neutron Irradiation on Mechanical Behavior of Ultra-Fine Grained Low Carbon Steel -- Application to Next Generation Fission Reactors

    NASA Astrophysics Data System (ADS)

    Alsabbagh, Ahmad Hesham Hasan

    Designing materials that can enhance performance and withstand extreme reactor operational conditions is a grand challenge in nuclear materials research. Irradiation induced defects result in embrittlement and hardening of reactor structural materials. Hence, the ability to mitigate the effects of radiation damage by removing in-situ radiation induced point defects is crucial to improving the mechanical properties of irradiated metals and enhancing their tailored response in irradiation environments. Ultra-fine grained steel provides large free surface to volume ratio, acting as sinks for migrating irradiation induced point defects. Annihilation of point defects at grain boundaries leads to lower net defect concentration in the grain interior compared to coarser grained counterpart thereby limiting radiation damage effects and resulting in enhanced radiation tolerant structural materials. Neutron irradiation effects on ultra-fine grain (UFG) low carbon ferritic steel prepared by equal channel angular pressing (ECAP) have been examined. Counterpart samples with conventional grain (CG) sizes were prepared by annealing at high temperatures and have been irradiated alongside with the UFG ones for comparison. Samples were irradiated in the PULSTAR reactor at North Carolina State University to relatively low dose (0.001 dpa) and in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 1.37 dpa. Low dose irradiation of ultrafine grained carbon steel revealed minute radiation effects in contrast to the distinct radiation hardening and reduction of ductility in its CG counterpart. At higher irradiation dose, atom probe tomography revealed manganese and silicon-enriched clusters in both UFG and CG steel after neutron irradiation. X-ray quantitative analysis showed that dislocation density in CG steel increased after irradiation while no significant change was observed in UFG steel, revealing better radiation tolerance. Quantitative correlations between

  5. Tomographic atom probe characterization of the microstructure of a cold worked 316 austenitic stainless steel after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Etienne, A.; Radiguet, B.; Pareige, P.; Massoud, J.-P.; Pokor, C.

    2008-11-01

    For the first time, chemical analyses using Atom Probe Tomography were performed on a bolt made of cold worked 316 austenitic stainless steel, extracted from the internal structures of a pressurized water reactor after 17 years of reactor service. The irradiation temperature of these samples was 633 K and the irradiation dose was estimated to 12 dpa (7.81 × 10 25 neutrons.m -2, E > 1 MeV). The samples were analysed with a laser assisted tomographic atom probe. These analyses have shown that neutron irradiation has a strong effect on the intragranular distribution of solute atoms. A high number density (6 × 10 23 m -3) of Ni-Si enriched and Cr-Fe depleted clusters was detected after irradiation. Mo and P segregations at the interfaces of these clusters were also observed. Finally, Si enriched atmospheres were seen.

  6. 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.

  7. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2012-03-01

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

  9. Irradiation-induced changes of the atomic distributions around the interfaces of carbides in a nuclear reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Toyama, T.; Tsuchiya, N.; Nagai, Y.; Almazouzi, A.; Hatakeyama, M.; Hasegawa, M.; Ohkubo, T.; van Walle, E.; Gerard, R.

    2010-10-01

    Irradiation-induced changes of the atomic distributions of solute and impurity elements around carbides in a reactor pressure vessel steel of a Belgium nuclear power reactor were investigated by laser-assisted local electrode-type three-dimensional atom probe, before and after in-service irradiation of 12 years. Before irradiation, nano-scale Fe-Mn-Cr-Mo carbides were found to be intragranular. The atomic distributions of Mn, Cr and Mo inside the carbide indicate that their concentrations around the inner carbide-matrix interface were enhanced, while a clear segregation of P at the interface was observed. After irradiation, the Mn concentration in the carbide increased substantially. In addition, the enhancement of Mn, Cr and Mo concentrations around the interface and the segregation of P were markedly intensified.

  10. 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.

  11. 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.

  12. Radiolysis driven changes to oxide stability during irradiation-corrosion of 316L stainless steel in high temperature water

    NASA Astrophysics Data System (ADS)

    Raiman, Stephen S.; Bartels, David M.; Was, Gary S.

    2017-09-01

    316L stainless steel samples were irradiated with a proton beam while simultaneously exposed to high temperature water with hydrogen (320 °C, 3 wppm H2, neutral pH) to study the effect of radiation on corrosion. The inner oxides on irradiated samples were found to be depleted in chromium when compared to the inner oxides on unirradiated samples exposed to the same conditions. Additionally, hematite was found on the oxide surfaces of irradiated samples, but not on unirradiated samples. Sample areas which were not directly irradiated but were exposed to the flow of irradiated water also exhibited chromium-deficient inner oxides and had hematite on their surfaces, so it is concluded that water radiolysis is the primary driver of both effects. Thermodynamic calculations and radiolysis modeling were used to show that radiolytic production of hydrogen peroxide was sufficient to raise corrosion potential high enough to cause the dissolution of chromium-rich spinel oxides which make up the inner oxide layer on stainless steel in high temperature water.

  13. 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.

  14. Charpy Impact Properties of Reduced-Activation Ferritic/Martensitic Steels Irradiated in HFIR up to 20 dpa

    SciTech Connect

    Tanigawa, H.; Shiba, K.; Sokolov, M.A.; Klueh, R.L.

    2003-07-15

    The effects of irradiation up to 20 dpa on the Charpy impact properties of reduced-activation ferritic/martensitic steels (RAFs) were investigated. The ductile-brittle transition temperature (DBTT) of F82H-IEA shifted up to around 323K. TIG weldments of F82H showed a fairly small variation on their impact properties. A finer prior austenite grain size in F82H-IEA after a different heat treatment resulted in a 20K lower DBTT compared to F82H-IEA after the standard heat treatment, and that effect was maintained even after irradiation. Helium effects were investigated utilizing Ni-doped F82H, but no obvious evidence of helium effects was obtained. ORNL9Cr-2WVTa and JLF-1 steels showed smaller DBTT shifts compared to F82H-IEA.

  15. Isolation of the role of radiation-induced segregation in irradiation-assisted stress corrosion cracking of proton-irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Busby, Jeremy Todd

    2001-11-01

    The role of radiation-induced segregation (RIS) in irradiation-assisted stress corrosion cracking (IASCC) was studied in order to better understand the underlying mechanisms of IASCC. High-purity 304L (HP-304L), commercial purity 304 (CP-304) and commercial purity 316 (CP-316) stainless steel alloys were irradiated with 3.2 MeV protons at 400°C (HP-304L) and 360°C (CP-304 and CP-316) to doses ranging from 0.1 and 5.0 dpa. Grain boundary chemistry was measured using scanning transmission electron microscopy with energy-dispersive spectroscopy (STEM/EDS) in both unirradiated and irradiated samples. Unirradiated and irradiated samples of the two commercial purity alloys were also strained to failure in an aqueous environment representative of boiling water reactor cores. The cracking susceptibility and RIS in the proton-irradiated CP-304 is very similar to that from the neutron-irradiated samples. The CP-316 alloy did not crack. Radiation-induced segregation, cracking susceptibility, and dislocation loop microstructure developed at the same rate as a function of dose in the CP-304 alloy. To isolate the effects of RIS in IASCC, post-irradiation annealing was utilized. Simulations of post-irradiation annealing of RIS and dislocation loop microstructure show that dislocation loops are removed preferentially over RIS due to the density of vacancies required and kinetic considerations. Experimental anneals were conducted on HP-304L samples irradiated to 1.0 dpa and CP-304 samples irradiated to 1.0 and 2.5 dpa. Post-irradiation anneals were performed at temperatures ranging from 400°C to 650°C for times between 45 minutes and 5 hours. At all temperatures, the hardness and dislocation densities decreased with increasing annealing time much faster than RIS did. Annealing at 600°C for 90 minutes removed virtually all dislocation microstructure while leaving RIS intact. Cracking susceptibility in the CP-304 alloy was mitigated rapidly during post-irradiation annealing

  16. Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

    DOE PAGES

    Mamivand, Mahmood; Yang, Ying; Busby, Jeremy T.; ...

    2017-03-11

    The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300–400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10–7 dpa/s and 390 °C) and then usemore » the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10–8 dpa/s and 275 °C). The model accurately predicts the γ' (Ni3Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γ' for typical CW 316 SS (with 0.054 wt% carbon) under LWR extended life conditions. Finally, this work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.« less

  17. Characterization of neutron-irradiated HT-UPS steel by high-energy X-ray diffraction microscopy

    NASA Astrophysics Data System (ADS)

    Zhang, Xuan; Park, Jun-Sang; Almer, Jonathan; Li, Meimei

    2016-04-01

    This paper presents the first measurement of neutron-irradiated microstructure using far-field high-energy X-ray diffraction microscopy (FF-HEDM) in a high-temperature ultrafine-precipitate-strengthened (HT-UPS) austenitic stainless steel. Grain center of mass, grain size distribution, crystallographic orientation (texture), diffraction spot broadening and lattice constant distributions of individual grains were obtained for samples in three different conditions: non-irradiated, neutron-irradiated (3dpa/500 °C), and irradiated + annealed (3dpa/500 °C + 600 °C/1 h). It was found that irradiation caused significant increase in grain-level diffraction spot broadening, modified the texture, reduced the grain-averaged lattice constant, but had nearly no effect on the average grain size and grain size distribution, as well as the grain size-dependent lattice constant variations. Post-irradiation annealing largely reversed the irradiation effects on texture and average lattice constant, but inadequately restored the microstrain.

  18. 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.

  19. 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.

  20. Void swelling induced by 1 MeV electron irradiation in Ti- and Nb-modified 316 stainless steels

    NASA Astrophysics Data System (ADS)

    Liu, B. X.; Lai, S. L.; Sun, J. G.; Shang, C. H.; Xu, D.

    1990-12-01

    Four 316 stainless steels, cold-worked and modified with minor elements of Ti and Nb, were irradiated by 1 MeV electrons in a HVEM at temperatures ranging from 823 to 883 K, up to 80 dpa. Void swelling, void densities and mean diameters were determined in each case. The experimental results show that swelling increases linearly with increasing of dose after an incubation period. Compared with the results obtained from solution-annealed 316 stainless steels, it is found that the influence of cold work on void swelling depends strongly on the modifying elements, and that Ti-modified 316 steels are superior to Nb-modified ones in swelling resistance.

  1. 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.

  2. Effects of grain boundary misorientation on solute segregation in thermally sensitized and proton-irradiated 304 stainless steel

    NASA Astrophysics Data System (ADS)

    Duh, T. S.; Kai, J. J.; Chen, F. R.

    2000-12-01

    The purpose of this study is to investigate the effects of the grain boundary misorientation on the radiation-induced segregation (RIS) in 304 stainless steels. There were four test conditions for the specimens: (1) as-received (AR) with enriched Cr at grain boundary, (2) AR + 1 dpa proton irradiation at 450°C, (3) thermally sensitized (SEN), and (4) SEN + 1 dpa proton irradiation at 450°C. The Cr/Ni-concentration profiles were measured by using FEGTEM/EDS and the grain boundary misorientation was determined with the aid of simulated Kikuchi patterns. A delayed Cr depletion compared to no pre-enrichment condition was found at grain boundaries in AR + 1 dpa specimens. The Cr-concentration profile gets narrower and deeper in SEN + 1 dpa specimens. The degree of grain boundary segregation was observed to be higher at random boundaries than special boundaries. The segregation cusps were measured at grain boundaries of Σ3,Σ9 and Σ15 in SEN + 1 dpa 304 stainless steel specimens. From the fitted segregation cusps, it seems that the Cr segregation level at special boundaries in irradiated sensitized 304 stainless steels increases with Σ for values up to Σ=15.

  3. The effect of low dose rate irradiation on the swelling of 12% cold-worked 316 stainless steel.

    SciTech Connect

    Allen, T. R.

    1999-03-02

    In pressurized water reactors (PWRs), stainless steel components are irradiated at temperatures that may reach 400 C due to gamma heating. If large amounts of swelling (>10%) occur in these reactor internals, significant swelling related embrittlement may occur. Although fast reactor studies indicate that swelling should be insignificant at PWR temperatures, the low dose rate conditions experienced by PWR components may possibly lead to significant swelling. To address these issues, JNC and ANL have collaborated to analyze swelling in 316 stainless steel, irradiated in the EBR-II reactor at temperatures from 376-444 C, at dose rates between 4.9 x 10{sup {minus}8} and 5.8 x 10{sup {minus}7} dpa/s, and to doses of 56 dpa. For these irradiation conditions, the swelling decreases markedly at temperatures less than approximately 386 C, with the extrapolated swelling at 100 dpa being around 3%. For temperatures greater than 386 C, the swelling extrapolated to 100 dpa is around 9%. For a factor of two difference in dose rate, no statistically significant effect of dose rate on swelling was seen. For the range of dose rates analyzed, the swelling measurements do not support significant (>10%) swelling of 316 stainless steel in PWRs.

  4. 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.

  5. 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.

  6. On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld.

    PubMed

    Lindgren, Kristina; Stiller, Krystyna; Efsing, Pål; Thuvander, Mattias

    2017-03-21

    Radiation induced clustering affects the mechanical properties, that is the ductile to brittle transition temperature (DBTT), of reactor pressure vessel (RPV) steel of nuclear power plants. The combination of low Cu and high Ni used in some RPV welds is known to further enhance the DBTT shift during long time operation. In this study, RPV weld samples containing 0.04 at% Cu and 1.6 at% Ni were irradiated to 2.0 and 6.4×1023 n/m2 in the Halden test reactor. Atom probe tomography (APT) was applied to study clustering of Ni, Mn, Si, and Cu. As the clusters are in the nanometer-range, APT is a very suitable technique for this type of study. From APT analyses information about size distribution, number density, and composition of the clusters can be obtained. However, the quantification of these attributes is not trivial. The maximum separation method (MSM) has been used to characterize the clusters and a detailed study about the influence of the choice of MSM cluster parameters, primarily on the cluster number density, has been undertaken.

  7. 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.

  8. 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).

  9. 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.

  10. Shear compression testing of glass-fibre steel specimens after 4K reactor irradiation: Present status and facility upgrade

    SciTech Connect

    Gerstenberg, H.; Kraehling, E.; Katheder, H.

    1997-06-01

    The shear strengths of various fibre reinforced resins being promising candidate insulators for superconducting coils to be used tinder a strong radiation load, e.g. in future fusion reactors were investigated prior and subsequent to reactor in-core irradiation at liquid helium temperature. A large number of sandwich-like (steel-bonded insulation-steel) specimens representing a widespread variety of materials and preparation techniques was exposed to irradiation doses of up to 5 x 10{sup 7} Gy in form of fast neutrons and {gamma}-radiation. In a systematic study several experimental parameters including irradiation dose, postirradiation storage temperature and measuring temperature were varied before the determination of the ultimate shear strength. The results obtained from the different tested materials are compared. In addition an upgrade of the in-situ test rig installed at the Munich research reactor is presented, which allows combined shear/compression loading of low temperature irradiated specimens and provides a doubling of the testing rate.

  11. Microstructural development due to long-term aging and ion irradiation behavior in weld metals of austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Nakata, K.; Ikeda, S.; Hamada, S.; Hishinuma, A.

    1996-10-01

    In a candidate austenitic stainless steel (316F) for fusion reactor structural materials, irradiation behavior of the weld metal produced by electron-beam welding (containing 7.9 vol% δ-ferrite) was investigated in terms of microstructural development. The densities of interstitial clusters in the γ-phase of the weld metal irradiated with He-ions at 673 and 773 K were about four times larger than those in 316F. Voids were formed in the δ-ferrite of the weld irradiated at 773 K. The number of clusters decreased in the weld metal (γ-phase) aged at 773 to 973 K, compared with that in the as-welded metal. The change in cluster density could be attributed to a Ni concentration increase in the γ-phase of the weld metal during aging.

  12. Mechanical properties of type 316 stainless steel materials after irradiation at 515/sup 0/C and 585/sup 0/C

    SciTech Connect

    Blackburn, L.D.; Greenslade, D.L.; Ward, A.L.

    1981-04-01

    Three different heats of type 316 SS base metal plate metals and three different weld metals produced by shielded metal arc, submerged arc, and gas tungsten arc processes with type 316 SS filler metal were used in the Gas-Cooled Fast Reactor (GCFR) Structural Materials Irradiation Experiment. Pre-irradiation strength and ductility properties over the range 24/sup 0/C to 650/sup 0/C were very similar for the three base metal heats and were within the expected range for this alloy. The three welds showed variations in strength and ductility before irradiation, but properties were generally within the range of previous experience for austenitic stainless steel welds. Weld metals showed higher yield strength but lower uniform and total elongations than those of base metal.

  13. 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.

  14. The effect of high-energy electron-beam irradiation on microstructural modification of a high-speed steel roll

    NASA Astrophysics Data System (ADS)

    Suh, Dongwoo; Lee, Sunghak; Koo, Yangmo; Lee, Hui Choon

    1996-10-01

    The purpose of this study is to investigate the microstructural modification in a high-speed steel (HSS) roll irradiated with an accelerated high-energy electron beam. The HSS roll samples were irradiated at the beam travel speeds of 2.5 to 25 mm/s using an electron accelerator (1.4 MeV). The microstructure was examined with a scanning electron microscope (SEM) capable of in situ fracture testing and simultaneous measurement of the apparent fracture toughness. Irradiation changed the matrix phase from tempered martensite to a mixture of retained austenite and martensite. Coarse primary carbides were partially or completely dissolved, depending on the heat input. Irradiation greatly improved the fracture properties because of the presence of retained austenite, which could retard crack propagation, although hardness was decreased. Occasional interior quench cracks were found in the heat-affected region. Appropriate processing methods, such as pre- or postirradiation, were suggested. A heat transfer analysis of the irradiated surface layer was also carried out to elucidate the influence of the irradiation parameters on the microstructure.

  15. 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.

  16. Examination of irradiated 304L stainless steel to 6061-T6 aluminum inertia welded transition joints after irradiation in a spallation neutron

    SciTech Connect

    Dunn, K.A.

    2000-04-28

    The Savannah River Technology Center (SRTC) designed and fabricated tritium target/blanket assemblies which were irradiated for six months at the Los Alamos Neutron Science Center (LANSCE). Cooling water was supplied to the assemblies through 1 inch diameter 304L Stainless Steel (SS) tubing. To attach the 304L SS tubing to the modules a 304L SS to 6061-T6 Aluminum (Al) inertia welded transition joint was used. These SS/Al inertia weld transition joints simulate expected transition joints in the Accelerator Production of Tritium (APT) Target/Blanket where as many as a thousand SS/Al weld transition joints will be used. Materials compatibility between the 304L SS and the 6061-T6 Al in the spallation neutron environment is a major concern as well as the corrosion associated with the cooling water flowing through the piping. The irradiated inertia weld examination will be discussed.

  17. 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.

  18. 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.

  19. Effect of neutron irradiation on the microstructure and the mechanical and corrosion properties of the ultrafine-grained stainless Cr-Ni steel

    NASA Astrophysics Data System (ADS)

    Maksimkin, O. P.; Gusev, M. N.; Tsai, K. V.; Yarovchuk, A. V.; Rybalchenko, O. V.; Enikeev, N. A.; Valiev, R. Z.; Dobatkin, S. V.

    2015-12-01

    It has been revealed that the neutron irradiation of ultrafine-grained (UFG) 08Kh18N10T steel after severe plastic deformation (SPD) does not lead to the appearance of defects of radiation origin up to a fluence of 2 × 1020 n/cm2 (~0.05 dpa) and that the strength properties of the material are retained after irradiation. At the same time, this irradiation reduces the corrosion resistance of the steel in a chlorine-containing medium, especially after heating at 550°C with a holding for 1 h after SPD.

  20. In-Situ NDT Measurements of Irradiation Induced Swelling PWR Core Internal Components; Phase 3: Correlation of Void Swelling and Material Properties of Austenitic Steels

    SciTech Connect

    I.Balachov; F. Garner; S-G. Kumatori-cho; Y. Isobe

    2004-04-01

    OAK-B135 The objective of the project is to examine and develop in-situ nondestructive testing (NDT) techniques for measuring irradiation induced swelling in the internal components for PWRs. This report documents the third phase effort on establishing experimental correlations of the irradiation induced void swelling and measurable material properties of austenitic steels and, eventually, correlation of swelling and signals of the developed swelling sensors. Experimental stainless steel irradiated at high neutron fluences are presented. Theoretical aspects of the influence of void swelling on electrical resistivity and ultrasound velocity are outlined. Swelling-material properties correlations were recommended for quantitative interpretation of swelling measurements.

  1. 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).

  2. Electrochemical and microstructural characterization of an austenitic stainless steel irradiated by heavy ions at 515°C

    NASA Astrophysics Data System (ADS)

    Bell, G. E. C.; Inazumi, T.; Kenik, E. A.; Kondo, T.

    1992-04-01

    The electrochemical and microstructural behavior of a solution-annealed, heavy-ion-irradiated, austenitic stainless steel. designated LS1A, have been investigated at 515°C after doses of 1.10 and 30 displacements per atom (dpa). Changes in electrochemical properties due to radiation-induced segregation in thin radiation-affected layers of the material were detected by the electrochemical potentiokinetic reactivation (EPR) technique using TEM disk specimens. At all doses, the Flade potential and reactivation charge were greater than those measured for thermally-aged control specimens. Grain face etching, similar to that found on EPR-tested neutron irradiated austenitic stainless steels, was observed on all specimens after testing. Duplicate heavy ion irradiated specimens were also examined by high resolution analytical electron microscopy (AEM). The 1 dpa specimen showed only a high density of small faulted dislocations (~ 10 nm), and no grain boundary precipitation or grain boundary segregation was detected. AEM confirmed chromium and molybdenum depletion at grain boundaries as measured by EPR for the 10 and 30 dpa specimens.

  3. 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.

  4. 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.

  5. 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.

  6. Effect of neutron irradiation on magnetic properties in the low alloy Ni-Mo steel SA508-3

    SciTech Connect

    Park, D.G.; Kim, C.G.; Kim, H.C.; Hong, J.H.; Kim, I.S.

    1997-04-01

    The B-H hysteresis loop and Barkhausen noise have been measured in the neutron irradiated SA508 steel of 45 {mu}m thickness. The coercive force of B-H loop showed a slow change up to a neutron dose of 10{sup 14} n/cm{sup 2} and increased by 15.4{percent} for a 10{sup 16} n/cm{sup 2} dose sample compared with that of the unirradiated one, related to the domain wall motion hindered by the increased defects. However, the amplitude of Barkhausen noise reflecting the wall motion decreased slowly up to 10{sup 14} n/cm{sup 2} irradiation, followed by a rapid decrease of 37.5{percent} at 10{sup 16} n/cm{sup 2}. {copyright} {ital 1997 American Institute of Physics.}

  7. 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.

  8. 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.

  9. Structural transformations in austenitic stainless steel induced by deuterium implantation: irradiation at 100 K

    NASA Astrophysics Data System (ADS)

    Morozov, Oleksandr; Zhurba, Volodymyr; Neklyudov, Ivan; Mats, Oleksandr; Rud, Aleksandr; Chernyak, Nikolay; Progolaieva, Viktoria

    2015-03-01

    Deuterium thermal desorption spectra were investigated on the samples of austenitic stainless steel 18Cr10NiTi preimplanted at 100 K with deuterium ions in the dose range from 3 × 1015 to 5 × 1018 D/cm2. The kinetics of structural transformation development in the implantation steel layer was traced from deuterium thermodesorption spectra as a function of implanted deuterium concentration. At saturation of austenitic stainless steel 18Cr10NiTi with deuterium by means of ion implantation, structural-phase changes take place, depending on the dose of implanted deuterium. The maximum attainable concentration of deuterium in steel is C = 1 (at.D/at.met. = 1/1). The increase in the implanted dose of deuterium is accompanied by the increase in the retained deuterium content, and as soon as the deuterium concentration attains C ≈ 0.5 the process of shear martensitic structural transformation in steel takes place. It includes the formation of bands, body-centered cubic (bcc) crystal structure, and the ferromagnetic phase. Upon reaching the deuterium concentration C > 0.5, the presence of these molecules causes shear martensitic structural transformations in the steel, which include the formation of characteristic bands, bcc crystal structure, and the ferromagnetic phase. At C ≥ 0.5, two hydride phases are formed in the steel, the decay temperatures of which are 240 and 275 K. The hydride phases are formed in the bcc structure resulting from the martensitic structural transformation in steel.

  10. Impact behavior of two low activation steels after irradiation to ˜67 dpa at 430°C

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

    Miniature CVN specimens of four martensitic steels, GA3X, F82H, GA4X and HT9, have been impact tested following irradiation at 430°C to 67 dpa. Comparison of the results with those obtained previously at lower doses indicates that the GA3X and F82H alloys, two primary candidate low activation alloys, exhibit similar behavior following irradiation at 430°C to ˜67 dpa and at 370°C to ˜15 dpa. Virtually no shift in either ductile to brittle transition temperature (DBTT) or upper shelf energy (USE) was observed in the F82H alloy at 67 dpa for either notched or precracked specimens. This absence of a shift in DBTT and USE in F82H compares with a slight increase in DBTT and a slight decrease in USE for GA3X, and much larger degradation in both properties in GA4X and HT9. The shifts in DBTT and USE observed in both GA4X and HT9 were smaller after irradiation at 430°C to ˜67 dpa than after irradiation at 370°C to ˜15 dpa.

  11. Influence of helium on deuterium retention in reduced activation ferritic martensitic steel (F82H) under simultaneous deuterium and helium irradiation

    NASA Astrophysics Data System (ADS)

    Yakushiji, K.; Lee, H. T.; Oya, M.; Hamaji, Y.; Ibano, K.; Ueda, Y.

    2016-02-01

    Deuterium and helium retention in Japanese reduced activation ferritic martensitic (RAFM) steel (F82H) under simultaneous D-He irradiation at 500, 625, 750, and 818 K was studied. This study aims to clarify tritium retention behavior in RAFM steels to assess their use as plasma facing materials. The irradiation fluence was kept constant at 1 × 1024 D m-2. Four He desorption peaks were observed with He retention greatest at 625 K. At T > 625 K a monotonic decrease in He retention was observed. At all temperatures a systematic reduction in D retention was observed for the simultaneous D-He case in comparison to D-only case. This suggests that He implanted at the near surface in RAFM steels may reduce the inward penetration of tritium in RAFM steels that would result in lower tritium inventory for a given fluence.

  12. 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.

  13. Correlation between irradiation-induced changes of microstructural parameters and mechanical properties of RPV steels

    NASA Astrophysics Data System (ADS)

    Böhmert, J.; Viehrig, H.-W.; Ulbricht, A.

    2004-08-01

    Radiation hardening, displayed by the yield stress increase, and irradiation embrittlement, described by the Charpy transition temperature shift, were experimentally determined for a broad variety of irradiation specimens machined from different reactor pressure vessel base and weld materials and irradiated in several VVER-type reactors. Additionally, the same specimens were investigated by small angle neutron scattering. The analysis of the neutron scattering data suggests the presence of nano-scaled irradiation defects. The volume fraction of these defects depends on the neutron fluence and the material. Both irradiation hardening and irradiation embrittlement correlate linearly with the square root of the defect volume fraction. However, a generally valid proportionality is only a rough approximation. In detail, chemical composition and technological pretreatment clearly affect the correlation.

  14. Effects of chemical composition and dose on microstructure evolution and hardening of neutron-irradiated reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kuramoto, A.; Kameda, J.; Toyama, T.; Nagai, Y.; Hasegawa, M.; Ohkubo, T.; Yoshiie, T.; Nishiyama, Y.; Onizawa, K.

    2010-07-01

    The correlation of microstructure evolution and hardening was studied in two kinds of A533B-1 steel with high and low levels of Cu irradiated in a range of dose from 0.32 to 9.9 × 10 19 n cm -2 ( E > 1 MeV) under a high flux of about 1.7 × 10 13 n cm -2 s -1 using three-dimensional local electrode atom probe (3DAP), positron annihilation (PA) techniques, and Vickers microhardness. The early rapid hardening was found to be caused by mainly matrix defects such as mono- or di-vacancies ( V1 - V2) and/or dislocations indicated by the PA analysis. The 3DAP analysis showed that dense dispersion of dilute Cu rich clusters and lean distribution of Mn-Ni-Si rich clusters, which were identified to possess the same dislocation-pinning effect by applying a Russell and Brown model, were responsible for large and small hardening in high- and low-Cu steels irradiated above 0.59 × 10 19 n cm 2, respectively.

  15. Influence of the copper impurity level on the irradiation response of reactor pressure vessel steels investigated by SANS

    NASA Astrophysics Data System (ADS)

    Wagner, Arne; Ulbricht, Andreas; Bergner, Frank; Altstadt, Eberhard

    2012-06-01

    Reactor pressure vessel (RPV) steel, when exposed to neutron irradiation, induces the formation of nano-sized features. Using small angle neutron scattering (SANS) we have studied the neutron fluence dependence of the precipitate volume fraction for high-Cu and low-Cu materials separately. Cu-rich precipitates have long been recognized to play the dominant role in embrittlement of Cu-bearing RPV steels. In contrast, Mn-Ni-rich precipitates seem to govern embrittlement in the case of low levels of impurity Cu. The objective is to work out the resulting differences from the microstructural point of view. For low-Cu materials, the volume fraction was found to be within the detection limit of SANS at fluences below an apparent threshold fluence, whereas the slope increases considerably beyond. The relationship between irradiation-induced yield stress increase and precipitate volume fraction was also considered. We have derived estimates of the obstacle strength for Cu-rich precipitates and for Mn-Ni-rich precipitates.

  16. Heat treatment effects on toughness of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated at 365{degrees}C

    SciTech Connect

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

    1991-12-31

    The 9Cr-1MoVNb and 12Cr-1MoVW steels were austenitized at 1040 and 1100{degrees}C to produce different prior austenite grain sizes, after which they were given different tempering treatments (1 h at 760 or 2.5 h at 780{degrees}C). Subsize Charpy impact specimens from these materials were irradiated at 365{degrees}C up to 5 dpa. For 9Cr-1MoVNb steel in the unirradiated condition, the smaller the prior austenite grain size and the higher the tempering temperature, the lower the ductile-brittle transition temperature (DBTT). Regardless of the DBTT in the unirradiated condition, however, the DBTT shift for 9Cr-1MoVNb steel due to irradiation was the same for all heat treatments. This means heat treatment can be used to ensure a lower DBTT before and after irradiation. The 12Cr-1MoVW steel showed little effect of heat treatment on DBTT in the unirradiated condition, and the shift in DBTT was relatively constant. Thus, it appears that heat treatment cannot be used to reduce the effect of irradiation on DBTT for this steel.

  17. Heat treatment effects on toughness of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated at 365 degrees C

    SciTech Connect

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

    1991-01-01

    The 9Cr-1MoVNb and 12Cr-1MoVW steels were austenitized at 1040 and 1100{degrees}C to produce different prior austenite grain sizes, after which they were given different tempering treatments (1 h at 760 or 2.5 h at 780{degrees}C). Subsize Charpy impact specimens from these materials were irradiated at 365{degrees}C up to 5 dpa. For 9Cr-1MoVNb steel in the unirradiated condition, the smaller the prior austenite grain size and the higher the tempering temperature, the lower the ductile-brittle transition temperature (DBTT). Regardless of the DBTT in the unirradiated condition, however, the DBTT shift for 9Cr-1MoVNb steel due to irradiation was the same for all heat treatments. This means heat treatment can be used to ensure a lower DBTT before and after irradiation. The 12Cr-1MoVW steel showed little effect of heat treatment on DBTT in the unirradiated condition, and the shift in DBTT was relatively constant. Thus, it appears that heat treatment cannot be used to reduce the effect of irradiation on DBTT for this steel.

  18. 55Fe effect on enhancing ferritic steel He/dpa ratio in fission reactor irradiations to simulate fusion conditions

    SciTech Connect

    Liu, Haibo; Abdou, Mohamed A.; Greenwood, Lawrence R.

    2013-11-01

    How to increase the ferritic steel He(appm)/dpa ratio in a fission reactor neutron spectrum is an important question for fusion reactor material testing. An early experiment showed that the accelerated He(appm)/dpa ratio of about 2.3 was achieved for 96% enriched 54Fe in iron with 458.2 effective full power days (EFPD) irradiation in the High Flux Isotope Reactor (HFIR), ORNL. Greenwood suggested that the transmutation produced 55Fe has a thermal neutron helium production cross section which may have an effect on this result. In the current work, the ferritic steel He(appm)/dpa ratio is studied in the neutron spectrum of HFIR with 55Fe thermal neutron helium production taken into account. The available ENDF-b format 55Fe incident neutron cross section file from TENDL, Netherlands, is first input into the calculation model. A benchmark calculation for the same sample as used in the aforementioned experiment was used to adjust and evaluate the TENDL 55Fe (n, a) cross section values. The analysis shows a decrease of a factor of 6700 for the TENDL 55Fe (n, a) cross section in the intermediate and low energy regions is required in order to fit the experimental results. The best fit to the cross section value at thermal neutron energy is about 27 mb. With the adjusted 55Fe (n, a) cross sections, calculation show that the 54Fe and 55Fe isotopes can be enriched by the isotopic tailoring technique in a ferritic steel sample irradiated in HFIR to significantly enhance the helium production rate. The results show that a 70% enriched 54Fe and 30% enriched 55Fe ferritic steel sample would produce a He(appm)/dpa ratio of about 13 initially in the HFIR peripheral target position (PTP). After one year irradiation, the ratio decreases to about 10. This new calculation can be used to guide future isotopic tailoring experiments designed to increase the He(appm)/dpa ratio in fission reactors. A benchmark experiment is suggested to be performed to evaluate the 55Fe (n, a) cross section

  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. Gamma-ray irradiation effect on corrosion rates of stainless steel, Ti and Ti-5Ta in boiling 9N nitric acid

    NASA Astrophysics Data System (ADS)

    Yamamoto, Takao; Tsukui, Shigeki; Okamoto, Shinichi; Nagai, Takayuki; Takeuchi, Masayuki; Takeda, Seiichiro; Tanaka, Yasumasa

    1996-03-01

    Irradiation effect of γ-rays on corrosion rates of stainless steel (type 304L), titanium and a titanium-tantalum alloy (Ti-5Ta) in 9N boiling nitric acid was investigated by measuring weight losses of specimens leached under a 60Co γ-ray environment of 1 kCkg -1/h (4 MR/h). Tests without irradiation were as well performed to obtain reference data. Plots of the weight loss normalized to specimen's surface area against total leaching time exhibited linear relations when the first leaching batch is neglected. The corrosion rates calculated from the gradients indicated slight, though significant, irradiation effects, an enhancement in stainless steel while suppressions in Ti and Ti-5Ta. Corrosion modes were found to be insensitive to the irradiation.

  2. Structural transformations in austenitic stainless steel induced by deuterium implantation: irradiation at 100 K.

    PubMed

    Morozov, Oleksandr; Zhurba, Volodymyr; Neklyudov, Ivan; Mats, Oleksandr; Rud, Aleksandr; Chernyak, Nikolay; Progolaieva, Viktoria

    2015-01-01

    Deuterium thermal desorption spectra were investigated on the samples of austenitic stainless steel 18Cr10NiTi preimplanted at 100 K with deuterium ions in the dose range from 3 × 10(15) to 5 × 10(18) D/cm(2). The kinetics of structural transformation development in the implantation steel layer was traced from deuterium thermodesorption spectra as a function of implanted deuterium concentration. At saturation of austenitic stainless steel 18Cr10NiTi with deuterium by means of ion implantation, structural-phase changes take place, depending on the dose of implanted deuterium. The maximum attainable concentration of deuterium in steel is C = 1 (at.D/at.met. = 1/1). The increase in the implanted dose of deuterium is accompanied by the increase in the retained deuterium content, and as soon as the deuterium concentration attains C ≈ 0.5 the process of shear martensitic structural transformation in steel takes place. It includes the formation of bands, body-centered cubic (bcc) crystal structure, and the ferromagnetic phase. Upon reaching the deuterium concentration C > 0.5, the presence of these molecules causes shear martensitic structural transformations in the steel, which include the formation of characteristic bands, bcc crystal structure, and the ferromagnetic phase. At C ≥ 0.5, two hydride phases are formed in the steel, the decay temperatures of which are 240 and 275 K. The hydride phases are formed in the bcc structure resulting from the martensitic structural transformation in steel.

  3. 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.

  4. Small punch tests on martensitic/ferritic steels F82H, T91 and Optimax-A irradiated in SINQ Target-3

    NASA Astrophysics Data System (ADS)

    Jia, X.; Dai, Y.

    2003-12-01

    Small punch (SP) tests were conducted in a temperature range from -190 to 80 °C on martensitic/ferritic steels F82H, T91 and Optimax-A irradiated in SINQ Target-3 up to 9.4 dpa in a irradiation temperature range of 90-275 °C. Results demonstrate: (a) the irradiation hardening deduced from SP tests is reasonably consistent with the results obtained by tensile tests; (b) with increasing irradiation dose, the SP yield load increases at all test temperatures, while the displacement at the maximum load and the total displacement at failure decrease; (c) the ductile-to-brittle transition temperature (DBTT SP) increases with increasing irradiation dose, and does so more quickly at irradiation doses above ˜6-7 dpa; in addition, the ΔDBTT SP increases linearly with helium content.

  5. 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.

  6. 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.

  7. In situ TEM Study of G-phase Precipitates under Heavy Ion Irradiation in CF8 Cast Austenitic Stainless Steel

    SciTech Connect

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

    2015-09-01

    Thermally-aged cast austenitic stainless steels (CASS) CF8 was irradiated with 1 MeV Kr ions at 300, 350 and 400°C to 1.88x10^15 ions/cm2 (~3 dpa) at the IVEM-Tandem Facility at the Argonne National Laboratory. Before irradiation, the distribution of G-phase precipitates in the ferrite showed strong spatial variations, and both their size and density were affected by the ferrite-austenite phase boundary and presence of M23C6 carbides. Under 300°C irradiation, in-situ TEM observation showed G-phase precipitates were relatively unchanged in the vicinity of the phase boundary M23C6 carbides, while the density of G-phase precipitates increased with increasing dose within the ferrite matrix. Coarsening of G-phase precipitates was observed in the vicinity of phase boundary M23C6 carbides at 350°C and 400°C.

  8. Suppression effect of nano-sized oxide particles on helium irradiation hardening in F82H-ODS steel

    NASA Astrophysics Data System (ADS)

    Chen, S.; Wang, Y.; Tadaki, K.; Hashimoto, N.; Ohnuki, S.

    2014-12-01

    Helium implantation was performed to investigate irradiation hardening in ferritic/martensitic steels. Depth dependence of nano-hardness was obtained using a Berkovich nano-indenter, and then nano-hardness was extracted from Nix-Gao model. The correlation between irradiation hardening and the concentration 500-2000 appm of helium was plotted. Nano-hardness increases as a function of helium concentration. F82H-ODS with a higher nano-hardness provides a lower irradiation hardening than F82H-IEA. Cross-sectional transmission electron microscopy (XTEM) revealed that cavities with a uniform distribution were formed after helium implantation at 2000 appm helium concentration, showing a mean size of 1.1 nm with an average number density of 4.9 × 1023 m-3 in F82H-IEA and 1.3 nm with 7.4 × 1023 m-3 in F82H-ODS. Orowan model was applied to evaluate the hardening from dispersed cavities. The significant difference of hardening between calculation and nano-indentation result of F82H-ODS indicates that oxide particles may shield the hardening effect from cavities because of the complex multi-interaction.

  9. Capability demonstration of simultaneous proton beam irradiation during exposure to molten lead-bismuth eutectic for HT9 steel

    NASA Astrophysics Data System (ADS)

    Qvist, Staffan; Bolind, Alan Michael; Hosemann, Peter; Wang, Yongqiang; Tesmer, Joseph; De Caro, Magdalena Serrano; Bourke, Mark

    2013-01-01

    We report the design and assembly of a corrosion station to enable simultaneous proton irradiation of a metallic surface that was also in contact with molten lead-bismuth eutectic (LBE). The capability has been established at the ion beam materials laboratory at Los Alamos National Laboratory (LANL). The engineering design focused on temperature and oxygen content control in the LBE, as well as the ability to achieve doses significantly in excess of 1 dpa in the contact region over the irradiation campaigns. In the preliminary demonstration of capability reported here, a sample made of HT9 steel was placed in contact with LBE at 450 °C and irradiated for 58 h at an average proton beam current of 0.3 μA/mm2. SRIM [1] calculations indicate that the nominal surface dose ranged from approximately 3-22 dpa. This paper outlines the experimental setup and design constraints. Characterization of the sample will be reported in a subsequent paper.

  10. 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.

  11. Effect of neutron irradiation on the microstructure of the stainless steel electroslag weld overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Onizawa, K.; Suzuki, M.

    2013-11-01

    Microstructural changes in the stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation with a fluence of 7.2 × 1023 n m-2 (E > 1 MeV) and a flux of 1.1 × 1017 n m-2 s-1 at 290 °C were investigated by atom probe tomography. The results showed a difference in the microstructural changes that result from neutron irradiation and thermal aging. Neutron irradiation resulted in the slight progression of Cr spinodal decomposition and an increase in the fluctuation of the Si, Ni, and Mn concentrations in the ferrite phases, with formation of γ‧-like clusters in the austenite phases. On the other hand, thermal aging resulted in the considerable progression of the Cr spinodal decomposition, formation of G-phases, and a decrease in the Si and an increase in the Ni and Mn concentration fluctuations at the matrix in the ferrite phases, without the microstructural changes in the austenite phases.

  12. Stress state dependence of transient irradiation creep in 20% cold worked 316 stainless steel

    NASA Astrophysics Data System (ADS)

    Foster, John Paul; Bunde, Kermit; Gilbert, E. Robert

    1998-11-01

    Irradiation creep tests were performed in fast reactors using the stress states of uniaxial tension, biaxial tension, bending and torsion. In order to compare the saturated transient strain irradiation creep component, the test data were converted to equivalent strain and equivalent stress. The saturated transient irradiation creep component was observed to depend on the stress state. The highest value was exhibited by the uniaxial tension stress state, and the lowest by the torsion stress state. The biaxial tension and bending stress state transient component values were intermediate. This behavior appears to be related to the dislocation or microscopic substructure resulting from fabrication processing and the applied stress direction.

  13. AEM and AES of radiation-induced segregation in proton-irradiated stainless steels

    SciTech Connect

    Kenik, E.A.; Carter, R.D.; Damcott, D.L.; Atzmon, M.; Was, G.S.

    1994-06-01

    In order to avoid complications from long-term induced radioactivity of neutron-irradiated specimens, 4 type 304L alloys were irradiated to 1 dpa with 3.4 MeV protons at 400 C. Analytical electron microscopy and Auger electron spectrometry were used to measure composition at and near grain boundaries in controlled purity alloys. As a result of the narrow RIS profiles (<20 nm width) at grain boundaries induced in these materials by low temperature irradiation and the finite size of the excited volume for x-ray microanalysis, the measured profiles are convolutions of these two factors.

  14. 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

  15. 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.

  16. Fatigue behavior at 650/sup 0/C of 20%-cold-worked type 316 stainless steel irradiated at 550/sup 0/C in the HFIR

    SciTech Connect

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

    1984-01-01

    Type 316 stainless steel in the 20%-cold-worked condition was irradiated in the High Flux Isotope Reactor (HFIR) and subsequently tested in fatigue. The specimens were irradiated at 550/sup 0/C to damage levels of 8 to 12 dpa and transmutation helium levels of 300 to 500 at. ppM. Fatigue testing at 650/sup 0/C revealed that cyclic life was not significantly affected by the irradiation. However, unlike the results of tests of the same material at 550/sup 0/C, no endurance limit was observed. The absence of an endurance limit is interpreted in terms of thermal creep.

  17. Detection of Irradiation Effects on Reactor Vessel Steels by Magneto-Acoustic Emission.

    DTIC Science & Technology

    1988-04-21

    espe- mechanisms of neutron irradiation damage of alpha - iron cially in low nickel compositions. Both MAE and Bark- [10]. It showed that MAE responses...can be correlated with hausen waveforms showed double peaks during each half- the metallurgical conditions of alpha - iron . MAE can provi cycle of...Briggs, A Study of Neutron Irradiation Damage in I. MAE and SBN measurements and waveform analysis Alpha - Iron Using Magneto-Acoustic Emission, AERE

  18. 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)

  19. 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.

  20. 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.

  1. Properties of High-Frequency Sub-Wavelength Ripples on Stainless Steel 304L under Ultra Short Pulse Laser Irradiation

    NASA Astrophysics Data System (ADS)

    Mitko, V. S.; Römer, G. R. B. E.; Veld, A. J. Huis in `t.; Skolski, J. Z. P.; Obona, J. V.; Ocelík, V.; De Hosson, J. T. M.

    The paper concentrates on surface texturing on sub-micro meter scale with ultra short laser pulses that has several applications, e.g. changing the hydrophilic/hydrophobic performance, optical or tribological properties of materials. In general, the formations of wavy structures, or ripples on a surface irradiated by short pulse lasers has been observed experimentally since 1965, and are usually referred to as Laser Induced Periodic Surface Structures (LIPSS). Generally Low Spatial Frequency LIPSS (LSFL) and High Spatial Frequency LIPSS (HSFL) are observed. The existing theoretical models do not describe the origin, nor growth of the ripples satisfactorily. That is why the experimental approach still plays a leading role in the investigation of ripple formation. In this paper we study the development of HSFL and LSFL as a result of picosecond laser pulses on a surface of stainless steel. Influences of number of pulses and pulse overlap on ripples growth were examined.

  2. Neutron irradiation and high temperature effects on amorphous Fe-based nano-coatings on steel - A macroscopic assessment

    NASA Astrophysics Data System (ADS)

    Simos, N.; Zhong, Z.; Dooryhee, E.; Ghose, S.; Gill, S.; Camino, F.; Şavklıyıldız, İ.; Akdoğan, E. K.

    2017-06-01

    The study revealed that loss of ductility in an amorphous Fe-alloy coating on a steel substrate composite structure was essentially prevented from occurring, following radiation with modest neutron doses of ∼2 × 1018 n/cm2. At the higher neutron dose of ∼2 × 1019, macroscopic stress-strain analysis showed that the amorphous Fe-alloy nanostructured coating, while still amorphous, experienced radiation-induced embrittlement, no longer offering protection against ductility loss in the coating-substrate composite structure. Neutron irradiation in a corrosive environment revealed exemplary oxidation/corrosion resistance of the amorphous Fe-alloy coating, which is attributed to the formation of the Fe2B phase in the coating. To establish the impact of elevated temperatures on the amorphous-to-crystalline transition in the amorphous Fe-alloy, electron microscopy was carried out which confirmed the radiation-induced suppression of crystallization in the amorphous Fe-alloy nanostructured coating.

  3. 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.

  4. On the correlation between irradiation-induced microstructural features and the hardening of reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Lambrecht, M.; Meslin, E.; Malerba, L.; Hernández-Mayoral, M.; Bergner, F.; Pareige, P.; Radiguet, B.; Almazouzi, A.

    2010-11-01

    A correlation is attempted between microstructural observations by various complementary techniques, which have been implemented within the PERFECT project and the hardening measured by tensile tests of reactor pressure vessel steel and model alloys after irradiation to a dose of ˜7 × 10 19 n cm -2. This is done, using the simple hardening model embodied by the Orowan equation and applying the most suitable superposition law, as suggested by a parametric study using the DUPAIR line tension code. It is found that loops are very strong obstacles to dislocation motion, but due to their low concentration, they only play a minor role in the hardening itself. For the precipitates, the contrary is found, although they are quite soft (due to their very small sizes and their coherent nature), they still play the dominant role in the hardening. Vacancy clusters are important for the formation of both loops and precipitates, but they will play almost no role in the hardening by themselves.

  5. 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.

  6. Effect of irradiation temperature on microstructure of ferritic-martensitic ODS steel

    NASA Astrophysics Data System (ADS)

    Klimenkov, M.; Lindau, R.; Jäntsch, U.; Möslang, A.

    2017-09-01

    The EUROFER-ODS alloy with 0.5% Y2O3 was neutron irradiated with doses up to 16.2 dpa at 250 °C, 350 °C and 450 °C. The radiation induced changes in the microstructure (e.g. dislocation loops and voids) were investigated using transmission electron microscopy (TEM). The number density of radiation induced defects was found to be significantly lower than in EUROFER 97 irradiated at the same conditions. It was found that the appearance and extent of radiation damage strongly depend not only on the irradiation temperature but also on the local number density and size distribution of ODS particles. The higher number density of dislocation loops and voids was found in the local areas with low number density of ODS particles. The interstitial loops with Burgers vector of both ½<111> and <100> types were detected by imaging using different diffraction conditions.

  7. 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.

  8. 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.

  9. Field emission study from an array of hierarchical micro protrusions on stainless steel surface generated by femtosecond pulsed laser irradiation

    NASA Astrophysics Data System (ADS)

    Singh, A. K.; Suryawanshi, Sachin R.; More, M. A.; Basu, S.; Sinha, Sucharita

    2017-02-01

    This paper reports our results on femtosecond (fs) pulsed laser induced surface micro/nano structuring of stainless steel 304 (SS 304) samples and their characterization in terms of surface morphology, formed material phases on laser irradiation and field emission studies. Our investigations reveal that nearly uniform and dense array of hierarchical micro-protrusions (density: ∼5.6 × 105 protrusions/cm2) is formed upon laser treatment. Typical tip diameters of the generated protrusions are in the range of 2-5 μm and these protrusions are covered with submicron sized features. Grazing incidence X-ray diffraction (GIXRD) analysis of the laser irradiated sample surface has shown formation mainly of iron oxides and cementite (Fe3C) phases in the treated region. These laser micro-structured samples have shown good field emission properties such as low turn on field (∼4.1 V/μm), high macroscopic field enhancement factor (1830) and stable field emission current under ultra high vacuum conditions.

  10. Heavy-Section Steel Irradiation (HSSI) Program (W6953) Monthly Letter Status Report - February 2001 - ORNL/HSSI (6953) MLSR-2001/5.

    SciTech Connect

    Rosseel, T.M.

    2001-03-26

    The primary goal of the Heavy-Section Steel Irradiation (HSSI) Program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure vessel (RPV) integrity. The program includes studies of the effects of irradiation on the degradation of mechanical and fracture properties of vessel materials augmented by enhanced 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 post-irradiation mitigation are being examined on a wide range of fracture properties. This program will also maintain and upgrade computerized databases, calculational procedures, and standards relating to RPV fluence-spectra determinations and embrittlement assessments. Results from the HSSI studies will be incorporated into codes and standards directly applicable to resolving major regulatory issues that involve RPV irradiation embrittlement such as pressurized-thermal shock, operating pressure-temperature limits, low-temperature overpressurization, and the specialized problems associated with low upper-shelf welds. Six technical tasks and one for program management are now contained in the HSSI Program.

  11. Cavity nucleation and growth in dual beam irradiated 316L industrial austenitic stainless steel

    NASA Astrophysics Data System (ADS)

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

    2017-10-01

    Thin foils of 316L were simultaneously ion irradiated and He implanted in situ in a Transmission Electron Microscope at elevated temperatures. The resulting microstructure is carefully investigated in comparison with previous single ion irradiation experiments with a focus on the nucleation and growth of cavities. Helium is found to strongly enhance the nucleation of cavities in dual beam experiments. On the contrary, it does not induce more nucleation when implanted consecutively to an in situ ion irradiation but rather the growth of cavities by absorption at existing cavities, which shows the importance of synergistic effects and He injection mode on the microstructural changes. In both dual beam and single beam experiments, the characteristics of the populations of cavities, either stabilized by He or O atoms, are in qualitative agreement with the predictions of rate theory models for cavity growth. The evolutions of cavity population as a function of irradiation conditions can be reasonably well explained by the concept of relative sink strength of cavities and dislocations and the resulting partitioning of defects at sinks, or conversely recombination when either of the sinks dominates. The dislocations whose presence is a prerequisite to cavity growth in rate theory models are not observed in all studied conditions. In this case, the net influx of vacancies to cavities necessary to their growth and conversion to voids is believed to result from free surface effects, and possibly also segregation of elements close to the cavity surface. In any studied condition, the measured swelling is low, which is ascribed to the dilution of gaseous atoms among a high density of cavities as well as a high rate of point defect recombination and loss at traps. This high rate of recombination enhanced when dislocations are absent appears to result in the formation of overpressurized He bubbles.

  12. 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

  13. Effect of irradiation defects on the corrosion behaviors of steels exposed to lead bismuth eutectic in ADS: a first-principles study.

    PubMed

    Zhang, Yange; You, Yu-Wei; Li, Dong-Dong; Xu, Yichun; Liu, C S; Pan, B C; Wang, Zhiguang

    2015-05-14

    In accelerator driven systems (ADSs), steels will suffer not only from the irradiation damage produced by protons or neutrons, but also from the dissolution corrosion induced by the liquid lead-bismuth eutectic (LBE). In this work we investigate the interactions between LBE atoms (Pb, Bi) and the irradiation induced defects X (X is helium, vacancy or divacancy) in α-Fe based on first-principles calculations. It is found that LBE atoms repulse each other without irradiation defects, while they aggregate easily with the defects to form X-Pbn and X-Bin complexes. This indicates that the irradiation defects could promote the aggregation of LBE atoms in iron, especially Bi atoms. The total binding energies of the X-Pbn and X-Bin complexes increase with the number of Pb and Bi atoms, respectively. The origin of the total binding energies of the complexes is further discussed via the electronic structures and the distortion of the crystalline lattice. Finally, the concentration evolutions of the Vac-(Bi)n complexes and unbound vacancies with temperature are predicted by the mass action analysis. This work provides important information for the synergistic effect of irradiation and LBE corrosion on the steels in the ADSs, which can be used as basic parameters for further study.

  14. Physical and mechanical modelling of neutron irradiation effect on ductile fracture. Part 1. Prediction of fracture strain and fracture toughness of austenitic steels

    NASA Astrophysics Data System (ADS)

    Margolin, Boris; Sorokin, Alexander; Smirnov, Valeriy; Potapova, Vera

    2014-09-01

    A physical-and-mechanical model of ductile fracture has been developed to predict fracture toughness and fracture strain of irradiated austenitic steels taking into account stress-state triaxiality and radiation swelling. The model is based on criterion of plastic collapse of a material unit cell controlled by strain hardening of a material and criterion of voids coalescence due to channel shearing of voids. The model takes into account deformation voids nucleation and growth of deformation and vacancy voids. For justification of the model experimental data on fracture strain and fracture toughness of austenitic steel 18Cr-10Ni-Ti grade irradiated up to maximal dose 150 dpa with various swelling were used. Experimental data on fracture strain and fracture toughness were compared with the results predicted by the model. It has been shown that for prediction of the swelling effect on fracture toughness the dependence of process zone size on swelling should be taken into account.

  15. Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Bergner, F.; Gillemot, F.; Hernández-Mayoral, M.; Serrano, M.; Török, G.; Ulbricht, A.; Altstadt, E.

    2015-06-01

    Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.

  16. 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.

  17. Effect of neutron irradiation on the impact properties of A533B steel

    SciTech Connect

    Schubert, L.E.; Kumar, A.S.; Rosinski, S.T.; Hamilton, M.L.

    1994-10-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, ASTM type A 533 Grade B (A533B) having a low USE (USE < 100 J). The methodology appears to be more satisfactory than those methodologies proposed earlier. The USE was normalized by a normalization factor involving the dimensions of the Charpy specimen, the elastic stress concentration factor, and the plastic constraint at the notch root. The normalized values of the USE were found to be invariant with specimen size. In addition, it was also found that the ratio of the USE of unirradiated to that of irradiated materials was approximately the same for full, half, and third size specimens. The ductile-to-brittle transition temperature (DBTT) increased due to irradiation at 150 C to a nominal fluence of 1.0 {times} 10{sup 19} n/cm{sup 2} (E > 1 MeV) by 78 {degree}, 83{degree}, and 70{degree}C for full, half, and third size specimens, respectively. These shifts in DBTT appeared to be independent of specimen size and notch geometry.

  18. Effect of specimen size on the impact properties of neutron irradiated A533B steel

    NASA Astrophysics Data System (ADS)

    Schubert, L. E.; Kumar, A. S.; Rosinki, Stan T.; Hamilton, Margaret L.

    1995-08-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full-size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material. The methodology appears to be more satisfactory than those methodlogies proposed earlier. The USE was normalized by a normalization factor involving the dimensions of the Charpy specimen, the elastic stress concentration factor, and the plastic constraint at the notch root. The normalized values of the USE were found to be invariant with speciment size. In addition, it was also found that the ratio of the USE of unirradiated to that of irradiated materials was approximately the same for full-, half-, and third-size specimens. The ductile-to-brittle transition temperture (DBTT) increased due to irradiation at 150°C to a nominal fluence of 1.0 × 10 19 n/cm 2 ( E >MeV) by 78, 83 and 70°C for full-, half-, and third-size specimens, respectively. These shifts in DBTT appeared to be independent of specimen size and notch geometry.

  19. Isochronal annealing studies on 1.1 MeV Fe ion irradiated RAFM steel using variable energy slow positron beam

    NASA Astrophysics Data System (ADS)

    Ramachandran, Renjith; David, C.; Rajaraman, R.; Abhaya, S.; Panigrahi, B. K.; Amarendra, G.

    2017-05-01

    Indian Reduced Activation Ferritic Martensitic steel is irradiated with 1.1 MeV Fe ions to a dose of 0.1 dpa at room temperature. The positron annihilation study showed a decrease in S-parameter with annealing temperature due to vacancy annealing. A complete defect recovery is observed beyond 1073 K. The linear nature of (S, W) correlation plot shows that only one kind of defect is present throughout the annealing temperature.

  20. Enhanced Radiation-tolerant Oxide Dispersion Strengthened Steel and its Microstructure Evolution under Helium-implantation and Heavy-ion Irradiation

    NASA Astrophysics Data System (ADS)

    Lu, Chenyang; Lu, Zheng; Wang, Xu; Xie, Rui; Li, Zhengyuan; Higgins, Michael; Liu, Chunming; Gao, Fei; Wang, Lumin

    2017-01-01

    The world eagerly needs cleanly-generated electricity in the future. Fusion reactor is one of the most ideal energy resources to defeat the environmental degradation caused by the consumption of traditional fossil energy. To meet the design requirements of fusion reactor, the development of the structural materials which can sustain the elevated temperature, high helium concentration and extreme radiation environments is the biggest challenge for the entire material society. Oxide dispersion strengthened steel is one of the most popular candidate materials for the first wall/blanket applications in fusion reactor. In this paper, we evaluate the radiation tolerance of a 9Cr ODS steel developed in China. Compared with Ferritic/Martensitic steel, this ODS steel demonstrated a significantly higher swelling resistance under ion irradiation at 460 °C to 188 displacements per atom. The role of oxides and grain boundaries on void swelling has been explored. The results indicated that the distribution of higher density and finer size of nano oxides will lead a better swelling resistance for ODS alloy. The original pyrochlore-structured Y2Ti2O7 particles dissolved gradually while fine Y-Ti-O nano clusters reprecipitated in the matrix during irradiation. The enhanced radiation tolerance is attributed to the reduced oxide size and the increased oxide density.

  1. Enhanced Radiation-tolerant Oxide Dispersion Strengthened Steel and its Microstructure Evolution under Helium-implantation and Heavy-ion Irradiation

    PubMed Central

    Lu, Chenyang; Lu, Zheng; Wang, Xu; Xie, Rui; Li, Zhengyuan; Higgins, Michael; Liu, Chunming; Gao, Fei; Wang, Lumin

    2017-01-01

    The world eagerly needs cleanly-generated electricity in the future. Fusion reactor is one of the most ideal energy resources to defeat the environmental degradation caused by the consumption of traditional fossil energy. To meet the design requirements of fusion reactor, the development of the structural materials which can sustain the elevated temperature, high helium concentration and extreme radiation environments is the biggest challenge for the entire material society. Oxide dispersion strengthened steel is one of the most popular candidate materials for the first wall/blanket applications in fusion reactor. In this paper, we evaluate the radiation tolerance of a 9Cr ODS steel developed in China. Compared with Ferritic/Martensitic steel, this ODS steel demonstrated a significantly higher swelling resistance under ion irradiation at 460 °C to 188 displacements per atom. The role of oxides and grain boundaries on void swelling has been explored. The results indicated that the distribution of higher density and finer size of nano oxides will lead a better swelling resistance for ODS alloy. The original pyrochlore-structured Y2Ti2O7 particles dissolved gradually while fine Y-Ti-O nano clusters reprecipitated in the matrix during irradiation. The enhanced radiation tolerance is attributed to the reduced oxide size and the increased oxide density. PMID:28079191

  2. Enhanced Radiation-tolerant Oxide Dispersion Strengthened Steel and its Microstructure Evolution under Helium-implantation and Heavy-ion Irradiation.

    PubMed

    Lu, Chenyang; Lu, Zheng; Wang, Xu; Xie, Rui; Li, Zhengyuan; Higgins, Michael; Liu, Chunming; Gao, Fei; Wang, Lumin

    2017-01-12

    The world eagerly needs cleanly-generated electricity in the future. Fusion reactor is one of the most ideal energy resources to defeat the environmental degradation caused by the consumption of traditional fossil energy. To meet the design requirements of fusion reactor, the development of the structural materials which can sustain the elevated temperature, high helium concentration and extreme radiation environments is the biggest challenge for the entire material society. Oxide dispersion strengthened steel is one of the most popular candidate materials for the first wall/blanket applications in fusion reactor. In this paper, we evaluate the radiation tolerance of a 9Cr ODS steel developed in China. Compared with Ferritic/Martensitic steel, this ODS steel demonstrated a significantly higher swelling resistance under ion irradiation at 460 °C to 188 displacements per atom. The role of oxides and grain boundaries on void swelling has been explored. The results indicated that the distribution of higher density and finer size of nano oxides will lead a better swelling resistance for ODS alloy. The original pyrochlore-structured Y2Ti2O7 particles dissolved gradually while fine Y-Ti-O nano clusters reprecipitated in the matrix during irradiation. The enhanced radiation tolerance is attributed to the reduced oxide size and the increased oxide density.

  3. New method for detection of Li inside He bubbles formed in B10-alloyed steel after neutron irradiation.

    PubMed

    Klimenkov, M; Möslang, A; Materna-Morris, E

    2013-03-01

    Electron energy loss spectroscopy (EELS) was used to detect and study the spatial distribution on the nanoscale of He and Li in boron-alloyed steel after neutron irradiation. Li and He are the products of the (10)B(n, α)(7)Li nuclear transmutation reaction and knowledge of their distribution is important to understand their influence on mechanical properties. Here, a new method is presented for the direct detection of Li in Fe, which is based on the analysis of the plasmon structure in EELS spectra. Li drops or particles in He bubbles show pronounced Li plasmon line at 10eV which can be extracted from the Fe/Cr plasmon. The Gaussian or linear interpolation of the Fe/Cr plasmon and its subtraction allows for the calculation of Li and He two-dimensional maps and the study their spatial distribution. The analysis of Li plasmon fine structure allows imaging surface effects in the Li drops. Copyright © 2012 Elsevier Ltd. All rights reserved.

  4. Neutron irradiation and high temperature effects on amorphous Fe-based nano-coatings on steel – A macroscopic assessment

    DOE PAGES

    Simos, N.; Zhong, Z.; Dooryhee, E.; ...

    2017-03-23

    Here, this study revealed that loss of ductility in an amorphous Fe-alloy coating on a steel substrate composite structure was essentially prevented from occurring, following radiation with modest neutron doses of ~2 x 1018 n/cm2. At the higher neutron dose of ~2 x 1019, macroscopic stress-strain analysis showed that the amorphous Fe-alloy nanostructured coating, while still amorphous, experienced radiation-induced embrittlement, no longer offering protection against ductility loss in the coating-substrate composite structure. Neutron irradiation in a corrosive environment revealed exemplary oxidation/corrosion resistance of the amorphous Fe-alloy coating, which is attributed to the formation of the Fe2B phase in the coating.more » To establish the impact of elevated temperatures on the amorphous-to-crystalline transition in the amorphous Fe-alloy, electron microscopy was carried out which confirmed the radiation-induced suppression of crystallization in the amorphous Fe-alloy nanostructured coating.« less

  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. Stainless steel wire mesh-supported ZnO for the catalytic photodegradation of methylene blue under ultraviolet irradiation.

    PubMed

    Vu, Tan T; del Río, Laura; Valdés-Solís, Teresa; Marbán, Gregorio

    2013-02-15

    The aim of this study was to assess the activity of catalysts formed by nanostructured zinc oxide supported on stainless steel wire mesh for the photocatalytic degradation of methylene blue under UV irradiation. Catalysts prepared by means of different low temperature synthesis methods, as described in a previous work (Vu et al., Mater. Res. Bull. 47 (2012) 1577-1586) were tested. A new activity parameter was introduced in order to compare the catalytic activity of the different catalysts. The best catalyst showed a catalytic activity higher than that of the reference material TiO(2) P25 (Degussa-Evonik). This high activity is attributed to a higher quantum yield derived from the small particle length of the ZnO deposited on the wire mesh. The photocatalytic degradation kinetics of methylene blue fitted a potential model with n orders ranging from 0.5 to 6.9. Reaction orders over 1 were attributed to catalyst deactivation during the reaction resulting from the photocorrosion of ZnO. Copyright © 2012 Elsevier B.V. All rights reserved.

  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. Microstructure evolution of two model ferritic/martensitic steels under in situ ion irradiation at low doses (0-2 dpa)

    NASA Astrophysics Data System (ADS)

    Kaoumi, D.; Adamson, J.; Kirk, M.

    2014-02-01

    Ferritic/martensitic steels are candidate materials for structural and cladding components designed for Generation IV reactors because of their superior resistance to radiation damage at the high operating temperatures envisioned in these reactors. To enable the development and optimization of such advanced alloys for in-reactor use, a fundamental understanding of radiation damage accumulation in materials is required. In this work, two model F/M steels (12Cr model alloy and 9Cr model alloy) were irradiated with 1 MeV Kr ions at 50 K, 180 K, 298 K, 473 K and 573 K in situ in a TEM. The microstructure evolution under irradiation was followed and characterized at successive doses in terms of irradiation-induced defect formation and evolution, defect density, size distribution and interaction with the as-fabricated microstructure (e.g. dislocation networks, lath boundaries) using weak-beam dark-field imaging. The effect of the irradiation temperature on the defect kinetics is assessed at doses up to 2 dpa.

  9. 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

  10. 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.

  11. 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.

  12. Microstructural investigation, using small-angle neutron scattering (SANS), of Optifer steel after low dose neutron irradiation and subsequent high temperature tempering

    NASA Astrophysics Data System (ADS)

    Coppola, R.; Lindau, R.; Magnani, M.; May, R. P.; Möslang, A.; Valli, M.

    2007-08-01

    The microstructural effect of low dose neutron irradiation and subsequent high temperature tempering in the reduced activation ferritic/martensitic steel Optifer (9.3 Cr, 0.1 C, 0.50 Mn, 0.26 V, 0.96 W, 0.66 Ta, Fe bal wt%) has been studied using small-angle neutron scattering (SANS). The investigated Optifer samples had been neutron irradiated, at 250 °C, to dose levels of 0.8 dpa and 2.4 dpa. Some of them underwent 2 h tempering at 770 °C after the irradiation. The SANS measurements were carried out at the D22 instrument of the High Flux Reactor at the Institut Max von Laue - Paul Langevin, Grenoble, France. The differences observed in nuclear and magnetic SANS cross-sections after subtraction of the reference sample from the irradiated one suggest that the irradiation and the subsequent post-irradiation tempering produce the growth of non-magnetic defects, tentatively identified as microvoids.

  13. Post-irradiation annealing behavior of microstructure and hardening of a reactor pressure vessel steel studied by positron annihilation and atom probe tomography

    NASA Astrophysics Data System (ADS)

    Kuramoto, A.; Toyama, T.; Takeuchi, T.; Nagai, Y.; Hasegawa, M.; Yoshiie, T.; Nishiyama, Y.

    2012-06-01

    Post-irradiation annealing (PIA) behavior of irradiation-induced microstructural changes and hardening of an A533B (0.16 wt.% Cu) steel after neutron-irradiation of 3.9 × 1019 n cm-2 (0.061 displacement per atom (dpa)) at 290 °C was studied by positron annihilation spectroscopy (PAS), atom probe tomography (APT) and Vickers microhardness measurements. Coincidence Doppler broadening and positron lifetime measurements clearly reveal two recovery stages; (i) as-irradiated state to annealing at 450 °C and (ii) annealing from 450 to 600 °C. The first stage is due to annealing out of the most of irradiation-induced vacancy-related defects, while the second stage corresponds to dissolving of irradiation-induced solute nanoclusters (SCs). APT observations reveal that the SCs are enriched with Cu, Mn, Ni and Si and that their number densities decrease with increasing annealing temperature without coarsening to give almost complete recovery at 550 °C. The experimental hardening is almost twice the SC hardening estimated by the Russell-Brown model below 350 °C, whereas it is almost the same as that estimated in the range 400-550 °C.

  14. Effect of the bainitic and martensitic microstructures on the hardening and embrittlement under neutron irradiation of a reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Marini, B.; Averty, X.; Wident, P.; Forget, P.; Barcelo, F.

    2015-10-01

    The hardening and the embrittlement under neutron irradiation of an A508 type RPV steel considering three different microstructures (bainite, bainite-martensite and martensite)have been investigated These microstructures were obtained by quenching after autenitization at 1100 °C. The irradiation induced hardening appears to depend on microstructure and is correlated to the yield stress before irradiation. The irradiation induced embrittlement shows a more complex dependence. Martensite bearing microstructures are more sensitive to non hardening embrittlement than pure bainite. This enhanced sensitivity is associated with the development of intergranular brittle facture after irradiation; the pure martensite being more affected than the bainite-martensite. It is of interest to note that this mixed microstructure appears to be more embrittled than the pure bainitic or martensitic phases in terms of temperature transition shift. This behaviour which could emerge from the synergy of the embrittlement mechanisms of the two phases needs further investigations. However, the role of microstructure on brittle intergranular fracture development appears to be qualitatively similar under neutron irradiation and thermal ageing.

  15. Overview of the activities in Spain on irradiation embrittlement of RPV steel

    SciTech Connect

    Bros, J.; Ballesteros, A.

    1994-12-31

    Specific problems such as the embrittlement of the VVER reactors or the integrity evaluation of the Yankee Rowe reactor pressure vessel, show the necessity of continuing in the understanding of the mechanisms responsible for the embrittlement. Besides, in order to increase the safety and performance of the nuclear power plants, it is important to optimize the surveillance programs and develop the mitigation techniques of the damage caused by neutron radiation. The main activities carried out in Spain are related to the surveillance programs and to phase 3 of the IAEA Coordinated Research Program, although other smaller projects also show the interest of the utilities in the embrittlement issues. The extension of the work performed as well as some of the most relevant results are shown in this paper. Some new projects are being considered. The participation in the recently created European Action Group for RPV Materials Irradiation Effects and Studies (AMES) is of huge value for Spain. This Group tries mainly to give a European response in the field of the standards. In spite of the nuclear moratorium that has been in vigor in Spain for several years and the present economic crisis, there is a general interest in participation in international projects. That is the case of the Spanish contribution to the certification of Advanced Light-Water Reactors. In particular, regarding the embrittlement problem, in the development of a reactor pressure vessel and support skirt material surveillance program for the Simplified Boiling Water Reactor (SBWR).

  16. Effect of heat treatment and irradiation temperature on mechanical properties and structure of reduced-activation Cr-W-V steels of bainitic, martensitic, and martensitic-ferritic classes

    NASA Astrophysics Data System (ADS)

    Gorynin, I. V.; Rybin, V. V.; Kursevich, I. P.; Lapin, A. N.; Nesterova, E. V.; Klepikov, E. Yu

    2000-12-01

    Effects of molybdenum replacement by tungsten in steels of the bainitic, martensitic, and martensitic-ferritic classes containing 2.5%, 8% and 11% Cr, respectively, were investigated. The phase composition and structure of the bainitic steels were varied by changing the cooling rates from the austenitization temperature (from values typical for normalization up to V=3.3 × 10-2°C/s) and then tempering. The steels were irradiated to a fluence of 4×1023 n/m2 (⩾0.5 MeV) at 270°C and to fluences of 1.3×1023 and 1.2×1024 n/m2 (⩾0.5 MeV) at 70°C. The 2.5Cr-1.4WV and 8Cr-1.5WV steels have shown lower values of the shifts in ductile-brittle transition temperature (DBTT) under irradiation in comparison with corresponding Cr-Mo steels. Radiation embrittlement at elevated irradiation temperature was lowest in bainitic 2.5Cr-1.4WV steel and martensitic-ferritic 11Cr-1.5WV steel. The positive effect of molybdenum replacement by tungsten at irradiation temperature ∼300°C is reversed at Tirr=70∘C.

  17. Comparison of the effects of long-term thermal aging and HFIR irradiation on the microstructural evolution of 9Cr-1MoVNb steel

    SciTech Connect

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

    1990-01-01

    Both thermal aging at 482--704{degree}C for up to 25,000h and HFIR irradiation at 300--600{degree}C for up to 39 dpa produce substantial changes in the as-tempered microstructure of 9Cr-1MoVNb martensitic/ferritic steel. However, the changes in the dislocation/subgrain boundary and the precipitate structures caused by thermal aging or neutron irradiation are quite different in nature. During thermal aging, the as-tempered lath/subgrain boundary and carbide precipitate structures remain stable below 650{degree}C, but coarsen and recover somewhat at 650--704{degree}C. The formation of abundant intergranular Laves phase, intra-lath dislocation networks, and fine dispersions of VC needles are thermal aging effects that are superimposed upon the as-tempered microstructure at 482--593{degree}C. HFIR irradiation produces dense dispersions of very small black-dot'' dislocations loops at 300{degree}C and produces helium bubbles and voids at 400{degree}C At 300--500{degree}C, there is considerable recovery of the as-tempered lath/subgrain boundary structure and microstructural/microcompositional instability of the as-tempered carbide precipitates during irradiation. By contrast, the as-tempered microstructure remains essentially unchanged during irradiation at 600{degree}C. Comparison of thermally aged with irradiation material suggests that the instabilities of the as-tempered lath/subgrain boundary and precipitate structures at lower irradiation temperatures are radiation-induced effects, whereas the absence of both Laves phase and fine VC needles during irradiation is a radiation-retarded thermal effect.

  18. The strong influence of displacement rate on void swelling in variants of Fe-16Cr-15Ni-3Mo austenitic stainless steel irradiated in BN-350 and BOR-60

    NASA Astrophysics Data System (ADS)

    Budylkin, N. I.; Bulanova, T. M.; Mironova, E. G.; Mitrofanova, N. M.; Porollo, S. I.; Chernov, V. M.; Shamardin, V. K.; Garner, F. A.

    2004-08-01

    Recent irradiation experiments conducted on a variety of austenitic stainless steels have shown that void swelling appears to be increased when the dpa rate is decreased, primarily by a shortening of the transient regime of swelling. This paper presents results derived from nominally similar irradiations conducted on six Russian steels, all laboratory heat variants of Fe-16Cr-15Ni-3Mo-Nb-B, with each irradiated in two fast reactors, BOR-60 and BN-350. The BN-350 irradiation proceeded at a dpa rate three times higher than that conducted in BOR-60. In all six steels, a significantly higher swelling level was attained in BOR-60, agreeing with the results of earlier studies.

  19. “Measurement of void swelling in thick non-uniformly irradiated 304 stainless steel blocks using nondestructive ultrasonic techniques”

    SciTech Connect

    F. A. Garner; T. Okita; Y. Isobe; M. Sagisaki; J. Etoh; T. Matsunaga; P. D. Freyer; Y. Huang; J. M. K. Wiezorek; D. L. Porter

    2001-09-01

    Void swelling is of potential importance in PWR austenitic internals, especially in components that will see higher doses during plant lives beyond 40 years. Proactive surveillance of void swelling is required to identify its emergence before swelling reaches levels that cause high levels of embrittlement and distortion. Non-destructive measurements of ultrasonic velocity can measure swelling at fractions of a percent. To demonstrate the feasibility of this technique for PWR application we have investigated five blocks of 304 stainless steel that were irradiated in the EBR-II fast reactor. These blocks were of hexagonal cross-section, with thickness of ~50 mm and lengths of ~218-245 mm. They were subjected to significant axial and radial gradients in gamma heating, temperature and dpa rate, producing complex internal distributions of swelling, reaching ~3.5% maximum at an off-center mid-core position. Swelling decreases both the density and elastic modulii, thereby impacting the ultrasonic velocity. Concurrently, carbide precipitates form, producing increases in density and decreases in elastic modulii. Using blocks from both low and high dpa levels it was possible to separate the ultrasonic contributions of voids and carbides. Time-of-flight ultrasonic measurements were used to non-destructively measure the internal distribution of void swelling. These distributions were confirmed using non-destructive profilometry followed by destructive cutting to provide density change and electron microscopy data. It was demonstrated that the four measurement types produce remarkably consistent results. Therefore ultrasonic measurements offer great promise for in-situ surveillance of voids in PWR core internals.

  20. Effect of gadolinium nitrate concentration on molecular product yield during gamma irradiation and on corrosion of stainless steel

    NASA Astrophysics Data System (ADS)

    Mal, D.; Puspalata, R.; Rangarajan, S.; Velmurugan, S.

    2017-09-01

    Effect of high concentrations of soluble neutron poison gadolinium nitrate, Gd(NO3)3, in the moderator system of a proposed advanced Indian nuclear reactor, was evaluated from the safety point of view. The radiolytic yields of H2 and H2O2 was expected to be high as moderator water system pH would be lowered and conductivity also would be high by the addition of higher concentration Gd(NO3)3 solutions during various shutdown states. Experiments were carried out to estimate this increase in radiolytic yield of molecular products with the addition of Gd(NO3)3 in the concentration range of 15-400 mg kg-1. Both the H2O2 and H2 yields were found to increase with absorbed dose and also with increasing Gd3+ concentration up to 100 mg kg-1 but the increase were marginal in 100-400 mg kg-1 range. For a given concentration of Gd(NO3)3 solution, radiolysis in high purity D2O showed a lower D2 formation than H2 in light water. In a simulated moderator temperature of 65 °C, a higher yield of H2 was observed. The headspace provided above the liquid phase in irradiation zone had shown to have a substantial effect on the generation of H2. With decreasing headspace, H2 generation increased and went through a maximum. Considering the expected long operational life ( 100 years) for the proposed reactor, the corrosion rate of the structural materials (stainless steel 304 LN) in contact with this high concentration Gd(NO3)3 solution was also estimated at 65 °C which showed a negligible effect.

  1. Radiation hardening and -embrittlement due to He production in F82H steel irradiated at 250 °C in JMTR

    NASA Astrophysics Data System (ADS)

    Wakai, E.; Jitsukawa, S.; Tomita, H.; Furuya, K.; Sato, M.; Oka, K.; Tanaka, T.; Takada, F.; Yamamoto, T.; Kato, Y.; Tayama, Y.; Shiba, K.; Ohnuki, S.

    2005-08-01

    The dependence of helium production on radiation hardening and -embrittlement has been examined in a reduced-activation martensitic F82H steel (8Cr-2W-0.2V-0.04Ta-0.1C) irradiated at 250 °C to 2.3 dpa. In this study, 10B and 11B-doped specimens were irradiated to minimize the errors from the effect of B on mechanical properties by comparing the results. The specimens used were 10B-doped, 10B + 11B-doped and 11B-doped F82H steels. The total amounts of doping boron were about 60 mass ppm. The range of helium concentration produced in the specimens was from about 5 to about 330 appm. Tensile and fracture toughness tests were performed after neutron irradiation. 50 MeV-He 2+ irradiation was also performed to implant about 85 appm He atoms at 120 °C by AVF cyclotron to 0.03 dpa, and small punch testing was performed to obtain ductile-to-brittle transition temperatures (DBTT). Radiation hardening of the neutron-irradiated specimens increased slightly with increasing helium production. The 100 MPa m 1/2 DBTT for the F82H + 11B, F82H + 10B + 11B, and F82H + 10B specimens were 40, 110, and 155 °C, respectively. The shifts of DBTT due to helium production were evaluated as about 70 °C by 190 appm He and 115 °C by 330 appm He. In cyclotron experiment using standard F82H, a similar DBTT shift due to He was measured. These results suggest that helium production can increase the DBTT.

  2. The effect of low dose irradiation on the impact fracture energy and tensile properties of pure iron and two ferritic martensitic steels

    NASA Astrophysics Data System (ADS)

    Belianov, I.; Marmy, P.

    1998-10-01

    Two batches of subsize V-notched impact bend specimens and subsize tensile specimens have been irradiated in the Saphir test reactor of the Paul Scherrer Institute (PSI). The first batch of specimen has been irradiated at 250°C to a dose of 2.65 × 10 19 n/cm 2 (0.042 dpa) and the second batch has been irradiated at 400°C to a dose of 8.12 × 10 19 n/cm 2 (0.13 dpa). Three different materials in three different microstructures were irradiated: pure iron and two ferritic steels, the alloy MANET 2 and a low activation composition CETA. The results of the impact tests and of the corresponding tensile tests are presented. Despite the very low neutron dose, a significant shift of the ductile to brittle transition temperature (DBTT) is observed. The influence of the test temperature on the impact energy is discussed for the irradiated and unirradiated conditions, with special emphasis on the microstructure.

  3. Effect of carbon on irradiation-induced grain-boundary phosphorus segregation in reactor pressure vessel steels using first-principles-based rate theory model

    NASA Astrophysics Data System (ADS)

    Ebihara, Ken-ichi; Yamaguchi, Masatake; Nishiyama, Yutaka; Onizawa, Kunio; Matsuzawa, Hiroshi

    2011-07-01

    In this paper, we incorporated the effect of carbon atoms on the irradiation-induced grain-boundary phosphorus segregation into the rate theory model by considering a carbon atom as a trap site of vacancies and self-interstitial atoms, and simulated the grain-boundary phosphorus coverage in the reactor pressure vessel steels, A533B steels which were neutron-irradiated using the Halden reactor. As a result, by selecting the sink strength of vacancies and self-interstitial atoms, the simulation reproduced the experimental grain-boundary phosphorus coverage that was measured using the scanning Auger electron microprobe analysis. It was observed that the grain-boundary phosphorus coverage does not depend on the dose rate regardless of the presence of carbon atoms. Furthermore, it was confirmed that vacancies scarcely transport phosphorus atoms to grain-boundaries as compared to the transport by self-interstitial atoms and it was found that carbon atoms influence the irradiation-induced phosphorus segregation by mainly suppressing the migration of vacancies.

  4. Observation and rate theory modeling of grain boundary segregation in Σ3 twin boundaries in ion-irradiated stainless steel 316

    NASA Astrophysics Data System (ADS)

    Lee, Gyeong-Geun; Jin, Hyung-Ha; Lee, Yong-Bok; Kwon, Junhyun

    2014-06-01

    Radiation-induced segregation (RIS) is the phenomenon of compositional change at point defect sinks in alloys irradiated at a moderate temperature. Owing to the potential relevance of RIS by way of the susceptibility of structural materials to irradiation-assisted stress corrosion cracking, basic research on austenitic stainless steels used in nuclear reactors has been carried out in recent years. In this work, commercial stainless steel 316 specimens were irradiated with Fe ions, and the resulting changes in Cr and Ni compositions were characterized using transmission electron microscopy and energy-dispersive X-ray spectroscopy. The samples with various grain boundary orientations, including the special Σ3 orientation, were analyzed. The ledges of a few special Σ3 twin boundaries showed significantly higher RIS compared to the coherent regions. The RIS behavior of a parallel twin pair was observed, and two profiles of RIS were found in them. The inner twins in multi-twins showed considerably lower RIS compared to the outer twins. For the calculation of RIS, time-dependent differential equations based on the rate theory were established and numerically integrated. An additional variable, representing the sink strength of the grain boundary, was introduced in the differential equations, and the concentration profiles of the Σ3 twins were calculated. The calculated results were in good agreement with the experimental results.

  5. Evaluation of irradiation effects of 16 MeV proton-irradiated 12Cr-1MoV steel by small punch (SP) tests

    SciTech Connect

    Chi, S.H.; Hong, J.H. ); Kim, I.S. . Dept. of Nuclear Engineering)

    1994-06-15

    Recently, interest in small-scale specimens for testing irradiated materials has arisen in conjunction with the need to develop materials for fusion reactor materials and to study irradiation effects using an ion irradiation facility. Several attempts have been made to evaluate material property changes due to irradiation using a small specimen technique. The SP (small punch) test is an example of small-scale specimen test techniques, originally developed by Baik et al. to estimate DBTT (ductile-to-brittle transition temperature) using broken standard CVN (Charpy 5-notch) specimens. The objective of the present study is to evaluate 16 MeV proton irradiation effects on a fusion reactor candidate material in terms of changes in energy up to failure and J[sub IC] fracture toughness (SP J[sub IC]) by using a SP test technique and a J[sub IC] - [bar [epsilon

  6. Effect of chloride ion on corrosion behavior of SUS316L-grade stainless steel in nitric acid solutions containing seawater components under γ-ray irradiation

    NASA Astrophysics Data System (ADS)

    Sano, Y.; Ambai, H.; Takeuchi, M.; Iijima, S.; Uchida, N.

    2017-09-01

    Concerning the Fukushima Daiichi nuclear power plant accident, we investigated the effect of chloride ion on the corrosion behavior of SUS316L stainless steel, which is a typical material for the equipment used in reprocessing, in HNO3 solution containing seawater components, including under γ-ray irradiation condition. Electrochemical and immersion tests were carried out using a mixture of HNO3 and artificial seawater (ASW). In the HNO3 solution containing high amounts of ASW, the cathodic current densities increased and uniform corrosion progressed. This might be caused by strong oxidants, such as Cl2 and NOCl, generated in the reaction between HNO3 and Cl- ions. The corrosion rate decreased with the immersion time at low concentrations of HNO3, while it increased at high concentrations. Under γ-ray irradiation condition, the corrosion rate decreased due to the suppression of the cathodic reactions by the reaction between the above oxidants and HNO2 generated by radiolysis.

  7. SANS and TEM of ferritic-martensitic steel T91 irradiated in FFTF up to 184 dpa at 413 °C

    NASA Astrophysics Data System (ADS)

    Van den Bosch, J.; Anderoglu, O.; Dickerson, R.; Hartl, M.; Dickerson, P.; Aguiar, J. A.; Hosemann, P.; Toloczko, M. B.; Maloy, S. A.

    2013-09-01

    Ferritic-martensitic steel T91 was previously irradiated in the Materials Open Test Assembly (MOTA) program of the Fast Flux Test Reactor Facility (FFTF) at 413 °C up to 184 dpa. The microstructure was analyzed by small angle neutron scattering (SANS) and transmission electron microscopy (TEM). Both SANS and TEM revealed a large fraction of voids with an average size of 29-32 nm leading to a calculated void swelling of 1.2-1.6% based on the volume fraction of the voids in the sample. SANS gave no indication of second phase particles having formed under irradiation in the material. Using TEM, one zone was found where a few G-phase particles were analyzed. Quantities were however too low to state reliable particle densities. No alpha prime (α') or Laves phase were observed in any of the investigated zones.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Void formation and helium effects in 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated in HFIR and FFTF at 400/degree/C

    SciTech Connect

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

    1988-01-01

    Martensitic/ferritic 9Cr-1MoVNb and 12Cr-1MoVW steels doped with up to 2 wt% Ni have up to 450 appm He after HFIR irradiation to /approximately/38 dpa, but only 5 appm He after 47 dpa in FFTF. No fine He bubbles and few or no larger voids were observable in any of these steels after FFTF irradiation at 407/degree/C. By contrast, many voids were found in the undoped steels (30-90 appm He) irradiated in HFIR at 400/degree/C, while voids plus many more fine He bubbles were found in the Ni-doped steels (400-450 appm He). Irradiation in both reactors at /approximately/400/degree/C produced significant changes in the as-tempered lath/subgrain boundary, dislocation, and precipitation structures that were sensitive to alloy composition, including doping with Ni. However, for each specific alloy the irradiation-produced changes were exactly the same comparing samples irradiated in FFTF and HFIR, particularly the Ni-doped steels. Therefore, the increased void formation appears solely due to the increased helium generation found in HFIR. While the levels of void swelling are relatively low after 37-39 dpa in HFIR (0.1-0.4%), details of the microstructural evolution suggest that void nucleation is still progressing, and swelling could increase with dose. The effect of helium on void swelling remains a valid concern for fusion application that requires higher dose experiments. 15 refs., 14 figs., 8 tabs.

  14. Atom probe tomographic analysis of high dose oxide-dispersion strengthened steel (alloy MA957) at selected irradiation conditions

    NASA Astrophysics Data System (ADS)

    Bailey, Nathan Alexander

    In an effort to understand the effect of high dose neutron irradiation on fast reactor cladding candidate materials, oxide-dispersion strengthened (ODS) alloy MA957 was irradiated to doses exceeding 100 displacements per atom (dpa) at various irradiation temperatures. The finely distributed Y-Ti-O particles, which provide MA957 its attractive properties, were examined by atom probe tomography (APT). Significant increases in oxide cluster number density and reductions in oxide cluster size were observed in specimens irradiated at 412 °C and below. A substantial hardness increase, measured by nanoindentation, was also observed at these low irradiation temperatures. It was found that the increase in oxide cluster number density, reduction in oxide cluster size, and associated increase in hardness is due to the inhibition of reformation processes of the Y-Ti-O particles following ballistic dissolution by incident radiation. Redistribution of oxide particle material along the grain boundaries is also observed at the low irradiation temperatures. The intermetallic phase alpha' was observed in the low temperature samples. This observation of this phase provides additional experimental evidence for the location of the phase boundary for this low temperature precipitate. The conclusion of this work is that MA957 is microstructurally stable under neutron irradiation at and above 495 °C.

  15. 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%.

  16. Stabilization of the spark-discharge point on a sample surface by laser irradiation for steel analysis.

    PubMed

    Matsuta, Hideyuki; Kitagawa, Kuniyuki; Wagatsuma, Kazuaki

    2006-10-01

    A combined technique with laser irradiation is suggested to control spark discharge for analytical use, having a unique feature that firing points of the spark discharge can be fixed by laser irradiation. Because the spark discharge easily initiates at particular surface sites, such as non-metallic inclusions, called selective discharge, the concentration of some elements sometimes deviates from their average one in spark discharge optical emission spectrometry. Therefore, stabilization of firing points on a sample surface could improve the analytical precision.

  17. 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.

  18. Experimental studies of the effect of irradiation on the anaerobic corrosion of carbon steel in relation to the Belgian supercontainer concept

    NASA Astrophysics Data System (ADS)

    Smart, N. R.; Fennell, P. A. H.; Rance, A. P.; Winsley, R. J.; Reddy, B.; Kursten, B.

    2011-04-01

    This paper describes recent results from an investigation of the effects of γ-radiation on the anaerobic corrosion of carbon steel in cement, in relation to the Belgian Supercontainer Concept for radioactive waste disposal. Anaerobic corrosion rates were measured by monitoring hydrogen evolution and the corresponding electrochemical behaviour was investigated by measuring open circuit potential and linear polarisation resistance. The test medium was alkaline simulated porewater, at γ-irradiation dose rates of 0 and 25 Gy hr-1, temperatures of 25 °C and 80 °C and chloride concentrations of 0 and 100 mg/l. The effects of radiation on the corrosion behaviour were found to be small.

  19. In-situ SCC observation of thermally-sensitized and cold-worked type 304 stainless steel irradiated to a neutron fluence of 1 × 10 25 n/m 2

    NASA Astrophysics Data System (ADS)

    Nakano, Junichi; Nemoto, Yoshiyuki; Miwa, Yukio; Usami, Koji; Tsukada, Takashi; Hide, Koichiro

    2009-04-01

    Crack initiation and crack growth processes of irradiation assisted stress corrosion cracking of stainless steels were studied by slow strain rate testing (SSRT) in oxygenated high temperature 561 K water. In-situ observation was carried out during SSRT for type 304 stainless steel irradiated to a neutron fluence of 1.0 × 10 25 n/m 2 ( E > 1 MeV) at 323 K in the Japan material testing reactor. The specimens were subjected to solution annealing, thermal sensitization, or cold working prior to neutron irradiation. The solution annealed material exhibited a combination of transgranular stress corrosion cracking (TGSCC) and ductile fracture, and almost all intergranular stress corrosion crackings were observed in the thermally-sensitized material. In the cold-worked material, cracking was introduced before the maximum stress was reached, and the fracture mode changed from TGSCC to ductile fracture to transgranular cracking together with the progress of crack growth in one direction.

  20. 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.

  1. 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.

  2. Modelling of the effect of dislocation channel on intergranular microcrack nucleation in pre-irradiated austenitic stainless steels during low strain rate tensile loading

    NASA Astrophysics Data System (ADS)

    Evrard, Pierre; Sauzay, Maxime

    2010-10-01

    In the present article, the effect of dislocation channel on intergranular microcrack nucleation during the tensile deformation of pre-irradiated austenitic stainless steels is studied. Because several slip planes are activated within the dislocation channel, the simple dislocation pile-up model seems not well suited to predict grain boundary stress field. Finite element computations, using crystal plasticity laws and meshes including a channel of finite thickness, are also performed in order to study the effect of some microstructural characteristics on grain boundary stress field. Numerical results show that: the thickness and the length of the dislocation channel influence strongly the grain boundary normal stress field. The grain boundary orientation with respect the stress axis does not affect so much the grain boundary normal stresses close to the dislocation channel. On the contrary far away the dislocation channel, the grain boundary stress field depends on the grain boundary orientation. Based on these numerical results, an analytical model is proposed to predict grain boundary stress fields. It is valuable for large ranges of dislocation channel thickness, length as well as applied stress. Then, a macroscopic microcrack nucleation criterion is deduced based on the elastic-brittle Griffith model. The proposed criterion predicts correctly the influence of grain boundary characteristics (low-angle boundaries (LABs), non-coincident site lattice (non-CSL) high-angle boundaries (HABs), special grain boundaries (GBs)) on intergranular microcrack nucleation and the macroscopic tensile stress required for grain boundary microcrack nucleation for pre-irradiated austenitic stainless steels deformed in argon environment. The criterion based on a dislocation pile-up model (Smith and Barnby) underestimates strongly the nucleation stress. These results confirm that pile-up models are not well suited to predict microcrack nucleation stress in the case of dislocation

  3. Microstructural behavior of VVER-440 reactor pressure vessel steels under irradiation to neutron fluences beyond the design operation period

    NASA Astrophysics Data System (ADS)

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Nikolaev, Yu. A.; Pechenkin, V. A.

    2005-06-01

    Electron-microscopy and fractographic studies of the surveillance specimens from base and weld metals of VVER-440/213 reactor pressure vessel (RPV) in the original state and after irradiations to different fast neutron fluences from ˜5 × 10 23 n m -2 ( E > 0.5 MeV) up to over design values have been carried out. The maximum specimens irradiation time was 84 480 h. It is shown that there is an evolution in radiation-induced structural behavior with radiation dose increase, which causes a change in relative contribution of the mechanisms responsible for radiation embrittlement of RPV materials. Particularly, radiation coalescence of copper-enriched precipitates and extensive density increase of dislocation loops was observed. Increase in dislocation loop density was shown to provide the dominant contribution to radiation hardening at the late irradiation stages (after reaching double the design end-of-life neutron fluence of ˜4 × 10 24 n m -2). The fracture mechanism of the base metal at those stages was observed to change from transcrystalline to intercrystalline.

  4. BWRVIP-140NP: BWR Vessel and Internals Project Fracture Toughness and Crack Growth Program on Irradiated Austenitic Stainless Steel

    SciTech Connect

    Gilman, J

    2005-03-15

    To prepare for this project, EPRI and BWRVIP conducted a workshop at Ponte Vedra Beach, Florida during February 19-21, 2003 (EPRI report 1007822). Attendees were invited to exchange relevant information on the effects of irradiation on austenitic materials in light water reactors and to produce recommendations for further work. EPRI reviewed the data, recommendations, and conclusions derived from the workshop and developed prioritized test matrices defining new data needs. Proposals were solicited, and selected proposals are the basis for the program described in this report. Results The planned test matrix for fracture toughness testing includes 21 tests on 5 materials.

  5. Detection of irradiation embrittlement of low-alloy steel for nuclear reactor pressure vessels using a probe type eddy current sensor

    SciTech Connect

    Maeda, Noriyoshi; Yamaguchi, Atsunori; Sugibayashi, Takuya; Kohno, Katsumi

    1999-10-01

    This report describes the results of studies made for the purpose of detecting the irradiation embrittlement of low-alloy steel used for nuclear reactor pressure vessels. For the method of using eddy current to detect material degradation, the device and the sensor employed are light in weight and compact in size, allowing testing without contact. In this study the frequency of input current to the excitation coil is changed in steps of 1 kHz. The output signal is processed by phase detection method, and displayed on a complex plane. It depicts a trajectory as the frequency is changed. To extract features of the trajectories, averaged radius and averaged phase angle are defined and plotted as function of neutron fluence or ductile-brittle transition temperature. Experiment shows that the averaged phase angle and transition temperature decrease as the neutron fluence is increased. Behavior of the averaged phase angle is interpreted employing magnetic permeability and electric conductivity of the test specimens. It becomes clear that electric conductivity decreases as the neutron fluence is increased.

  6. 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.

  7. TEM observations and finite element modelling of channel deformation in pre-irradiated austenitic stainless steels - Interactions with free surfaces and grain boundaries

    NASA Astrophysics Data System (ADS)

    Sauzay, Maxime; Bavard, Karine; Karlsen, Wade

    2010-11-01

    Transmission electron microscopy (TEM) observations show that dislocation channel deformation occurs in pre-irradiated austenitic stainless steels, even at low stress levels (˜175 MPa, 290 °C) in low neutron dose (˜0.16 dpa, 185 °C) material. The TEM observations are utilized to design finite element (FE) meshes that include one or two "soft" channels (i.e. low critical resolved shear stress (CRSS)) of particular aspect ratio (length divided by thickness) embedded at the free surface of a "hard" matrix (i.e. high CRSS). The CRSS are adjusted using experimental data and physically based models from the literature. For doses leading to hardening saturation, the computed surface slips are as high as 100% for an applied stress close to the yield stress, when the observed channel aspect ratio is used. Surface slips are much higher than the grain boundary slips because of matrix constraint effect. The matrix CRSS and the channel aspect ratio are the most influential model parameters. Predictions based on an analytical formula are compared with surface slips computed by the FE method. Predicted slips, either in surface or bulk channels, agree reasonably well with either atomic force microscopy measures reported in the literature or measures based on our TEM observations. Finally, it is shown that the induced surface slip and grain boundary stress concentrations strongly enhance the kinetics of the damage mechanisms possibly involved in IASCC.

  8. Using complimentary microscopy methods to examine Ni-Mn-Si-precipitates in highly-irradiated reactor pressure vessel steels

    DOE PAGES

    Edmondson, P. D.; Parish, C. M.; Nanstad, R. K.

    2017-05-29

    Nano-scale Ni-Mn-Si-rich precipitates formed in a reactor pressure vessel steel under high neutron fluence have been characterized using highly complimentary atom probe tomography (APT) and scanning transmission electron microscopy with energy dispersive spectroscopy (STEM-EDS) combined with STEM-EDS modeling. Using these techniques in a synergistic manner to overcome the well-known trajectory aberrations in APT data, the average upper limit Fe concentration within the precipitates was found to be ~6 at.%. Using this knowledge, accurate compositions of the precipitates was determined and it was found that the spread of precipitate compositions was large, but mostly centered around the Γ2-and G-phases. The usemore » of STEM-EDS also allowed for larger areas to be examined, and segregation of minor solutes was observed to occur on grain boundaries, along with Ni-Mn-Si-rich precipitates that were smaller in size than those in the matrix. Solute segregation at the grain boundaries is proposed to occur through a radiation induced segregation or radiation enhanced diffusion mechanism due to the presence of a denuded zone about the grain boundary. It is also proposed that the reduced precipitate size at the grain boundaries is due to the structure of the grain boundary. The lack of Ni-Mn-Si precipitates observed in larger Mo-rich precipitates is also discussed, and the absence of the minor solutes required to form the Ni-Mn-Si precipitates results in the lack of nucleation. This is in contrast to cementite phases in which Ni-Mn-Si precipitates have been seen to be formed. It was also determined through this work that the exclusion of all the Fe ions during atom probe analysis is a reasonable approximation.« less

  9. 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.

  10. The Effect of Oversize Solute Additions on the Irradiation-Assisted Stress Corrosion Cracking Resistance of Austenitic Stainless Steels

    SciTech Connect

    M Hackett; G Was

    2005-08-12

    Solute additions of zirconium are believed to decrease RIS and dislocation density through point defect trapping and recombination, which in turn reduces grain boundary sensitization and IGSCC. In this work, the effect of zirconium on the microstructure, microchemistry, hardening and IGSCC behavior of 316SS doped with zirconium to levels of 0.31 and 0.45 wt% was studied. These alloys were then irradiated with 3.2 MeV protons to doses up to 7 dpa at a temperature of 400 C. Zr additions had relatively little effect on radiation hardening. Dislocation densities were reduced and average sizes slightly increased for the +Zr alloys relative to the 316SS. Although a low amount of swelling was seen in 316SS at 3 dpa, no voids were observed in either of the +Zr alloys at 3 or 7 dpa. The difference in RIS of Cr and Ni between 316SS and 316+LoZr at 3 dpa was negligible, though RIS for 316+HiZr was considerably less than 316+LoZr at 7 dpa. The link between the oversize solute addition of Zr and its effect on IASCC shows that although the percent strain to failure increased substantially for 316+LoZr compared to the 316SS, cracking behavior was substantially worse as the number of cracks and total crack length was increased by more than an order of magnitude.

  11. Results of Charpy V-Notch Impact Testing of Structural Steel Specimens Irradiated at ~30°C to 1 x 1016 neutrons/cm2 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 ~30°C (~ 85°F) in the cavity of a commercial nuclear power plant to a fluence of 1 x 1016 neutrons/cm2 (>1 MeV). 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 ~ 99 J, while the energy of 11 unirradiated specimens from other locations of the same plate was 45 J, a difference of ~ 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.

  12. 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.

  13. Spectral and raw quasi in-situ energy dispersive X-ray data captured via a TEM analysis of an ODS austenitic stainless steel sample under 1 MeV Kr(2+) high temperature irradiation.

    PubMed

    Brooks, Adam J; Yao, Zhongwen

    2017-10-01

    The data presented in this article is related to the research experiment, titled: 'Quasi in-situ energy dispersive X-ray spectroscopy observation of matrix and solute interactions on Y-Ti-O oxide particles in an austenitic stainless steel under 1 MeV Kr(2+) high temperature irradiation' (Brooks et al., 2017) [1]. Quasi in-situ analysis during 1 MeV Kr(2+) 520 °C irradiation allowed the same microstructural area to be observed using a transmission electron microscope (TEM), on an oxide dispersion strengthened (ODS) austenitic stainless steel sample. The data presented contains two sets of energy dispersive X-ray spectroscopy (EDX) data collected before and after irradiation to 1.5 displacements-per-atom (~1.25×10(-3) dpa/s with 7.5×10(14) ions cm(-2)). The vendor software used to process and output the data is the Bruker Esprit v1.9 suite. The data includes the spectral (counts vs. keV energy) of the quasi in-situ scanned region (512×512 pixels at 56k magnification), along with the EDX scanning parameters. The.raw files from the Bruker Esprit v1.9 output are additionally included along with the.rpl data information files. Furthermore included are the two quasi in-situ HAADF images for visual comparison of the regions before and after irradiation. This in-situ experiment is deemed 'quasi' due to the thin foil irradiation taking place at an external TEM facility. We present this data for critical and/or extended analysis from the scientific community, with applications applying to: experimental data correlation, confirmation of results, and as computer based modeling inputs.

  14. 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.

  15. Results from the irradiation of stainless steel and copper by 23 MeV γ-quanta in the atmosphere of molecular deuterium at a pressure of 2 kbar

    NASA Astrophysics Data System (ADS)

    Didyk, A. Yu.; Wisniewski, R.

    2014-05-01

    Metal samples were arranged inside a "finger-type" high-pressure chamber (DHPC-FT) filled by deuterium. They were two aluminum rods, a copper rod, two specimens of homogeneous YMn2 alloy, and a stainless steel wire. The pressure of molecular deuterium in DHPC-FT was about 2 kbar. The samples were irradiated by braking γ-quanta at a threshold energy of 23 MeV. All the samples were studied using scanning electron microscopy (SEM) and X-ray (roentgen) microelement probe analysis (RMPA). Considerable changes in the surface structure and elemental composition were found for the samples of copper, aluminum, YMn2 alloy, and stainless steel. Unusual results were analyzed in detail and compared with the earlier data.

  16. 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

    NASA Astrophysics Data System (ADS)

    Maloy, S. A.; Saleh, T. A.; Anderoglu, O.; Romero, T. J.; Odette, G. R.; Yamamoto, T.; Li, S.; Cole, J. I.; Fielding, R.

    2016-01-01

    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(ɛ) laws for all of these alloys. In the irradiated condition, the σ(ɛ) 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 the latter can be understood in terms of the alloy's σ(ɛ) behavior. Increases in the average σ(ɛ) in the range of 0-10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are also 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. Notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.

  17. 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; ...

    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

  18. Microstructure and mechanical properties of austenitic stainless steel 12X18H9T after neutron irradiation in the pressure vessel of BR-10 fast reactor at very low dose rates

    SciTech Connect

    Porollo, S. I.; Dvoriashin, Alexander M.; Konobeev, Yury V.; Ivanov, A. A.; Shulepin, S. V.; Garner, Francis A.

    2006-12-01

    Results are presented for void swelling, microstructure andmechanical properties of Russian 12X18H9T (0.12C-18Cr-9Ni-Ti) austenitic stainless steel irradiated as a pressure vessel structure material of the BR-10 fast reactor at ~350C to only 0.64 dpa, produced by many years of exposure at the very low displacement rate of only 1.9x10-9 dpa/s. In agreement with a number of other recent studies it appears that lower dpa rates have a pronounced effect on the microstructure and resultant mechanical properties. In general, loweer dpa rates lead to the onset of swelling at much lower doses compared to comparable irradiations conducted at higher dpa rates.

  19. 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.)

  20. 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.

  1. Generation and Retention of Helium and Hydrogen in Austenitic Steels Irradiated in a Variety of LWR and Test Reactor Spectral Environments

    SciTech Connect

    Garner, Francis A.; Oliver, Brian M.; Greenwood, Lawrence R.; Edwards, Danny J.; Bruemmer, Stephen M.; Grossbeck, Martin L.

    2002-03-31

    In fission and fusion reactor environments stainless steels generate significant amounts of helium and hydrogen by transmutation. The primary sources of helium are boron and nickel, interacting with both fast and especially thermal neutrons. Hydrogen arises primarily from fast neutron reactions, but is also introduced into steels at often much higher levels by other environmental processes. Although essentially all of the helium is retained in the steel, it is commonly assumed that most of the hydrogen is not retained. It now appears that under some circumstances, significant levels of hydrogen can be retained, especially when helium-nucleated cavities become a significant part of the microstructure. A variety of stainless steel specimens have been examined from various test reactors, PWRs and BWRs. These specimens were exposed to a wide range of neutron spectra with different thermal/fast neutron ratios. Pure nickel and pure iron have also been examined. It is shown that all major features of the retention of helium and hydrogen can be explained in terms of the composition, thermal/fast neutron ratio and the presence or absence of helium-nucleated cavities. In some cases, the hydrogen retention is very large and can exceed that generated by transmutation, with the additional hydrogen arising from either environmental sources and/or previously unidentified radioisotope sources that may come into operation at high neutron exposures.

  2. Effects of alloying elements on radiation hardening based on loop formation of electron-irradiated light water reactor pressure vessel model steels

    NASA Astrophysics Data System (ADS)

    Nishi, Takakuni; Hashimoto, N.; Ohnuki, S.; Yamamoto, T.; Odette, G. R.

    2011-10-01

    Electron irradiations using a high voltage electron microscope were conducted on several reactor pressure vessel model alloys in order to investigate the effects of alloying elements on the formation and development of defect clusters. In addition, the effects of alloying elements on yield stress change after irradiation were considered, comparing the mean size and number density of dislocation loops with the irradiation-induced hardening. High Cu alloys formed Cu and Mn-Ni-Si rich clusters, and these are important in determining the yield stress increase. High Ni alloys formed a high density of small dislocation loops and probably Mn-Ni-Si rich cluster, which have the effect of increasing the yield stress. High P enhanced radiation-induced segregation on grain boundary, helping prevent dislocation movement.

  3. 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.

  4. On the effects of irradiation and helium on the yield stress changes and hardening and non-hardening embrittlement of ˜8Cr tempered martensitic steels: Compilation and analysis of existing data

    NASA Astrophysics Data System (ADS)

    Yamamoto, Takuya; Odette, G. Robert; Kishimoto, Hirotatsu; Rensman, Jan-Willem; Miao, Pifeng

    2006-09-01

    Data on irradiation hardening and embrittlement of 8-10Cr normalized and tempered martensitic steel (TMS) alloys has been compiled from the literature, including results from neutron, spallation proton (SP) and He-ion (HI) irradiations. Limitations of this database are briefly described. Simple, phenomenological-empirical fitting models were used to assess the dose (displacement-per-atom, dpa), irradiation temperature ( Ti) and test temperature ( Tt) dependence of yield stress changes (Δ σy), as well as the corresponding dependence of sub-sized Charpy V-notch impact test transition temperature shifts (Δ Tc). The Δ σy are generally similar for SP and neutron irradiations, with very high and low helium to dpa ratios, respectively. Further, the Δ σy trends were found to be remarkably consistent with the Ti and dpa hardening-dependence of low alloy steels irradiated at much lower doses. The similar Ti and (low) dose dependence of Δ σy and Δ Tc, as well as an analysis of paired Δ Tc-Δ σy datasets, show that embrittlement is typically dominated by a hardening mechanism below about 400 °C. However, the corresponding hardening-Charpy shift coefficient, Cc = Δ Tc/Δ σy ≈ 0.38 ± 0.18 °C/MPa is lower than that for the fracture toughness reference temperature, T0, with Δ T0/Δ σy ≈ 0.58 ± 0.1 °C/MPa, indicating that sub-sized Charpy tests provide non-conservative estimates of embrittlement. The Cc increases at Ti > 400 °C, and Δ Tc > 0 are sometimes observed in association with Δ σy ⩽ 0, indicative of a non-hardening embrittlement (NHE) contribution. Analysis of limited data on embrittlement due to thermal aging supports this conclusion, and we hypothesize that the NHE regime may be shifted to lower temperatures by radiation enhanced diffusion. Possible effects of helium on embrittlement for Ti between 300 and 400 °C are also assessed based on observed trends in Cc. The available data is limited, scattered, and potentially confounded. However

  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. 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.

  8. 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.

  9. Tensile properties of irradiated surveillance coupons

    SciTech Connect

    Huang, F.H.; Blackburn, L.D.

    1994-06-01

    Tensile testing of austenitic steel and superalloy samples irradiated in the HMO 13 assembly was performed in support of the Fast Flux Test Facility (FFTF) Surveillance Program. Postirradiation yield stress, ultimate tensile stress, uniform elongation, total elongation, and reduction in area of 304 stainless steel (SS), 308 SS weld, 316 SS, A286, In718, and In718 weld were determined. Results showed the strength of austenitic steels increased while the ductility decreased as a result of irradiation. Low irradiation exposure produced little property change in In718. Overall, the tensile properties of HMO 13 surveillance coupons showed a lower magnitude of irradiation-induced property change than was expected based on earlier studies. Results from these tests gave no indications of unexpectedly severe irradiation damage to FFTF components.

  10. Heavy-section steel technology program: An overview

    SciTech Connect

    Pugh, C.E.

    1986-01-01

    The major focus of the current irradiation work is on understanding the fracture characteristics of irradiated high-copper weld materials and stainless steel cladding. Testing in the Fifth and Sixth HSST Irradiation Series are underway and include two high-copper weld materials (CU = 0.25 and 0.30%). Both crack initiation and arrest-toughness properties are being investigated. The tests also include Charpy V-notch, tensile, and compact specimens, with the latter ranging in size up to 4TCS for the irradiated high-copper weld materials. The Seventh Irradiation Series is examining the effects of neutron exposure on the fracture properties of stainless steel cladding.

  11. (Irradiation embrittlement of reactor pressure vessels)

    SciTech Connect

    Corwin, W.R.

    1990-09-24

    The traveler served as a member of the two-man US Nuclear Regulatory Commission sponsored team who visited the Prometey Complex in Leningrad to assess the potential for expanded cooperative research concerning integrity of the primary pressure boundary in commercial light-water reactors. The emphasis was on irradiation embrittlement, structural analysis, and fracture mechanics research for reactor pressure vessels. At the irradiation seminar in Cologne, presentations were made by German, French, Finnish, Russian, and US delegations concerning many aspects of irradiation of pressure vessel steels. The traveler made presentations on mechanisms of irradiation embrittlement and on important aspects of the Heavy-Section Steel Irradiation Program results of irradiated fracture mechanics tests.

  12. Aircraft Steels

    DTIC Science & Technology

    2009-02-19

    NAWCADPAX/TR-2009/ 12 AIRCRAFT STEELS by E. U. Lee R. Taylor C. Lei H. C. Sanders 19 February 2009...MARYLAND NAWCADPAX/TR-2009/ 12 19 February 2009 AIRCRAFT STEELS by E. U. Lee R. Taylor C. Lei H. C. Sanders...Prescribed by ANSI Std. Z39-18 NAWCADPAX/TR-2009/ 12 ii SUMMARY Five high strength and four stainless steels have been studied, identifying their

  13. 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.

  14. Mechanism of the influence of transmutation helium produced in the claddings of fuel elements made of an austenitic steel ChS-68 during neutron irradiation on the formation of pores

    NASA Astrophysics Data System (ADS)

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

    2009-09-01

    Results of the determination of the porosity characteristics of steel ChS-68 are presented. To describe the processes of pore formation in this steel, the model of point-defect migration was used. The participation of helium both in the formation of the pore nuclei and in the processes of their further growth is considered.

  15. Fracture mechanism of borated stainless steel

    SciTech Connect

    He, J.Y.; Soliman, S.E.; Baratta, A.J.; Balliett, T.A.

    2000-05-01

    The mechanical properties and fracture mechanism of irradiated and unirradiated boron containing Type 304 stainless steel are studied. Four different batches with different boron weight percentages are used. One of these batches was manufactured by a conventional wrought technique, while the others were manufactured by a powder metallurgy technique. The irradiated specimens were subjected to a fluence level of 5 x 10{sup 19} or 1 {times} 10{sup 21} n/m{sup 2}. The mechanical and fracture tests were performed at temperatures of 233, 298, and 533 K. No significant effects on the mechanical properties or fracture behavior were observed as a result of neutron irradiation and/or temperature. The ductility and toughness of the borated steel were found to decrease with increasing boron content. The effect of boride on void nucleation and linkage was found to play an important role in the fracture behavior of borated steel.

  16. Electron Microscopy Study of Stainless Steel Radiation Damage Due to Long-Term Irradation by Alpha Particles Emitted From Plutonium

    SciTech Connect

    Unlu, Kenan; Rios-Martinez, Carlos; Saglam, Mehmet; Hart, Ron R.; Shipp, John D.; Rennie, John

    1998-04-16

    Radiation damage and associated surface and microstructural changes produced in stainless steel encapsulation by high-fluence alpha particle irradiations from weapons-grade plutonium of 316-stainless steel are being investigated.

  17. Stainless steel

    SciTech Connect

    Lula, R.A.

    1985-01-01

    This book discusses the stainless steels for high-strength, heat-resistant or corrosion-resistant applications. It is a treatment of the properties and selection of stainless steels. Up-to-date information covers physical, mechanical and chemical properties of all stainless grades, including the new ferritic and duplex grades. The book covers physical metallurgy as well as processing and service characteristics, including service in corrosive environments. It deals with wrought and cast stainless steels and reviews fabrication from cold-forming to powder metallurgy.

  18. APT characterization of high nickel RPV steels

    NASA Astrophysics Data System (ADS)

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

    2006-06-01

    The microstructures of three high nickel content pressure vessel steels have been characterized by atom probe tomography to investigate the influence of high nickel levels on the response to neutron irradiation of high and low copper pressure vessel steels. The high-nickel, low-manganese, low-copper VVER-1000 weld and forging exhibited lower than predicted levels of embrittlement during neutron irradiation. The Palisades weld exhibits a Δ T41 J of 102 °C which was significantly lower than the value of 154 °C predicted by Reg. Guide 1.99 Rev. 2. Atom probe tomography revealed nickel-, manganese-, and silicon-enriched precipitates in both the VVER-1000 base and weld materials after neutron irradiation. A high number density of copper-, nickel-, manganese-, silicon- and phosphorus-enriched precipitates were observed in the Palisades weld after neutron irradiation. Atom probe tomography also revealed high levels of phosphorus segregation to the dislocations in all three materials.

  19. 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.

  20. 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.

  1. Irradiation performance of FFTF drivers using the D9 alloy

    SciTech Connect

    Pitner, A.L.; Gneiting, B.C.; Bard, F.E.

    1994-12-31

    Five test assemblies similar in design to the Fast Flux Test Facility driver fuel assembly , but employing the alloy D9 in place of stainless steel 316 for duct, cladding, and wire wrap compnents were irradiated to demonstrate the improved performance of the new design. Results of post-irradiation examinations are discussed.

  2. Comparison of swelling for structural materials on neutron and ion irradiation

    SciTech Connect

    Loomis, B.A.

    1986-03-01

    The swelling of V-base alloys, Type 316 stainless steel, Fe-25Ni-15Cr alloys, ferritic steels, Cu, Ni, Nb-1% Zr, and Mo on neutron irradiation is compared with the swelling for these materials on ion irradiation. The results of this comparison show that utilization of the ion-irradiation technique provides for a discriminative assessment of the potential for swelling of candidate materials for fusion reactors.

  3. JPDR vessel steel examination

    SciTech Connect

    Corwin, W.R.; Broadhead, B.L.; Sokolov, M.A.

    1995-10-01

    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 which has been irradiated during normal service. This task has been included with the HSSI Program to provide just such an evaluation of material from the wall of the pressure vessel from the JPDR. The JPDR was a small BWR that began operation in 1963. It operated until 1976, accumulating {approximately}17,000 h of operation, of which a little over 14,000 h were with the original 45-MWTh core, and the remaining fraction, late in life, with an upgraded 90-MWTh core. The pressure vessel of the JPDR, fabricated from A 302, grade B, modified steel with an internal weld overlay cladding of 304 stainless steel, is approximately 2 m ID and 73 mm thick. It was fabricated from two shell halves joined by longitudinal seam welds located 180{degrees} from each other. The rolling direction of the shell plates is parallel to the axis of the vessel. It operated at 273{degrees}C and reached a maximum fluence of about 2.3 x 10{sup 18} n/cm{sup 2} (> 1 MeV). The impurity contents in the base metal are 0.10 to 0.11% Cu and 0.010 to 0.017% P with a nickel content of 0.63 to 0.65%. Impurity contents of the weld metal are 0.11 to 0.14% Cu and 0.025 to 0.039% P with a nickel content of 0.59%.

  4. 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...

  5. 10 CFR 36.33 - Irradiator pools.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 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 meter...

  6. 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 meter...

  7. 10 CFR 36.33 - Irradiator pools.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 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 meter...

  8. 10 CFR 36.33 - Irradiator pools.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 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 meter...

  9. Evaluation of the reactor pressure vessel steels by positron annihilation

    NASA Astrophysics Data System (ADS)

    Slugeň, V.; Hein, H.; Sojak, S.; Simeg Veterníková, J.; Petriska, M.; Sabelová, V.; Pavúk, M.; Hinca, R.; Stacho, M.

    2013-11-01

    This paper presents a comparison of commercially used German and Russian reactor pressure vessel steels from the positron annihilation spectroscopy (PAS) point of view, having in mind knowledge obtained also from other techniques from the last decades. The second generation of Russian RPV steels seems to be fully comparable with German steels and their quality allows prolongation of NPP operating lifetime over projected 40 years. The embrittlement of CrMoV steels is relatively low due to effect of higher temperature which implies partial in situ annealing of primary microstructural point defects and therefore delays the degradation processes caused by neutron irradiation.

  10. Crack arrest in thick section steel plate

    NASA Astrophysics Data System (ADS)

    Smith, E.

    1983-03-01

    Crack arrest in thick section steel plate is considered in relation to the conditions for crack arrest in a nuclear reactor pressure vessel, when this is subjected to thermal stresses resulting from a hypothetical loss of coolant accident. The results of a theoretical analysis, based primarily on recent developments in quasi-static crack propagation theory, provide further support for the view that the arrest toughness KIa is essentially a material property. However, since the theoretical results also suggest that KIa is reduced by neutron irradiation, and because there is, as yet, no conclusive experimental data on the effect of neutron irradiation on KIa, it is proposed that with highly irradiated steel, instead of using a KIa crack arrest criterion, it is better to use a more conservative criterion, based on the concept that arrest occurs within the vessel at a position where the temperature exceeds that temperature above which the cleavage fracture mode is unable to operate.

  11. Diffusion Couple Alloying of Refractory Metals in Austenitic and Ferritic/Martensitic Steels

    DTIC Science & Technology

    2012-03-01

    raised in the samples, the main mode was dislocation channeling [8]. b. Irradiation -Induced Precipitation and Grain Boundary Segregation In materials...centered Cubic FIB Focused Ion Beam HTGR High Temperature Gas Reactor IASCC Irradiation -assisted Stress Corrosion Cracking LMFBR Liquid Metal Fast...displacements per atom (dpa) (From [8]). Observations on irradiated austenitic steel determined that two deformation modes were present: dislocation

  12. The effect of tantalum on the mechanical properties of a 9Cr 2W 0.25V 0.07Ta 0.1C steel

    NASA Astrophysics Data System (ADS)

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

    1999-07-01

    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 (FFTF) 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 ≈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 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 ≈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.

  13. 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.

  14. [Food irradiation].

    PubMed

    Migdał, W

    1995-01-01

    A worldwide standard on food irradiation was adopted in 1983 by Codex Alimentarius Commission of the Joint Food Standard Programme of the Food and Agriculture Organization (FAO) of the United Nations and the World Health Organization (WHO). As a result, 41 countries have approved the use of irradiation for treating one or more food items and the number is increasing. Generally, irradiation is used to: food loses, food spoilage, disinfestation, safety and hygiene. The number of countries which use irradiation for processing food for commercial purposes has been increasing steadily from 19 in 1987 to 33 today. In the frames of the national programme on the application of irradiation for food preservation and hygienization an experimental plant for electron beam processing has been established in Institute of Nuclear Chemistry and Technology. The plant is equipped with a small research accelerator Pilot (19MeV, 1 kW) and an industrial unit Elektronika (10MeV, 10 kW). On the basis of the research there were performed at different scientific institutions in Poland, health authorities have issued permission for irradiation for: spices, garlic, onions, mushrooms, potatoes, dry mushrooms and vegetables.

  15. Steel Rattler

    NASA Astrophysics Data System (ADS)

    Trudo, Robert A.; Stotts, Larry G.

    1997-07-01

    Steel Rattler is a multi-phased project to determine the feasibility of using commercial off-the-shelf components in an advanced acoustic/seismic unattended ground sensor. This project is supported by the Defense Intelligence Agency through Sandia National Laboratories as the lead development agency. Steel Rattler uses advanced acoustic and seismic detection algorithms to categorize and identify various heavy vehicles down to the number of cylinders in the engine. This detection is accomplished with the capabilities of new, high-speed digital signal processors which analyze both acoustic and seismic data. The resulting analysis is compared against an onboard library of known vehicles and a statistical match is determined. An integrated thermal imager is also employed to capture digital thermal images for subsequent compression and transmission. Information acquired by Steel Rattler in the field is transmitted in small packets by a built-in low-power satellite communication system. The ground station receivers distribute the coded information to multiple analysis sites where the information is reassembled into coherent messages and images.

  16. 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)

  17. 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.

  18. 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.

  19. Irradiated foods

    MedlinePlus

    ... it reduces the risk for 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.

  20. Supertough Stainless Bearing Steel

    NASA Technical Reports Server (NTRS)

    Olson, Gregory B.

    1995-01-01

    Composition and processing of supertough stainless bearing steel designed with help of computer-aided thermodynamic modeling. Fracture toughness and hardness of steel exceeds those of other bearing steels like 440C stainless bearing steel. Developed for service in fuel and oxidizer turbopumps on Space Shuttle main engine. Because of strength and toughness, also proves useful in other applications like gears and surgical knives.

  1. Irradiation response on mechanical properties of neutron irradiated F82H

    NASA Astrophysics Data System (ADS)

    Shiba, K.; Suzuki, M.; Hishinuma, A.

    1996-10-01

    Tensile and Charpy impact properties of neutron irradiated F82H (Fe8Cr2WVTa) with and without boron have been investigated to obtain the basic irradiation response on mechanical properties in low damage regime less than 1 dpa at the temperature ranging from 300° to 590°C. Boron-doped steel was used for the helium effect due to (n, α) reaction. Typical irradiation hardening was observed at 300°C. The irradiation above 520°C did not reveal increase in yield stress, but the specimen irradiated at 590°C showed some reduction in elongation in room temperature tensile testing. Slight difference in the tensile properties between boron-doped and boron-free were observed at 590°C. No changes in ductile brittle transition temperature (DBTT) occurred at a temperature between 335° and 460°C by Charpy impact testing.

  2. Preliminary analysis of irradiation effects on CLAM after low dose neutron irradiation

    NASA Astrophysics Data System (ADS)

    Peng, Lei; Huang, Qunying; Li, Chunjing; Liu, Shaojun

    2009-04-01

    To investigate the irradiation effects on a new version of reduced activation ferritic/martensitic steels (RAFMs) i.e. China Low Activation Martensitic steel (CLAM), neutron irradiation experiments has been being carried out under wide collaboration in China and overseas. In this paper, the mechanical properties of CLAM heats 0603A, 0408B, and 0408D were investigated before and after neutron irradiation to ˜0.02 dpa at 250 °C. The test results showed that ultimate strength and yield stress of CLAM HEAT 0603A increased about 10-30 MPa and ductile to brittle transition temperature (DBTT) shift was about 5 °C. For CLAM heats 0408B and 0408D, ultimate strength and yield stress increased about 80-150 MPa.

  3. R&D of low activation ferritic steels for fusion in japanese universities*1

    NASA Astrophysics Data System (ADS)

    Kohyama, Akira; Kohno, Yutaka; Asakura, Kentaro; Kayano, Hideo

    1994-09-01

    Following the brief review of the R&D of low activation ferritic steels in Japanese universities, the status of 9Cr-2W type ferritic steels development is presented. The main emphasis is on mechanical property changes by fast neutron irradiation in FFTF. Bend test, tensile test, CVN test and in-reactor creep results are provided including some data about low activation ferritic steels with Cr variation from 2.25 to 12%. The 9Cr-2W ferritic steel, denoted as JLF-1, showed excellent mechanical properties under fast neutron irradiation as high as 60 dpa. As potential materials for DEMO and beyond, innovative oxide dispersion strengthened (ODS) quasi-amorphous low activation ferritic steels are introduced. The baseline properties, microstructural evolution under ion irradiation and the recent progress of new processes are provided.

  4. Radiation embrittlement of manganese-stabilized martensitic stainless steel

    SciTech Connect

    Gelles, D.S.; Hu, W.L.

    1986-12-01

    Fractographic examination has been performed on selected Charpy specimens of manganese stabilized martensitic stainless steels in order to identify the cause of irradiation embrittlement. Embrittlement was found to be partly due to enhanced failure at grain boundaries arising from precipitation. Microstructural examination of a specimen irradiated at higher temperature has demonstrated the presence of Fe-Cr-Mn chi phase, a body centered cubic intermetallic phase known to cause embrittlement. This work indicated that manganese stabilized martensitic stainless steels are prone to intermetallic phase formation which is detrimental to mechanical properties.

  5. Saturation of the DBTT shift of irradiated 12Cr-1MoVW with increasing fluence

    SciTech Connect

    Vitek, J.M.; Corwin, W.R.; Klueh, R.L.; Hawthorne, J.R.

    1986-01-01

    Ferritic/martensitic steels are considered one of the prime candidate materials for structural applications in fusion reactors. This class of steel provides many advantages including excellent swelling resistance and high thermal conductivity. A reiew of current results on the suitability of ferritic/martensitic steels reveals their corrosion behavior and postirradiation properties are, in general, quite promising. However, a major concern is the existence in these steels of a ductile-to-brittle transition temperature (DBTT), below which the steel has limited ductility and fails in a brittle mode. Although the DBTT, as measured in a Charpy impact test, may be below room temperature in the unirradiated condition, irradiation can increase the DBTT significantly, and the increase depends on the irradiation temperature. Increases in the DBTT of greater than 100/sup 0/C have been measured following irradiation at 400/sup 0/C and below.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Development of Outer Surface Irradiated Laser Stress Improvement Process (L-SIP)

    SciTech Connect

    Noriaki Sugimoto; Hironori Onitsuka; Koji Okimura; Takahiro Ohta; Kazuhiko Kamo

    2006-07-01

    Improvement of residual stress is effective in a countermeasure to deal with the stress corrosion cracks in pipe welds. A irradiated laser stress improvement process (L-SIP) will be introduced as a method to improve residual stress inside steel pipes. This work method is to improve inner surface residual stress from tensile stress to compressive stress by irradiating laser beam around the welds of steel pipe and utilizing the temperature differences between inner and outer surface. (authors)

  11. 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.

  12. Evaluation of stainless steels for their resistance to intergranular corrosion

    NASA Astrophysics Data System (ADS)

    Korostelev, A. B.; Abramov, V. Ya.; Belous, V. N.

    1996-10-01

    Austenitic stainless steels are being considered as structural materials for first wall/blanket systems in the International Thermonuclear Reactor (ITER). The uniform corrosion of stainless steels in water is well known and is not a critical issue limiting its application for the ITER design. The sensitivity of austenitic steels to intergranular corrosion (IGC) can be estimated rather accurately by means of calculation methods, considering structure and chemical composition of steel. There is a maximum permissible carbon content level, at which sensitization of stainless steel is eliminated: K = Cr eff - αC eff, where α-thermodynamic coefficient, Cr eff-effective chromium content (regarding molybdenum influence) and C eff-effective carbon content (taking into account nickel and stabilizing elements). Corrosion tests for 16Cr11Ni3MoTi, 316L and 316LN steel specimens, irradiated up to 2 × 10 22 n/cm 2 fluence have proved the effectiveness of this calculation technique for determination of austenitic steels tendency to IGC. This method is directly applicable in austenitic stainless steel production and enables one to exclude complicated experiments on determination of stainless steel susceptibility to IGC.

  13. 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.

  14. 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)

  15. 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)

  16. 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.

  17. 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.

  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. Phytosanitary Irradiation

    PubMed Central

    Hallman, Guy J.; Blackburn, Carl M.

    2016-01-01

    Phytosanitary treatments disinfest traded commodities of potential quarantine pests. Phytosanitary irradiation (PI) treatments use ionizing radiation to accomplish this, and, since their international commercial debut in 2004, the use of this technology has increased by ~10% annually. Generic PI treatments (one dose is used for a group of pests and/or commodities, although not all have been tested for efficacy) are used in virtually all commercial PI treatments, and new generic PI doses are proposed, such as 300 Gy, for all insects except pupae and adult Lepidoptera (moths). Fresh fruits and vegetables tolerate PI better than any other broadly used treatment. Advances that would help facilitate the use of PI include streamlining the approval process, making the technology more accessible to potential users, lowering doses and broadening their coverage, and solving potential issues related to factors that might affect efficacy. PMID:28231103

  20. 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.